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Interim Assessment of the Denatured 233U Fuel Cycle: Feasibility and Nonproliferation Characteristics


National Technical Information Service U.S. Department of Commerce 5285 Port Royal Road, Springfield, Virginia 22161 Price: Printed Copy $11.75; Microfiche $3.00

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government norany agency thereof, nor any of their employees, contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility for any third party's use or the results of such use of any information, apparatus, product or process disclosed in this report, nor represents that its use by such third party would not infringe privately owned rights.

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ORNL-5388 Distribution Category UC-80

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Contract No. W-7405-eng-26

Engineering Physics D i v i s i o n

INTERIM ASSESSMENT OF THE DENATURED

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23%

FUEL CYCLE:

FEASIBILITY AND NONPROLIFERATION CHARACTERISTICS Edited by

L. S. Abbott, D. E. Bartine, T. J . Burns NOTICE nport was prepared u m account of w o k vponsond by the United Sates Gwemmnt. Neither the United Sulci nor the United States Depvtmcnt of Fnerw, nor MY of their amployus. nor MY of their

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W i t h Contributions from

Argonne N a t i o n a l L a b o r a t o r y Brookhaven N a t i o n a l L a b o r a t o r y Combustion E n g i n e e r i n g , Inc. Hanf o r d E n g i n e e r i n g D e vel opment L a b o r a t o r y Oak R i d g e Gaseous D i f f u s i o n P l a n t Oak R i d g e N a t i o n a l L a b o r a t o r y

wnlnnon, wbcontnc:mr, or meir employas, mkn m y warranty, cxpnn or implied. or p ~ y m c sm y kgnl liability or mpmIJility for the -cy. cnnpleteneu OT Urefulnnr of m y infomution. ap1an:uI. product or proocn dirlacd. or rcpnrnb du: itr we would no: hfrhlEC D r i W C l V o m d dhb.

C. Sege DOE Program Manager D. E. Bartine ORNL Program Manager

I.Spiewak ORNL Program Director

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December, 1978 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 operated by UNION CARBIDE CORPORATION f o r the DEPARTMENT OF ENERGY DISTRIB~ION c OF THIS DOCUMENT 19 UN


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PRINCIPAL AUTHORS

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T. J. Burns Oak Ridge National Laboratory

J. C. Cleveland Oak Ridge National Laboratory ld

E. H. G i f t Oak Ridge Gaseous Diffusion Plant

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R. W. Hardie Hanford Engineering Development Laboratory C. M. Newstead Brookhaven National Laboratory R. P. Omberg Hanford Engineering Development Laboratory

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N. L. Shapiro Combustion Engineering

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I . Spiewak Oak Ridge National Laboratory CONTRIBUTING AUTHORS

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1. B. Arthur, oak Ridge Gaseous Diffusion Plant

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E. Black, Hanford Engineering Development Laboratory E. Brooksbank, Oak Ridge National Laboratory

Y. I. P. M. D. R. T. M. D. T.

Chang , Argonne National Laboratory Haas, Oak Ridge National Laboratory Haffner, Hanford Engineering Development Laboratory Helm, Hanford Engineering Development Laboratory I n g e r s o l l , Oak Ridge National Laboratory

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G. JOl l y , Hanford Engineering Development Laboratory

H. H, D. M,

E. Knee, Oak Ridge National Laboratory

Jenkins, oak Ridge National Laboratory

R. Meyer, Oak Ridge National Laboratory L. $ e l by, Oak Ridge National Laboratory

R. Shay, Hanford Engineering Development Laboratory J. E. T i l 1, Oak Ridge National Laboratory


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PREFACE AND ACKNOWLEDGMENTS T h i s r e p o r t d e s c r i b i n g a study o f the f e a s i b i l i t y o f t h e denatured 233U f u e l c y c l e i n t e g r a t e s t h e data and c o n t r i b u t i o n s o f a number o f n a t i o n a l l a b o r a t o r i e s and government contractors.

Those o f us a t ORNL who have been responsible f o r compiling and e d i t i n g the

r e p o r t wish t o acknowledge t h e assistance o f many i n d i v i d u a l s who a c t i v e l y p a r t i c i p a t e d i n the study throughout t h e many i t e r a t i o n s leading t o t h i s f i n a l d r a f t .

I n particular,

we wish t o thank Carol Sege and Saul Strauch o f t h e U.S. Department o f Energy f o r t h e i r guidance d u r i n g t h e e n t i r e program; C. M. Newstead o f Brookhaven National Laboratory f o r

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t h e p r o l i f e r a t i o n - r i s k assessment; E. H. G i f t o f t h e Oak Ridge Gaseous D i f f u s i o n P l a n t f o r t h e a n a l y s i s o f t h e p o t e n t i a l circumvention o f t h e f u e l isotope b a r r i e r ; Y. Chang o f Argonne National Laboratory, J. C. Cleveland and P. R. Kasten o f Oak Ridge National Laboratory, R. P. Omberg o f t h e Hanford Engineering Development Laboratory, and N. L. Shapiro o f Combustion Engineering, Inc. f o r the c h a r a c t e r i z a t i o n s o f r e a c t o r and f u e l c y c l e technologies; and R.

P. Omberg and

R. W. Hardie o f Hanford Engineering Development

Laboratory f o r t h e system economics-resources analysis.

These, i n turn, were a s s i s t e d

by several c o n t r i b u t i n g authors as l i s t e d on page iii. Many o t h e r s have provided data o r p a r t i c i p a t e d i n reviews o f t h e various chapters, and t o each o f them we express our appreciation. F i n a l l y , we wish t o thank the many s e c r e t a r i e s and r e p o r t production s t a f f members who have so p a t i e n t l y prepared numerous d r a f t s o f t h i s report. I r v i n g Spiewak David E. B a r t i n e Thomas J. Burns L o r r a i n e S. Abbott


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CONTENTS

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....................................................... ABSTRACT .......................................................................... 1. INTRODUCTION : BACKGROUND ..................................................... 2 . RATIONALE FOR DENATURED FUEL CYCLES ........................................... 2.0. I n t r o d u c t i o n ............................................................ 2.1. I n t e r n a t i o n a l Plutonium Economy ......................................... 2.2. The Denatured 233U Fuel Cycle ........................................... 2.3. Some I n s t i t u t i o n a l Considerations f o r t h e Denatured Fuel Cycle .......... 3 . ISOTOPIC CHARACTERISTICS OF DENATURED 233U FUEL ............................... 3.0. I n t r o d u c t i o n ............................................................ 3.1. Estimated 232U Concentrations i n Denatured 233U Fuels ................... 3.1.1. Light-Water Reactor Fuels ....................................... 3.1.2. High-Temperature Gas-Cooled Reactor Fuels ....................... 3.1.3. Liquid-Metal Fast Breeder Reactor Fuels ......................... 3.1.4. Conclusions ..................................................... 3.2. Radiological Hazards o f Denatured Fuel Isotopes ......................... 3.2.1. T o x i c i t y o f 233U and 232U ....................................... 3.2.2. T o x i c i t y o f 232Th ............................................... 3.2.3. Hazards Related t o Gama-Ray Emissions .......................... 3.2.4. Conclusions ..................................................... 3.3. I s o t o p i c s Impacting Fuel Safeguards Considerations ...................... 3.3.1. Enrichment C r i t e r i a f o r Denatured Fuel .......................... 3.3.2. F a b r i c a t i o n and Handling o f Denatured Fuel ...................... 3.3.3. Detection and Assay o f Denatured Fuel ........................... 3.3.4. P o t e n t i a l Circumvention o f I s o t o p i c B a r r i e r o f Denatured Fuel ............................................... 3.3.5. Deterrence Value o f 232U Contamination i n Denatured Fuel ........ 4 . IMPACT OF DENATURED 233U FUEL ON REACTOR PERFORMANCE .......................... 4.0. I n t r o d u c t i o n ............................................................ 4.1. Light-Water Reactors .................................................... 4.1 .1. Pressurized Water Reactors ...................................... 4.1.2. B o i l i n g Water Reactors .......................................... 4.2. Spectral-Shift-Control1 ed Reactors ...................................... PREFACE AND ACKNOWLEDGMENTS

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4.3.

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4.4.

4.5. 4.6.

.................................................... Gas-Cooled Thermal Reactors ...................... .,...................... 4.4.1. High-Temperature Gas-Cooled Reactors ............................ 4.4.2. Pebble-Bed High-Temperature Reactors ............................ Liquid-Metal Fast Breeder Reactors ...................................... A1 t e r n a t e Fast Reactors ................................................. 4.6.1. Advanced Oxide-Fueled LMFBRs .................................... 4.6.2. Carbide- and Metal -Fueled LMFBRs ................................ Heavy-Water Reactors

4.6.3.

Gas-Cooled Fast Breeder Reactors

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xi 1-1 2-1 2-3 2-4 2-5 2-9 3-1 3-3 3-6 3-6 3-7 3-8 3-9 3-10 3-10 3-14 3-14 3-15 3-17 3-17 3-20 3-22 3-24 3-35 4-1

4-3 4-12 4-12 4-19 4-23 4-30 4-33 4-33 4-41 4-48 4-54 4-54 4-58 4-62


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....................................... ............................................................ 5.1. Reactor Research and Development Requirements ........................... 5.1.1. Light-Water Reactors ............................................ 5.1.2. High-Temperature Gas-Cooled Reactors ............................ 5.1.3. Heavy-Water Reactors ............................................ 5.1.4. Spectral-Shift-Controlled Reactors .............................. 5.1.5. R,D&D Schedules ................................................. 5.1.6. Summary and Conclusions ......................................... 5.2. Fuel Recycle Research and Development Requirements ...................... 5.2.1. Technology Status Sumnary ....................................... 5.2.2. Research, Development, and Demonstration Cost Ranges and Schedules ................................................... 5.2.3. Conclusions ..................................................... EVALUATION OF NUCLEAR POWER SYSTEMS UTILIZING DENATURED FUEL .................. 6.0. I n t r o d u c t i o n ............................................................ 6.1. Basic Assumptions and Analysis Technique ................................ 6.1 .l. The U308 Supply ................................................. 6.1.2. Reactor Options ................................................. 6.1.3. Nuclear P o l i c y Options .......................................... 6.1.4. The A n a l y t i c a l Method ........................................... IMPLEMENTATION OF DENATURED FUEL CYCLES 5.0. Introduction

............... 6.2.1. The Throwaway/Stowaway Option ................................... 6.2.2. Converter System w i t h Plutonium Recycle ......................... 6.2.3. Converter System w i t h Plutonium Throwaway ....................... 6.2.4. Converter System w i t h Plutonium Production Minimized; P u - ~ o - ~ "Transmutation" ~ ~ U ...................................... 6.2.5. Converter System w i t h Plutonium Production Not Minimized; P u - ~ o - ~3U "Transmutation" ...................................... 6.2.6. Converter-Breeder System w i t h L i g h t Plutonium "Transmutation" ................................................. 6.2.7. Converter-Breeder System w i t h Heavy Plutonium "Transmutation" ................................................. 6.3. Conclusions ............................................................. OVERALL ANALYSIS OF DENATURED FUEL SYSTEMS .................................... 7.0. I n t r o d u c t i o n ............................................................ 7.1. Pro1 i f e r a t i o n - R e s i s t a n t C h a r a c t e r i s t i c s o f Denatured 233U Fuel .......... 6.2.

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Discussion o f Results f o r Selected Nuclear P o l i c y Options

.................................. ........................... ...................................... ..................................................... Impact o f Denatured 233U Fuel on Reactor Performance and Selection: Comparison w i t h Other Fuel Cycles ....................................... 7.2.1. Thermal Reactors ................................................ Once-Through Systems ....................................... Recycle Systems ............................................ 7.2.2. Fast Reactors ................................................... 7.2.3. Symbiotic Reactor Systems ....................................... 7.2.4. Conclusions ..................................................... Prospects f o r Implementation and Comnercialization o f Denatured 233U Fuel Cycle ......................................................... 7.1.1. 7.1.2. 7.1.3. 7.1.4.

7.2.

7.3.

I s o t o p i c B a r r i e r o f Fresh Fuel Gamma-Radiation B a r r i e r o f Fresh Fuel Spent Fuel F i s s i l e Content Conclusions

5-1 5-3 5-4 5-8 5-1 1 5-13 5-14 5-1 7 5-17 5-21

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5-21 5-24 5-26 6- 1 6-3

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6-5 6-5 6-6 6-10

6-1 1 6-23 6-23 6-30 6-33

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6-35 6-38 6-41 6-44 6-47 7-1 7-3 7-4 7-4 7-6 7-7 7-9 7-10 7-10 7-10 7-13 7-14 7-16 7-19 7-21

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7.3.1. 7.3.2. 7.3.3. 7.4.

7.4.4. 7.4.5. 7.4.6.

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Adequacy o f Nuclear Power Systems U t i l i z i n g Denatured 233U Fuel f o r Meeting E l e c t r i c a l Power Demands 7.4.1. 7.4.2. 7.4.3.

7.5.

Possible Procedure f o r Implementing and Comnercial i z i n g t h e Denatured Fuel Cycle Considerations i n Commercial i z i ng Reactors Opera t i ng on A l t e r n a t e Fuels Conclusions

.................................... ........................................... ....................................................... ...................... .................... ........................ ................. ....... .....................................................

The A n a l y t i c a l Method Data Base Results f o r Price-Limi t e d Uranium Supplies Non-FBR Systems. Options 1. 2. 4. and 5 FBR Systems. Options 3. 6. 7. and 8 Results f o r Unconstrained Resource A v a i l a b i l i t y Systems Employing Improved LWRs and Enrichment Technology Conclusions

Tradeoff Analysis and Overall Strategy Considerations

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.............................................. ................................................. ......................... APPENDICES ........................................................................ App . A . ISOTOPE SEPARATION TECHNOLOGIES ..................................... A.l. Current Separation C a p a b i l i t y ................................. The Gaseous D i f f u s i o n Process .............................. 7.5.1. 7.5.2. 7.5.3.

No-Recycle Options Recycle Options Overall Conclusions and Recomnendations

................................. ............................... .................................. .................. A.2. ................................... ............................ .................................. ........................................ ..................................... ................. App . B . ECONOMIC DATA BASE USED FOR EVALUATIONS OF NUCLEAR POWER SYSTEMS ............................................................. The Gas Centrifuge Process The Becker Separation Nozzle The South Helikon Process Current and Projected Enrichment Capacity New Separation Technologies Photoexcitation (Laser) Methods Chemical Exchange Methods Aerodynamic Methods Plasma-Based Processes Comparison o f Advance Separation Processes

. C. App . D . App

DETAILED RESULTS FROM EVALUATIONS OF VARIOUS NUCLEAR POWER SYSTEMS UTILIZING DENATURED FUEL

.................................... CALCULATIONS OF NUCLEAR AND FOSSIL PLANT COMPETITION BASED ON ECONOMICS ........................................................

7-23 7-25 7-27 7-29 7-30 7-31 7-31 7-33 7-35 7-35 7-38 7-40 7-42 7-43 7-44 7-'48 A-1 A-3 A-3 A-3 A- 3 A- 7 A-9 A-10 A-10 A-13 A-15 A-17 A-17 A-18 B-1 c-1 D- 1


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ABSTRACT

A f u e l c y c l e t h a t employs 23% denatured w i t h

23811

and mixed w i t h thorium f e r t i l e

m a t e r i a l i s examined w i t h respect t o i t s p r o l i f e r a t i o n - r e s i s t a n c e c h a r a c t e r i s t i c s and i t s technical and economic f e a s i b i l i t y .

The r a t i o n a l e f o r considering t h e denatured 233U f u e l

c y c l e i s presented, and the impact o f the denatured f u e l on the performance o f Light-Water Reactors, Spectral-Shift-Control l e d Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors i s discussed.

The scope o f t h e R,D&D programs t o commercialize

these r e a c t o r s and t h e i r associated f u e l cycles i s a l s o summarized and t h e resource requirements and economics o f denatured 233U cycles are compared t o those o f t h e conventional Pu/U cycle.

I n addition, several nuclear power systems t h a t employ denatured 23% f u e l

and are based on t h e energy center concept are evaluated.

Under t h i s concept, dispersed

power r e a c t o r s fueled w i t h denatured o r low-enriched uranium f u e l are supported by secure energy centers i n which s e n s i t i v e a c t i v i t i e s o f t h e nuclear c y c l e are performed. These a c t i v i t i e s i n c l u d e 233U production by Pu-fueled "transmuters" (thermal o r f a s t reactors) and reprocessing. A summary chapter presents t h e most s i g n i f i c a n t conclusions from the study and recommends areas f o r f u t u r e work.

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CHAPTER 1 INTRODUCTION: BACKGROUND I

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B a r t i n e , L. S. Abbott, and T. J. Burns Oak Ridge National Laboratory


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INTRODUCTION:

BACKGROUND

I n t h e mid-l940s, as t h e nuclear era was j u s t beginning, a p r e s t i g i o u s group i n c l u d -

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i n g Robert Oppenheimer and l e d by David L i l i e n t h a l , t h e f i r s t chairman o f t h e

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Energy Comnission, was comnissioned by Under Secretary o f S t a t e Dean Acheson t o recommend ways t h a t t h e b e n e f i t s o f nuclear energy could be shared with t h e world w i t h o u t t h e dangers o f what we now r e f e r t o as "nuclear p r o l i f e r a t i o n " : t h a t i s , t h e c r e a t i o n o f numerous

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U.S. Atomic

nuclear weapons states. The r e p o r t ' they submitted s t a t e s t h a t "the proposed s o l u t i o n i s an i n t e r n a t i o n a l i n s t i t u t i o n and framework o f t r e a t i e s and agreements f o r cooperative

A t t h e same time, t h e committee proposed

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operation o f s e n s i t i v e nuclear technology."

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several p o s s i b l e technological developments t o help implement an i n t e r n a t i o n a l system, They a l s o suggested t h e r e s t r i c t i o n o f t h e

i n c l u d i n g t h e denaturing of r e a c t o r f u e l s . . -

most s e n s i t i v e a c t i v i t i e s w i t h i n a nuclear c y c l e t o nuclear energy a r e n a s .

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Lili I n t h e subsequent years several steps have been taken toward i n t e r n a t i o n a l cooperat i o n i n t h e p o l i t i c a l c o n t r o l o f t h e p o t e n t i a l f o r making nuclear weapons. Atoms f o r Peace Program was i n i t i a t e d by t h e U.S.

I n 1953 t h e

and i n 1957 t h e I n t e r n a t i o n a l Atomic

Energy Agency was formed, one o f i t s chartered r e s p o n s i b i l i t i e s being t h e safeguarding o f

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f i s s i l e m a t e r i a l and t h e reduction o f t h e p o t e n t i a l f o r t h e production o f nuclear weapons. I n 1970 these e f f o r t s r e s u l t e d i n a n o n p r o l i f e r a t i o n t r e a t y t h a t was d r a f t e d by t h e U.S. and t h e U.S.S.R.

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As t h e d i a l o g has continued, i n e v i t -

and subscribed t o by 116 nations.

a b l y a l l serious studies o f t h e problem, i n c l u d i n g t h e most recent studies, have a r r i v e d a t t h e same conclusion as the Acheson comnittee: w i t h technological supports a r e mandatory

i n t e r n a t i o n a l cooperation and safeguards

-- o r t o s t a t e

i t another way, no p u r e l y tech-

n o l o g i c a l f i x t o prevent nuclear p r o l i f e r a t i o n i s possible. . -

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I t was against t h i s background and l a r g e l y through t h e i n i t i a t i v e s o f President

Carter t h a t an I n t e r n a t i o n a l Nuclear Fuel Cycle Evaluation Program (INFCE) was established i n the Fall of 1977 t o study how p r o l i f e r a t i o n - r e s i s t a n t nuclear fuel cycles could be developed f o r world-wide nuclear generation of e l e c t r i c a l power. A t t h e same time a U.S. N o n p r o l i f e r a t i o n A1 t e r n a t i v e Systems Assessment Program (NASAP) was formed t o c a r r y o u t i n t e n s i v e studies t h a t would both provide i n p u t t o INFCE and recommend t e c h n i c a l and i n s t i t u t i o n a l approaches t h a t could be implemented w i t h various nuclear f u e l cycles

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proposed f o r t h e U.S. The p r i n c i p a l p r o l i f e r a t i o n concern i n c i v i l i a n nuclear power f u e l cycles i s t h e poss i b l e d i v e r s i o n o f f i s s i l e m a t e r i a l t o t h e f a b r i c a t i o n of nuclear weapons.

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s u f f i c i e n t q u a n t i t i e s , t h e f i s s i l e m a t e r i a l employed i n any nuclear f u e l c y c l e can be processed i n t o weapons-usable m a t e r i a l ,but

f u e l cycles t h a t are considered t o o f f e r t h e l e a s t

r e s i s t a n c e t o d i v e r s i o n a r e those t h a t i n c l u d e weapons usable m a t e r i a l t h a t can be chemic a l l y separated from a l l t h e o t h e r m a t e r i a l s

in the

cycle.

The 235U

in

t h e low-enriched

uranium (LEU) f u e l used by c u r r e n t l y operating Light-Water Reactors (LWRs) cannot be chemic a l l y separated because i t i s embedded i n a m a t r i x of 238U. To e x t r a c t t h e 235U from t h e 2 3 e U


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would r e q u i r e i s o t o p i c separation which i s t e c h n o l o g i c a l l y d i f f i c u l t and f o r which few f a c i l i t i e s i n t h e world c u r r e n t l y e x i s t .

The uranium m i x t u r e i t s e l f could n o t be used f o r

weapons f a b r i c a t i o n because t h e concentration o f t h e f i s s i l e component i s t o o low.

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By contrast, t h e plutonium i n t h e Pu/U mixed oxide f u e l c y c l e developed f o r f a s t

breeder reactors such as t h e L i q u i d Metal Fast Breeder (LFMBR) can be chemically separated from t h e o t h e r m a t e r i a l s i n t h e cycle. Thus, as p r e s e n t l y developed, t h e Pu/U f u e l c y c l e i s perceiyed t o be l e s s p r o l i f e r a t i o n r e s i s t a n t than t h e LEU cycle. This facet o f t h e FBR-Pu/U f u e l c y c l e was obviously a major f a c t o r i n t h e A d m i n i s t r a t i o n ' s decision i n A p r i l , 1977, t o defer commercialization o f t h e LMFBR i n t h e United States.

, Another concern about plutonium centers on i t s presence i n t h e "back end" o f t h e LEU f u e l cycle.

While i t does n o t e x i s t i n t h e " f r o n t end" o f t h e c y c l e ( t h a t i s , i n t h e

f r e s h f u e l ) , plutonium i s produced i n t h e tions. able.

238U

o f t h e f u e l elements during r e a c t o r opera-

Thus t h e spent LWR elements contain f i s s i l e plutonium t h a t i s chemically e x t r a c t The f u e l c y c l e technology includes steps f o r reprocessing t h e elements t o recover

and r e c y c l e t h e plutonium, together w i t h other unburned f i s s i l e m a t e r i a l i n t h e elements, b u t t o date t h i s has n o t been done i n t h e U.S. and c u r r e n t l y a moratorium on U.S.' comnercial reprocessing i s i n e f f e c t .

As a r e s u l t , t h e spent f u e l elements now being removed from

LWRs are being stored on s i t e .

Because i n i t i a l l y they are h i g h l y r a d i o a c t i v e due t o a

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f i s s i o n - p r o d u c t buildup, t h e spent elements must be h e a v i l y shielded, b u t as t h e i r r a d i o a c t i v i t y decays w i t h time l e s s s h i e l d i n g w i l l be required. Various nuclear " a l t e r n a t i v e s " are being proposed by t h e U.S. and o t h e r c o u n t r i e s f o r i n t e r n a t i o n a l consideration i n l i e u o f t h e c l a s s i c a l Pu/U cycle. One proposal i s t h a t nations continue marketing LWRs and o t h e r types o f thermal reactors f u e l e d with n a t u r a l o r low-enriched uranium. A moratorium on reprocessing would be adopted, and t h e spent f u e l would be s t o r e d i n secure n a t i o n a l o r i n t e r n a t i o n a l centers such as has r e c e n t l y been proposed by t h e United States, t h e s e c u r i t y o f t h e f u e l being transported t o t h e centers being provided by i t s f i s s i o n - p r o d u c t r a d i o a c t i v i t y .

This scenario assumes a guarantee t o the nuclear-power-consuming nations o f a f u e l supply f o r t h e approximately 30-year economic l i f e o f t h e i r nuclear plants. Other proposals t h a t assume t h e absence o f reprocessing (and thus do n o t i n c l u d e r e c y c l e o f uranium and/or plutonium) are aimed a t improving t h e in-sttu u t i l i z a t i o n o f f i s s i l e m a t e r i a l within t h e framework o f c u r r e n t l i g h t - w a t e r technology. Light-water r e a c t o r options such as improved r e f u e l i n g patterns and c y c l e "coastdownn procedures, as w e l l as more extensive m o d i f i c a t i o n s (such as increasing t h e design burnup), are being studied.

S i g n i f i c a n t gains i n resource u t i l i z a t i o n a l s o appear p o s s i b l e w i t h t h e i n t r o -

duction o f "advanced converter" designs based on Heavy-Water Reactors (HWRs) , Spectral S h i f t - C o n t r o l l e d Reactors (SSCRs) , o r High-Temperature Gas-Cooled Reactors (HTGRs).

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While these various proposals could be useful f o r increasing t h e energy generated

b

from t h e uranium resource base w h i l e r e c y c l i n g i s disallowed, they w i l l n o t provide t h e "inexhaustible" supply o f nuclear f u e l t h a t has been a n t i c i p a t e d from t h e connnercial i z a t i o n o f f u e l r e c y c l e and breeder reactors.

To provide such a supply would r e q u i r e t h e separation

and reuse o f t h e " a r t i f i c i a l " f i s s i l e isotopes 239Pu and 233U.

It was under t h e assumption

t h a t r e c y c l e would occur, i n i t i a l l y i n LWRs, t h a t t h e technology f o r t h e Pu/U mixed-oxide

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fuel cycle, i n which 239Pu i s bred from 238U, was developed.

However, f o r t h e reasons

s t a t e d above, t h e p r o l i f e r a t i o n r e s i s t a n c e o f t h e c y c l e as c u r r e n t l y developed i s perceived as being inadequate.

I t s pro1 i f e r a t i o n r e s i s t a n c e could be increased by d e l i b e r a t e l y

k

" s p i k i n g " t h e f r e s h f u e l elements w i t h r a d i o a c t i v e contaminants o r a l l o w i n g them t o r e t a i n

id

some o f t h e f i s s i o n products from t h e previous cycle, e i t h e r o f which would discourage

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seizure by unauthorized groups o r states.

The f e a s i b i l i t y o f these and other p o s s i b l e

m o d i f i c a t i o n s t o t h e c y c l e are c u r r e n t l y under study.

I n a d d i t i o n , t h e employment o f

f u l l - s c o p e safeguards, i n c l u d i n g extensive f i s s i l e monitoring procedures, i s being i n v e s t i g a t e d f o r use w i t h t h e Pu/U cycle.

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Also under study a r e several " a l t e r n a t e " f u e l cycles based on t h e use o f t h e a r t i f i c i a l f i s s i l e isotope 2 3 % which i s bred i n 232Th. One such c y c l e i s t h e 33U/238U/2 32Th c y c l e proposed by Feiveson and Taylor,2 and i t i s t h i s c y c l e t h a t i s t h e s u b j e c t o f t h i s report.

I n t h e 23%/238U/232Th f u e l c y c l e t h e

denaturant. The f e r t i l e isotope 232Th i s included t o breed a d d i t onal 233U. The a d d i t i o n o f t h e 238U denaturant makes t h e proposed f u e l c y c l e sim l a r t o the 235U/238U c y c l e c u r r e n t l y employed i n

LWRs i n t h a t e x t r a c t i n g t h e

r e q u i r e isotope separation f a c i l i t i e s .

id

23% i s mixed w i t h 238U which serves as a

233U

f o r weapons f a b r i c a t i o n would

Since 23% does n o t occur i n nature, t h e c y c l e i s

a l s o s i m i l a r t o t h e 239Pu/238U c y c l e i n t h a t reprocessing w i l l be necessary t o u t i l i z e t h e bred f u e l .

However, as suggested by t h e Acheson Comnittee and again by Feiveson and Taylor,

reprocessing and o t h e r s e n s i t i v e a c t i v i t i e s could be r e s t r i c t e d t o secure energy centers

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and s t i l l a l l o w power t o be generated outside the centers. It i s t h e purpose o f t h i s r e p o r t t o assess i n t h e l i g h t o f today's knowledge t h e p o t e n t i a l o f t h e denatured

233U f u e l c y c l e f o r meeting t h e requirements f o r e l e c t r i c a l

power growth w h i l e a t t h e same time reducing p r o l i f e r a t i o n r i s k s .

Chapter 2 examines

t h e r a t i o n a l e f o r u t i l i z i n g t h e denatured f u e l c y c l e as a reduced p r o l i f e r a t i o n measure, and Chapter 3 attempts t o assess t h e impact o f t h e i s o t o p i c s o f the cycle, e s p e c i a l l y w i t h respect t o an imp1 i e d t r a d e o f f between chemical i n s e p a r a b i l i t y and i s o t o p i c s e p a r a b i l i t y o f t h e f u e l components. Chapter 4 examines t h e neutronic performance o f various r e a c t o r types u t i l i z i n g denatured 233U f u e l , and Chapter 5 discusses t h e r e q u i r e _ I

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I ,

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ments and p r o j e c t i o n s f o r implementing t h e cycle. ear power systems u t i l i z i n g denatured f u e l .

Chapter 6 then evaluates various nucl-

F i n a l l y , Chapter 7 gives sumnations of t h e

safeguards considerations and r e a c t o r neutronic and symbiotic aspects and discusses t h e prospects f o r deploying denatured r e a c t o r systems.

Chapter 7 a l s o presents t h e o v e r a l l

conclusions and recommendations r e s u l t i n g from t h i s study.


.

f

1-6

c' The reader w i l l note t h a t throughout t h e study the U.S. has been used as t h e base case.

This was necessary because t h e a v a i l a b l e i n p u t data

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t h a t i s , resource base

estimates, p r o j e c t e d r e a c t o r and f u e l c y c l e development schedules , and assumed power growth r a t e s -- are a l l o f U.S. o r i g i n . However, w i t h access t o corresponding data for an i n t e r n a t i o n a l base, t h e study could be scaled upward t o cover an interdependent world model. References f o r Chapter 1

1.

, , 2.

"A Report on t h e I n t e r n a t i o n a l Control o f Atomic Energy," prepared f o r t h e Secretary o f S t a t e ' s Committee on Atomic Energy by a Board o f Consultants: Chester I. Barnard, Dr. J. R. Oppenheimer, Dr. Charles A. Thomas, Harry Winne, and David E. L i l i e n t h a l (Chairman), Washington, D.C., March 16, 1946, pp. 127-213, Department o f S t a t e P u b l i c a t i o n 2493. H. A. Feiveson and T.

B. Taylor, "Security I m p l i c a t i o n s

o f A l t e r n a t i v e F i s s i o n Futures,"

BUZZ. Atom& S c i e n t i s t s , p. 14 (December 1976).

c

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CHAPTER 2 RATIONALE FOR DENATURED FUEL CYCLES

T.

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J. Burns

Oak Ridge National Laboratory

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Chapter O u t l i n e

I: 2.0.

Introduction

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2.1.

I n t e r n a t i o n a l Plutonium Economy

2.2.

The Denatured

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2.3.

Some I n s t i t u t i o n a l Considerations o f the Denatured Fuel Cycle

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233U

Fuel Cycle


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2.0.

INTRODUCTION

The primary r a t i o n a l e f o r considering the pro1i f e r a t i o n p o t e n t i a l o f t h e nuclear f u e l cycles associated w i t h c i v i l i a n power reactors derives from two opposing concerns: t h e p o s s i b i l i t y o f nuclear weapons p r o l i f e r a t i o n versus a need f o r and t h e perceived economic/resource b e n e f i t s o f a nuclear-based generating capacity. A t t h e o u t s e t i t should be emphasized t h a t a c i v i l i a n nuclear power program i s n o t t h e o n l y p r o l i f e r a t i o n r o u t e a v a i l a b l e t o nonnuclear weapons states.

The countries t h a t have developed nuclear explosives

t o date have n o t r e l i e d on a c i v i l i a n nuclear power program t o o b t a i n t h e f i s s i l e material.

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Rather, they have u t i l i z e d enrichment f a c i l i t i e s , plutonium-production reactors, and, more r e c e n t l y , a research reactor.

Moreover, as opposed t o a d e l i b e r a t e (and p o s s i b l y clande-

s t i n e ) weapons-development program based upon a n a t i o n a l decision, nuclear power programs are c u r r e n t l y s u b j e c t t o i n t e r n a t i o n a l monitoring and i n f l u e n c e i n most cases. Thus w h i l e c i v i l i a n nuclear power does represent one conceivable p r o l i f e r a t i o n route, i f i t i s made

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l e s s a t t r a c t i v e than o t h e r p o s s i b l e routes, p r o l i f e r a t i o n concerns should n o t i n h i b i t t h e development o f comnercial nuclear power.

bl

Pro1 i f e r a t i o n concerns regarding c i v i 1i a n nuclear power programs center on two i n t r i n s i c c h a r a c t e r i s t i c s o f t h e ' n u c l e a r f u e l cycle.

F i r s t , nuclear r e a c t o r f u e l

i n h e r e n t l y provides a p o t e n t i a l source o f f i s s i l e m a t e r i a l from which production o f

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weapons-grade m a t e r i a l i s possible.

t

enrichment and reprocessing f a c i l i t i e s , exacerbate t h e p r o l i f e r a t i o n problem since they

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Second, c e r t a i n f u e l c y c l e components, p a r t i c u l a r l y

provide a technological c a p a b i l i t y which could be d i r e c t e d towards weapons development. The term " l a t e n t p r o l i f e r a t i o n ' has been coined by Feiveson and Taylor' t o cover these c h a r a c t e r i s t i c s o f t h e nuclear f u e l c y c l e which, although 'not p e r t a i n i n g d i r e c t l y t o weapons development, by t h e i r existence f a c i l i t a t e a p o s s i b l e f u t u r e decision t o e s t a b l i s h such a c a p a b i l i t y .

id I t should be noted t h a t t h e problem o f l a t e n t p r o l i f e r a t i o n impacts even t h e "oncethrough" low-enriched uranium (LEU) c y c l e c u r r e n t l y employed i n 1ight-water r e a c t o r s (LWRS)

and a l s o t h e natural-uranium c y c l e u t i l i z e d i n t h e Canadian heavy-water systems (CANDUs). The technology r e q u i r e d t o e n r i c h n a t u r a l uranium t o LWR fuel represents a technological c a p a b i l i t y which could be r e d i r e c t e d from.peacefu1 purposes. I n a d d i t i o n , the plutoniumc o n t a i n i n g spent f u e l , a l b e i t d i l u t e and contaminated w i t h h i g h l y r a d i o a c t i v e f i s s i o n . products, represents a source o f p o t e n t i a l weapons m a t e r i a l .

Thus t h e p o s s i b i l i t y o f

p r o l i f e r a t i o n e x i s t s even f o r t h e f u e l cycles now i n use. This has already been recogn i z e d and i t has been proposed1n2 t h a t i n t e r n a t i o n a l l y c o n t r o l l e d f u e l c y c l e s e r v i c e

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centers be established whose purpose would be t o preclude subversion o f s e n s i t i v e technology (such as enrichment technology) and t o provide f a c i l i t i e s f o r t h e assay and secure storage o f spent once-through r e a c t o r f u e l .


2-4

The establishment of such fuel cycle service centers i s currently receiving serious consideration. As the costs of U308 production increase (and as i t i s preceived t h a t longterm reliance on nuclear power i s necessary), the expansion of the fuel cycle service center t o include reprocessing a c t i v i t i e s will become attractive. The expansion would allow the I t would also allow the a r t i f i c i a l ( t h a t 235U remaining in the spent fuel t o be utilized. i s , "manufactured") f i s s i l e isotopes produced as a direct result of the power production process t o be recycled. Of the l a t t e r , only two possible candidate isotopes exist: 239Pu In considering these two isotopes, i t appears that the proliferation aspects of and 233U. t h e i r possible recycle scenarios are considerably different. In f a c t , the rationale f o r the present study i s the need t o determine whether 233U-based recycle scenarios have significant pro1 iferation-resistant advantages compared w i t h plutonium-based recycle scenarios. 2.1.

INTERNATIONAL PLUTONIUM ECONOMY

Prior t o President Carter's April 7, 1977, nuclear policy statement, the reference recycle fuel scenario had been based on plutonium, referred t o by Feiveson and Taylor' as the "plutonium economy." In t h i s scenario the plutonium generated i:: the LEU cycle would be'recycled as feed material f i r s t into thermal reactors and l a t e r into f a s t breeders, these reactors then operating on mixed Pu/U oxides instead of on uranium oxide alone. As w i t h any recycle scenario, the plutonium-based nuclear power economy would require the operation of spent fuel reprocessing f a c i l i t i e s . I f dispersed throughout the world, such reprocessing technology, 1ike uranium enrichment technology, would markedly increase the l a t e n t proliferation potential inherent i n the nuclear fuel cycle. Of course, such f a c i l i t i e s could also be restricted t o the fuel cycle service centers. However, the plutonium recycle scenario introduces a f a r greater concern regarding nuclear proliferation since weapons-usable material can be produced from the fresh mixed oxide fuel through chemicaZ sepamtion of the plutonium from the uranium, whereas t o obtain weapons-usable material from LEU fuel requires isotopic enrichment i n 235U. Since the fresh mixed oxide (Pu/U) fuel of the reference cycle is vulnerable t o chemical separation, not only are the fuel fabrication f a c i l i t i e s of the cycle potential sources of directly usable weapons material, b u t also the reactors themselves. While restriction of mixed oxide fabrication f a c i l i t i e s t o safeguarded centers i s both feasible and advisable, i t i s unlikely t h a t the reactors can be centralized into a few such internationally controlled centers. Rather they will be dispersed outside the centers, which will necessitate t h a t fresh fuel containing plutonium be shipped and stockpiled on a global scale and that i t be safeguarded a t a l l points. Thus, as pointed out by Feiveson and Taylor,' the plutonium recycle scenario significantly increases the number of nuclear fuel cycle f a c i l i t i e s which must be safeguarded. The prospect of such widespread use of plutonium and i t s associated problems of security have led t o an examination of possible alternative fuel cycles aimed a t reducing the proliferation risk inherent i n recycle scenarios. One such alternative fuel cycle i s the denatured 233U fuel cycle which comprises the subject of this report.


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2.2. I n t h e denatured

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THE DENATURED 233U FUEL CYCLE

233U cycle, the f r e s h f u e l would c o n s i s t o f a mixture o f f i s s i l e 233U

d i l u t e d w i t h 23eU ( t h e denaturant) and combined w i t h the f e r t i l e isotope thorium.

The pre-

sence o f a s i g n i f i c a n t q u a n t i t y o f 23eU denaturant would preclude d i r e c t use o f t h e f i s s i l e m a t e r i a l f o r weapons purposes even i f t h e uranium and thorium were chemically separated.

As

i n t h e LEU cycle, an a d d i t i o n a l step, t h a t o f i s o t o p i c enrichment o f t h e uranium, t h i s time t o increase i t s 233U concentration, would be necessary t o produce weapons-grade m a t e r i a l ,

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and the development o f an enrichment c a p a b i l i t y would r e q u i r e a s i g n i f i c a n t decision and commitment w e l l i n advance o f t h e actual d i v e r s i o n o f f i s s i l e m a t e r i a l from the f r e s h fuel. This i s i n c o n t r a s t t o t h e reference Pu/U f r e s h f u e l f o r which o n l y chemical separation would be required. Moreover, even i f such an enrichment c a p a b i l i t y were developed, i t would appear t h a t e n r i c h i n g c l a n d e s t i n e l y obtained n a t u r a l uranium would be p r e f e r a b l e t o d i v e r t i n g and e n r i c h i n g r e a c t o r f u e l , whether i t be denatured 233U o r some other type, since the r e a c t o r

-.

f u e l would be more i n t e r n a t i o n a l l y "accountable."

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The primary advantage o f t h e denatured f u e l c y c l e i s the i n c l u s i o n of t h i s " i s o t o p i c b a r r i e r " i n t h e fuel.

Whereas ir, t h e plutonium c y c l e no denaturant comparable t o 238U e x i s t s

and t h e f r e s h f u e l safeguards ( t h a t i s , physical security, i n t e r n a t i o n a l monitoring, etc.) would a l l be external t o the f u e l , the denatured 233U f u e l c y c l e would incorporate an i n I

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herent safeguard advantage as a physical property o f the f u e l i t s e l f .

L i k e t h e plutonium

cycle, t h e denatured f u e l c y c l e would r e q u i r e the development o f f u e l c y c l e centers t o safeguard s e n s i t i v e f u e l c y c l e a c t i v i t i e s such as reprocessing ( b u t n o t necessarily r e f a b r i -

L

cation).

However, u n l i k e t h e plutonium cycle, t h e denatured f u e l c y c l e would n o t r e q u i r e

the extension o f such s t r i n g e n t safeguard procedures t o the reactors themselves, and they are the most numerous component o f the nuclear f u e l cycle. (As noted above, LEU f u e l i s a l s o "denatured" i n the sense t h a t a low concentration o f 2 3 % i s included i n a 23'3U matrix. S i m i l a r l y , n a t u r a l uranium f u e l i s denatured. Thus, these f u e l s a l s o have the p r o l i f e r a t i o n resistance advantages o f the isotopic barrier.) The concept o f denatured 233U f u e l as a p r o l i f e r a t i o n - r e s i s t a n t step i s addressed p r i n c i p a l l y a t t h e f r o n t end o f t h e nuclear f u e l cycle, t h a t i s , t h e f r e s h f u e l charged t o reactors.

The 238U denaturant w i l l , o f course, produce plutonium under i r r a d i a t i o n .

Thus, as i n the LEU and mixed oxide cycles, t h e spent f u e l from t h e denatured c y c l e i s a p o t e n t i a l source o f plutonium. However, a l s o as i n t h e LEU and mixed oxide cycles, t h e

i

plutonium generated i n the spent f u e l i s contaminated w i t h h i g h l y r a d i o a c t i v e f i s s i o n products. Moreover, t h e q u a n t i t y o f plutonium generated v i a t h e denatured f u e l c y c l e w i l l be s i g n i f i c a n t l y l e s s than t h a t o f t h e o t h e r two cycles.

Further, t h e decision t o use spent

r e a c t o r f u e l as a source o f weapons m a t e r i a l requires a previous commitment t o t h e develop-

i;

ment o f shielded e x t r a c t i o n f a c i l i t i e s . I n sumnary, t h e use o f a denatured f u e l as a source o f weapons m a t e r i a l i m p l i e s one o f two s t r a t e g i c decisions: t h e development o f an i s o t o p i c enrichment c a p a b i l i t y t o process d i v e r t e d fresh f u e l , o r the development o f a, f i s c s i l e e x t r a c t i o n c a p a b i l i t y (chemical o r i s o t o p i c ) t o process d i v e r t e d spent fuel.

In


2-6

contrast, w h i l e the plutonium c y c l e a l s o would r e q u i r e a s t r a t e g i c d e c i s i o n concerning t h e spent f u e l , the decision t o u t i l i z e t h e f r e s h mixed oxide f u e l would be e a s i e r and thus would be more t a c t i c a l i n nature.

A subsidiary p r o l i f e r a t i o n - r e l a t e d advantage o f t h e denatured f u e l c y c l e i s t h e The 232U, 232U (and i t s h i g h l y r a d i o a c t i v e decay daughters) i n t h e f r e s h f u e l .

presence o f

an unavoidable byproduct i n t h e production o f 233U from 232Th, c o n s t i t u t e s a chemically inseparable r a d i o a c t i v e contaminant i n t h e f r e s h f u e l , which would be a f u r t h e r d e t e r r e n t t o proliferation.

S i m i l a r contamination o f mixed Pu/U oxide f u e l has been proposed v i a

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"spiking" t h e f u e l w i t h f i s s i o n products o r p r e i r r a d i a t i n g i t t o produce t h e f i s s i o n products i n s i t u , but both these options would i n v o l v e s i g n i f i c a n t p e r t u r b a t i o n s t o t h e P U / ~ ~fuel ~ U c y c l e as opposed t o t h e " n a t u r a l " contamination o f thorium-based fuels.

Additionally, the

a r t i f i c i a l spike o f mixed oxide f u e l would be subject t o chemical e l i m i n a t i o n , a l b e i t req u i r i n g h e a v i l y shielded f a c i l i t i e s . The n a t u r a l spike o f t h e denatured f u e l ( t h a t i s , t h e 232U decay daughters) would a l s o be s u b j e c t t o chemical e l i m i n a t i o n , b u t t h e continuing

decay o f t h e 232U would replace t h e n a t u r a l spike w i t h i n a l i m i t e d p e r i o d o f time. 233U a l s o has t h e advantage o f a higher f i s s i l e worth i n thermal r e a c t o r s than 239Pu,

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both i n terms o f t h e energy release p e r atom destroyed and i n terms o f t h e conversion r a t i o (see Section 4.0). Comnercial thermal r e a c t o r s are c u r r e n t l y a v a i l a b l e and a r e p r o j e c t e d t o enjoy a c a p i t a l c o s t advantage over proposed f a s t breeder reactors.

Additionally, the

technological base r e q u i r e d f o r i n s t a l l a t i o n and operation o f a thermal system i s l e s s s o p h i s t i c a t e d than t h a t f o r f a s t systems such as LMFBRs. Thus i t appears l i k e l y t h a t nearterm scenarios w i l l be dominated by c u r r e n t and proposed thermal systems.

I n considering

possible replacement f i s s i l e m a t e r i a l s f o r t h e l i m i t e d 235U base, t h e worth o f t h e replacement f u e l s i n the thermal systems i s o f some importance. One important f a c t o r which must be considered i n discussing t h e denatured f u e l c y c l e i s t h e p o t e n t i a l source o f t h e r e q u i r e d f i s s i l e material, 233U. It appears l i k e l y t h a t

c c L

current-generation nuclear power r e a c t o r s operating on t h e denatured c y c l e w i l l r e q u i r e an external source of 233U t o provide makeup requirements. Moreover, even i f f u t u r e denatured r e a c t o r s could be designed t o be s e l f - s u f f i c i e n t i n terms o f 233U, t h e r e would s t i l l remain t h e question o f t h e i n i t i a l 2 3 3 U loading. One p o s s i b l e source o f the r e q u i r e d 2 3 3 U is a

233U

production r e a c t o r l o c a t e d i n t h e fuel c y c l e s e r v i c e center (now perhaps more This system would be f u e l e d w i t h plutonium and would

accurately termed an energy center).

both produce power and transmute 232Th i n t o

233U,

L

which could then be denatured f o r use out-

side t h e secure energy center. Loosely termed a transmuter, such a r e a c t o r would be cons t r a i n e d t o the energy center because o f i t s u t i l i z a t i o n o f plutonium f u e l . The r e q u i r e d plutonium f o r the transmuters i s envisioned as coming i n i t i a l l y from reprocessed LEU fuel, and l a t e r , i n t h e more mature system, from plutonium produced i n energy-center r e a c t o r s o r v i a the 238U denaturant i n dispersed reactors. Thus, i n mature form a symbiotic system such as t h a t depicted i n Fig. 2.2-1 w i l l evolve i n which the energy center transmuters produce fuel (233U) f o r the dispersed reactors and consume the plutonium produced by t h e dispersed

L


2-7

denatured reactors o r by energy-center reactors.

L

The dispersed reactors i n t u r n are

provided a source o f 233U f o r i n i t i a l loading and makeup requirements, as w e l l as a means f o r disposing o f the non-recyclable ( i n t h e dispersed reactors) plutonium. The s i g n i f i c a n t p o i n t o f such a system i s t h a t no plutonium-containing f r e s h f u e l c i r c u l a t e s outside t h e energy center.

The plutonium contained i n the spent f u e l i s returned t o the center f o r

u l t i m a t e destruction. oRI?L-Iwc 77-1041

,y38 ~KEUP

DENATUREDFUELASSEMBLIES (NO Pu)

ENERGY

I

(U-233, U-238, T~-232) OXIDE

/CENTER

[

BOUNDARY

DISPERSED REACTORS

-1

-

LWR

IRRADIATED FUEL

i'

,SPENT FUEL

u

t~

FABRICATION

rn Pu TO U-233 PU

- TH

TRANSMUTER

I

IRRADIATED FUEL

i Lt

L

TERMI NAL STORAGE

Schematic Fuel Flow f o r Symbiotic System Consisting o f an Fig. 2.2-1. Energy Center and Dispersed Reactors Operating on Denatured 233U Fuel. One obvious concern regarding such a coupled system i s the amount o f power produced by the dispersed systems r e l a t i v e t o t h a t produced i n the energy center reactors. The power ratio,*

defined a5 dispersed power generated r e l a t i v e t o c e n t r a l i z e d power, can be

viewed as a parameter c h a r a c t e r i z i n g the p r a c t i c a l i t y o f the system.

r-

L -t

b

L

While t h e power

r a t i o depends on t h e c h a r a c t e r i s t i c s o f the reactors a c t u a l l y u t i l i z e d f o r t h e various components and i s considered i n d e t a i l l a t e r i n t h i s report, c e r t a i n generic statements can be made.

.

I n a mature "safeguarded" plutonium cycle, the r a t i o would be zero since

a l l r e a c t o r s would, of necessity, be located i n energy centers.

I n t h e c u r r e n t open-ended

LEU cycles, t h i s r a t i o i s e s s e n t i a l l y i n f i n i t e since c u r r e n t nuclear generating capacity

i s dispersed v i a " n a t u r a l l y denatured" thermal systems. L

*

A1 so c a l l e d ''energy support r a t i o . "

The denatured 233U c y c l e w i l l f a l l


2-8

between these two extremes, and thus the proposed system's power r a t i o w i l l be a crucial evaluation parameter. The symbiotic system depicted by Fig. 2.2-1 can also be characterized by the type of reactors utilized inside and outside the center. In general, systems consisting of thermal (converter) reactors only, systems consisting of both thermal converters and f a s t breeder reactors, and systems consisting solely of f a s t breeder reactors can be envisioned.* One important characteristic of each system i s the extent t o which i t must rely on an external fuel supply t o meet the demand f o r nuclear-based generating capacity. The thermal-thermal system would be the most resource-dependent. The breeder-thermal system could be fuel-self-sufficient f o r a given power level and possibly also provide f o r moderate nuclear capacity growth. The breeder-breeder scenario, i f economically competitive w i t h a1 ternative energy sources, would permit the maximum resource-independent nuclear contribution t o energy production. While such considerations serve t o categorize the symbiotic systems themselves, the transition from the current once-through LEU cycles to the symbiotic systems i s of more immediate concern. A1 though a1 l-breeder systems would be resource-independent, commercial deployment of such systems i s uncertain. The transition t o the denatured cycle could be i n i t i a t e d relatively soon, however, by u s i n g moderately enriched 235U/2 38U mixed w i t h thorium (sometimes referred t o as the "denatured 23sU fuel cycle") i n existing and projected thermal systems. The addition of thorium (and the corresponding reduction of 238U over the LEU cycle) would serve a dual purpose: the quantity of plutonium generated would be significantly reduced, and an i n i t i a l stockpile of 2 3 % would be produced. I t should be noted t h a t this rationale holds even i f commercial fuel reprocessing i s deferred f o r some time. Use of denatured 23% fuel would reduce the amount of plutonium contained in the stored spent fuel. In addition, the spent fuel would represent a readily accessible source of denatured 233U should the need t o s h i f t from 23% arise. However, s u b s t i t u t i n g 232Th f o r some of the 23*U i n the LEU cycle would require higher f i s s i l e loadings and thus more 235U would be comnitted i n a shorter time frame than would be necessary w i t h the LEU cycle. An alternative would be t o u t i l i z e energy-center Pu-burning transmuters t o provide the i n i t i a l source of 233U f o r dispersed 233U-based reactors. From these s t a r t i n g points, various scenarios which employ thermal o r f a s t energy-center reactors coupled w i t h denatured thermal o r f a s t dispersed reactors can be developed. On the basis of the above, eight general scenarios have been postulated f o r this study, with two s e t s o f constraints on Pu utilization considered: e i t h e r plutonium wiZZ no& be a l lowed as a recycle fuel b u t recycle of denatured 233U will be permitted; or plutonium wiZZ be allowed within secure energy centers w i t h only denatured fuels being acceptable f o r use a t dispersed s i t e reactors. The eight scenarios can be summarized as follows:

*

See Section 4.0 for discussion of reactor terminology as applied i n this study.


2-9

1.

Nuclear power i s l i m i t e d t o low-enriched uranium-fueled (LEU) thermal reactors operati n g on a stowaway c y c l e (included t o allow comparisons w i t h c u r r e n t p o l i c y ) .

2.

LEU reactors w i t h uranium recycle are operated outside secure energy centers and thermal reactors w i t h plutonium r e c y c l e are operated i n s i d e the centers.

3.

Same as Scenario 2 plus f a s t breeder reactors (FBRs) operating on the Pu/U c y c l e are deployed w i t h i n t h e centers.

4.

LEU reactors and denatured 235U and denatured 233U reactors are operated w i t h uranium recycle, a l l i n dispersed areas; no plutonium r e c y c l e i s permitted.

5.

Same as Scenario 4 plus thermal reactors operating on t h e Pu/Th cycle are permitted w i t h i n secure energy centers,

6. - Same as Scenario 5 plus FBRs w i t h Pu/U cores and thorium blankets ( " l i g h t " transmutat i o n r e a c t o r s ) are permitted w i t h i n secure energy centers.

7.

Same as Scenario 6 plus denatured FBRs w i t h 233U/238U cores and thorium blankets are permitted i n dispersed areas.

8.

The " l i g h t " transmutation FBRs o f Scenario 7 are replaced w i t h "heavy" transmutation reactors with Pu/Th cores and thorium blankets.

2.3.

SOME INSTITUTIONAL CONSIDERATIONS OF THE DENATURED FUEL CYCLE

As s t a t e d above, the implementation of the denatured f u e l c y c l e w i l l e n t a i l the c r e a t i o n o f f u e l cycle/energy centers, which w i l l r e q u i r e i n s t i t u t i o n a l arrangements t o manage and c o n t r o l such f a c i l i t i e s .

The advantages and disadvantages o f such centers,

whether they be regional, m u l t i n a t i o n a l , o r i n t e r n a t i o n a l , as w e l l as t h e mechanisms r e q u i r e d f o r t h e i r implementation, have been rep0rted.3'~ Although a d e t a i l e d enumeration o f t h e conclusions o f such studies are beyond the scope o f t h i s p a r t i c u l a r discussion, c e r t a i n aspects o f the energy center concept as i t r e l a t e s t o the denatured f u e l c y c l e are relevant. Since o n l y a few thousand kilograms o f 233U c u r r e n t l y e x i s t , i t i s c l e a r t h a t production o f 2% w i l l be required p r i o r t o f u l l - s c a l e deployment o f t h e denatured 233U cycle. I f the reserves o f economically recoverable n a t u r a l uranium are allowed t o become extremely l i m i t e d before t h e denatured c y c l e i s implemented, most ifn o t a l l power produced a t t h a t time would be from energy-center transmuters. Such a s i t u a t i o n i s c l e a r l y i n c o n s i s t e n t w i t h t h e p r i n c i p l e t h a t the number o f such centers and the percentage o f t o t a l power produced i n them be minimized. A gradual t r a n s i t i o n i n which 235U-ba~ed dispersed reactors a r e replaced w i t h denatured 2 3 %-based dispersed reactors and t h e i r accompanying energy-center transmuter systems i s thus desirable. The proposed denatured fuel cycle/energy center scenario a l s o presents an a d d i t i o n a l dimension i n t h e formulation o f t h e energy p o l i c i e s o f n a t i o n a l s t a t e s

-

interdependence.

- t h a t o f nuclear

By t h e very nature o f t h e proposed symbiotic r e l a t i o n s h i p inherent i n


2-1 0

the denatured cycle, a condition of mutual dependence between the dispersed reactors and the energy-center reactors i s created. Thus while nations choosing to operate only denatured ( i .e. , dispersed) reactors must obtain t h e i r fuel from nations t h a t have energy-center transmuters, the nations operating the transmuters will i n t u r n rely on the nations operating dispersed reactors f o r t h e i r tvansmuter fuel requirements ( P u ) . Hence, i n addition t o the possible nonproliferation advantages of the denatured fuel cycle, the concept also introduces a greater f l e x i b i l i t y i n national energy policies. ,

References f o r Chapter 2 1.

H. A. Feiveson and T. B. Taylor, "Security Implications of Alternative Fission Futures," BUZZ. Atomic S c i e n t h t s , p. 14 (Dec. 1976).

2.

"A Report on the International Control of Atomic Energy," prepared f o r the Secretary of S t a t e ' s Committee on Atomic Energy by a Board of Consultants: Chester I. Barnard, Dr. J. R. Oppenheimer, Dr. Charles A. Thomas, Harry A. Winne, and David E. Lilienthal (Chairman), Washington, D.C., March 16, 1946, pp. 127-213, Department of State Publication 2493.

3.

"Nuclear Energy Center S i t e Survey Commi s s i on , January , 1 976.

4.

"Regional Nuclear Fuel Cycle Centers," 1977 Report of the IAEA Study Project, STI/TUB-445, International Atomic Energy Agency, 1977.

-

1975," Volumes 1-5, NUREG-0001, Nuclear Regulatory


e

Isr CHAPTER 3 ISOTOPIC CHARACTERISTICS OF DENATURED 233U FUEL

Chapter Out1 i n e

i

3.0.

I n t r o d u c t i o n , T. J . Burns and L .

3.1.

Estimated 232U Concentrations i n Denatured 233U Fuels, D . T. I n g e r s o l l , ORNL

3.2.

Radiological Hazards o f Denatured Fuel Isotopes, H . R . Meyer and J . E . Till, ORNL

3.3.

I s o t o p i c s Impacting Fuel Safeguards Considerations

S . Abbott, ORNL

L

3.3.1. 3.3.2.

Enrichment C r i t e r i a o f Denatured Fuel, c. M . Newstead, BNL F a b r i c a t i o n and Handling o f Denatured Fuel, J . D . Jenkins and

3.3.3. 3.3.4.

Detection and Assay o f Denatured Fuel , D . T. I n g e r s o l l , ORNL P o t e n t i a l Circumvention o f the I s o t o p i c B a r r i e r o f Denatured Fuel, E . H. G i f t

i-

3.3.5.

Deterrence Value o f 232U Contamination i n Denatured Fuel, c. M . Newstead, ORNL

b

t

R . E . Brooksbank, ORNL

and W . B . Arthur, ORGDP


3.0.

INTRODUCTION

T. J. Burns and L. S. Abbott Oak Ridge National Laboratory An assessment of t h e denatured 233U f u e l c y c l e

- both f o r meeting t h e

requirements

f o r e l e c t r i c a l power growth and f o r reducing the r i s k s of nuclear weapons p r o l i f e r a t i o n i n v a r i a b l y must i n c l u d e an examination o f t h e i s o t o p i c s o f the cycle. I t has been

-

p o i n t e d o u t i n Chapters 1 and 2 t h a t t h e concept of t h e denatured 233U c y c l e i s an attempt t o r e t a i n t h e i s o t o p i c b a r r i e r i n h e r e n t i n the c u r r e n t l y used LWR low-enriched 235U (LEU) c y c l e b u t a t t h e same time t o a l l o w the production and r e c y c l i n g o f new f u e l .

I n both t h e

denatured and t h e LEU cycles t h e i s o t o p i c b a r r i e r i s created by d i l u t i n g t h e f i s s i l e isotope w i t h 238U, so t h a t t h e concentration of t h e f i s s i l e n u c l i d e i n any uranium chemicall y e x t r a c t e d from f r e s h f u e l would be s u f f i c i e n t l y low t h a t the m a t e r i a l would n o t be

d i r e c t l y usable f o r weapons purposes.

This i s i n c o n t r a s t t o t h e two reference f u e l cycles,

t h e Pu/U cycle, and t h e HEU/Th cycle.

I n both o f these cycles, weapons-usable m a t e r i a l

could be e x t r a c t e d from the f r e s h f u e l v i a chemical separation. Table 3.0-1,

O f course, as shown i n

chemically e x t r a c t a b l e f i s s i l e m a t e r i a l i s present i n the spent f u e l elements

of a l l these cycles; however, t h e spent elements are n o t considered t o be p a r t i c u l a r l y vulnerable because o f the high r a d i o a c t i v i t y e m i t t e d by the f i s s i o n products initially.

-

a t least

I n t h i s assessment o f denatured 233U f u e l , t h e i m p l i c a t i o n s o f s u b s t i t u t i n g t h e denatured f u e l f o r t h e reference cycles o f various r e a c t o r s are examined. I n a d d i t i o n t o t h e obvious advantage o f t h e i s o t o p i c b a r r i e r i n t h e f r e s h f u e l , denatured 233U f u e l has an a d d i t i o n a l p r o t e c t i o n f a c t o r against d i v e r s i o n i n t h a t i t s f r e s h f u e l i s r a d i o a c t i v e t o a much g r e a t e r e x t e n t than any of t h e o t h e r f u e l s l i s t e d i n Table 3.0-1. This c h a r a c t e r i s t i c i s due t o t h e presence o f the contaminant 232U, which i s generated as a byproduct o f t h e 233U production process and which spawns a h i g h l y r a d i o a c t i v e decay chain. As shown i n Fig. 3.0-1, 232U decays through 22eTh t o s t a b l e 208Pb, e m i t t i n g numerous g a m rays i n the process, t h e most prominent being a 2.6-MeV gama r a y associated w i t h t h e decay o f 208T1. Table 3.0-1.

Comparison of P r i n c i p a l F i s s i l e and F e r t i l e Nuclides i n Some Reactor Fuels Fuel

Fresh Fuel Nuclides

Spent Fuel Nuclides

Denatured 233U f u e l (with recycle) LEU (no recycle)

23311,

2 3 8 ~ 2 3~2 ~ h

233U, Puf, 238U, 232Th

235u,

2 3 8 ~

235u, Puf,

23%

LEU ( w i t h r e c y c l e )

235u,

Puf,

23511, Puf,

2380

Pu/U (with recycle) HEU/Th (no recycle)

Puf, 2381) 2 3 5 ~ 2 3~2 ~ h

238u

Puf, 23% 2 3 3 ~2 ~3 5 ~2 3~2 ~ h


3-4

The r a d i o a c t i v i t y associated w i t h the 232"

ORNL-DWG 65-55DR3

233U s i g n i f i c a n t l y impacts t h e associated f u e l

cycle.

The f a b r i c a t i o n , shipping, and handling

o f t h e f r e s h denatured f u e l i s expected t o d i f f e r markedly from the o t h e r cycles , p r i m a r i l y due t o the f a c t t h a t remote procedures w i l l have t o be employed throughout.

To design t h e

necessary f a c i l i t i e s w i l l r e q u i r e a knowledge o f the concentrations o f 232U (and i t s daughter products) i n t h e f u e l as a f u n c t i o n o f time. To date, i n s u f f i c i e n t data a r e a v a i l a b l e on t h i s subject, b u t on t h e basis o f some prel i m i n a r y i n v e s t i g a t i o n s some estimates are given i n Section 3.1 on the 232U concentrations t h a t could be expected i n t h e recycled f u e l o f LWRs , HTGRs ,and FBRs operating on denatured 233u.

The r a d i o l o g i c a l hazards associated wfth the use o f denatured 2 3 3 U f u e l represent another aspect o f t h e c y c l e demanding a t t e n t i o n .

Again

l i t t l e i n f o r m a t i o n i s a v a i l a b l e , b u t Section 3.2 discusses the t o x i c i t y o f t h e various isotopes present i n t h e f u e l and a l s o i n thorium ore, Fig. 3.0-1.

Decay o f 232U.

as w e l l as t h e e f f e c t s o f exposure t o t h e gamma rays emitted from the f r e s h f u e l .

I n assessing t h e safeguard features o f denatured 233U f u e l , t h e i s o t o p i c s o f the c y c l e must be examined from several viewpoints.

While t h e 232U contamination w i l l be e s s e n t i a l l y

an i n h e r e n t property o f t h e 4denatured f u e l cycle, t h e concentration of the i s o t o p i c denaturant, 238U, i s c o n t r o l l a b l e . The presente o f both isotopes a f f e c t s t h e p r o l i f e r a t i o n p o t e n t i a l o f t h e denatured f u e l cycle. As t h e 238U concentration i s increased, the d i f f i c u l t y o f circumventing t h e i n t r i n s i c i s o t o p i c b a r r i e r i s increased.

However, i n c r e a s i n g t h e 238U f r a c t i o n

a l s o increases t h e 239Pu concentration i n the spent f u e l so t h a t an obvious t r a d e - o f f o f p r o l i f e r a t i o n concerns e x i s t s between t h e f r o n t and back ends o f t h e denatured f u e l cycle. As pointed o u t i n Section 3.3.1,

t h e enrichment c r i t e r i a f o r denatured

233U

fuel are s t i l l

being formulated. The requirement f o r remote operations throughout the f u e l c y c l e w i l l i n i t s e l f c o n s t i t u t e a safeguard f e a t u r e i n t h a t access t o f i s s i l e m a t e r i a l w i l l be d i f f i c u l t a t a l l stages o f t h e cycle. But t h i s requirement w i l l a l s o be a complicating f a c t o r i n t h e design o f the f u e l r e c y c l i n g steps and operations.

This s u b j e c t i s t r e a t e d i n more d e t a i l i n

Chapter 5, b u t Section 3.3.2 of t h i s chapter p o i n t s o u t t h a t t h e remote operation requirement could d i c t a t e t h e s e l e c t i o n o f techniques, as, f o r example, f o r t h e f u e l f a b r i c a t i o n process.


3- 5

4 -.

The radioactivity of the 232U chain would also make i t easier t o detect diverted denatured fuel and would complicate both the production of weapons-grade 233U from fresh denatured fuel and its subsequent use i n an explosive device. On the other hand, as discussed i n Section 3.3.3, the fadioactivity will i n h i b i t passive, nondestructive assays for f i s s i l e accountability. Finally, the possible circumvention of the isotopic barrier must be addressed. In Section 3.3.4 i t is postulated that a gas centrifuge isotope separation f a c i l i t y is available f o r isotopically enriching diverted fresh denatured 233U fuel, and estimates are made of the amounts of weapons-grade material that could be so obtained. Conclusions are then drawn as t o the r e l a t i v e attractiveness of denatured 233U fuel and other fuels t o would-be d i verters

.

1

Y

c

e


3-6

3.1.

ESTIMATED

U CONCENTRATIONS I N DENATURED

U FUELS

D. T. I n g e r s o l l Oak Ridge National Laboratory Although i t i s mandatory t h a t ’ t h e concentrations o f 2 3 2 U a t each stage o f the f u e l c y c l e be p r e d i c t a b l e f o r the various reactors operating on thorium-based f u e l s , l i t t l e information on t h e s u b j e c t i s a v a i l a b l e a t t h i s time.

This i s a t t r i b u t a b l e t o the f a c t

t h a t the i n t e r e s t i n thorium f u e l cycles i s r e l a t i v e l y recent and t h e r e f o r e the nuclear data required f o r c a l c u l a t i n g the production o f 2 3 2 U have n o t been adequately developed. O f primary importance are the (n,y) cross sections o f 23’Pa, 230Th, and 232Th and the (n,Zn) cross sections o f 2 3 3 U and 232Th, a l l o f which are intermediate i n t e r a c t i o n s t h a t can l e a d t o t h e formation of 2 3 2 U as i s i l l u s t r a t e d by t h e r e a c t i o n chain given i n Fig. These cross sections are under c u r r e n t evaluation’ and should appear i n the Version 3.1-1. V release o f t h e Evaluated Nuclear Data F i l e (ENDF/B-V).

I n s p i t e o f t h e nuclear data d e f i c i e n cies, some r e s u l t s f o r 2321) concentrations are a v a i l a b l e from c a l c u l a t i o n s f o r denatured fuels i n l i g h t - w a t e r r e a c t o r s (LWRs) and i n f a s t breeder reactors (FBRs) . A1 though no

I 232

r e s u l t s f o r denatured high-temperature gas-

9oTh

I

I

cooled reactors (HTGRs) are c u r r e n t l y a v a i l a b l e , 232U concentrations can be roughly i n f e r r e d

h2n)

from e x i s t i n g HTGR f u e l data.

Moreover, t h e

analysis o f 232U concentrations i n standard HTGR designs (HEU/Th) serves as an upper

t

A compilat i o n of the a v a i l a b l e r e s u l t s i s given below.

bound f o r t h e denatured systems.

E l h

The c u r r e n t s t a t e of the r e l a t e d 232U nuclear data i s amply r e f l e c t e d i n t h e l a r g e variances

Fig. 3.1-1 Important Reaction Chains Leading t o t h e Production o f 232U.

3.1.1.

of the c a l c u l a t e d concentrations.

Light-water Reactor Fuels

E x i s t i n g data on 232U concentrations i n denatured LWR f u e l s are p r i m a r i l y from c a l c u l a t i o n s based on t h e Combustion Engineering System 80TM r e a c t o r design.2 Results from CE3 f o r a denatured 235U c y c l e (20% 235U-enriched uranium i n 78% thorium) show t h e 232U

concentration a f t e r t h e zeroth generation t o be 146 ppm 232U i n uranium, w h i l e a f t e r f i v e generations o f r e c y c l e uranium, t h e concentration i s increased t o 251 ppm. These l e v e l s are i n good agreement w i t h ORNL calculation^,^ which i n d i c a t e 130 ppm 232U i n uranium f o r t h e z e r o t h generation. The discharge uranium i s o t o p i c s are summarized i n Table 3.1-1.

Also shown are t h e r e s u l t s from an ORNL c a l c u l a t i o n f o r a denatured 233U c y c l e

t ’

L

L L;

h


3-7

(10% 233U-enriched uranium i n 78% Th).

The s l i g h t c o n t r i b u t i o n from "'U

reactions i n -

creases the 2 3 2 U content t o 157 ppm a f t e r the zeroth generation. Table 3.1-1.

Discharge I s o t o p i c s f o r LWRS Operating on Denatured Fuels

i,

Iso t o p i c Fraction Cyc1e

-

2 3 2 ~

2 3 3 ~

2 3 4 ~

2 3 5 ~

2 3 6 ~

in U (ppm)

232U 2 3 6 ~

232~h

235U/Th Fuel'

f

L

CE(0)b

0.0029

1.07

0.11

1.56

0.50

16.81

76.21

146

ORNL(0)

0.0026

1.00

0.09

1.59

0.49

16.85

76.23

130

CE(5)

0.0061

1.60

0.69

1.27

1.86

18.78

75.79

251

0.0031

1.16

0.29

0.056

0.0052

18.32

75.99

%/Th Fuel

ORNL (0) 'Initial ?he 0

isotopics:

157

,

4.4% 235U, 17.6% 238U, 78% 232Th.

number i n parentheses represents the f u e l generation number,

I n i t i a l isotopics:

2.8% 233U, 19.2% 238U, 78% 232Th.

I

ii

3.1.2.

t

U b T e m p e r a t u r e Gas-Cooled Reactor Fuels

Although c a l c u l a t i o n s f o r 232U concentrations i n denatured HTGR f u e l s are n o t a v a i l -

t

able, i t i s possible t o roughly i n f e r t h i s information from e x i s t i n g HTGR c a l c u l a t i o n s i f

.-

the expected changes i n t h e thorium content are known.

I

w i t h 93% 235U-enriched uranium fuel and thorium f e r t i l e material, On successive cycles, the 233U produced i n t h e thorium i s recycled, thus reducing t h e required amount o f 235U makeup. The 232U content o f the recycled f u e l becomes appreciable a f t e r o n l y a few genera-

u

tions.

Table 3.1-2

The conventional HTGR c y c l e begins

gives the uranium i s o t o p i c s of t h e recycle f u e l batches a t the beginning

o f r e c y c l e and a t e q u i l i b r i u m recycle,5 t h e l a t t e r showing a maximum 232U concentration o f 362 ppm i n uranium.

--

Table 3.1-2.

Uranium I s o t o p i c s f o r Commercial HTGR Recycled Fuel (HEU/Th)

1 i,

Beginning o f recycle Equilibrium recycle

23211

233u

0.000126 0.000362

Isotopic Fraction

in U (ppm)

232U

234u

23511

0.921

0.0735

0.00568

0.000245

126

0.614

0.243

0.0802

0.0630

362

23611


3-8

The values i n Table 3.1-2 are a r e s u l t o f a standard HTGR f u e l composition which has an average Th/233U r a t i o o f about 20.

Preliminary estimates have been made o f dena-

t u r e d HTGR f u e l s which assume a 20% denatured 235U, leading t o a 15% denatured 233U.6 Because o f t h e added 2381) f e r t i l e material, t h e amount o f thorium i s correspondingly r e duced by about 30%, r e s u l t i n g i n a s i m i l a r reduction i n t h e 232U production. The conc e n t r a t i o n o f 232U i n t o t a l uranium would a l s o be reduced by t h e mere presence of t h e d i l u t i n g 238U, so t h a t i t can be estimated t h a t a 15% denatured 233U HTGR would c o n t a i n approximately 40 ppm 232U i n uranium a f t e r e q u i l i b r i u m recycle. The lower 232U l e v e l s i n the HTGR are p r i m a r i l y due t o a s o f t e n i n g o f t h e neutron energy spectrum compared w i t h t h a t o f t h e LWR. This r e s u l t s i n a marked reduction i n t h e 232Th(n,2n') which i s a prime source o f 232U.

3.1.3.

reaction rate,

L L

Fast Breeder Reactor Fuels

232U concentrations c a l c u l a t e d by Mann and Schenter' and by Burns8 f o r various Except f o r Case 2, tnese conunercial-sized FBR f u e l cycles are given i n Table 3.1-3.

values were determined from r e a c t i on-rate c a l c u l a t i ons using 42 energy groups and onedimensional geometry; t h e Case 2 r e s u l t s were determined from a coarse nine-group twodimensional d e p l e t i o n calculation. It i s important t o note t h a t Cases 1 and 2 represent t h e "transmuter" concept.

All

the discharged uranium (23211, 233U, 2s4U, and 235U) i s bred from t h e 232Th i n i t i a l l y charged and c o n s i s t s p r i n c i p a l l y o f 2331). This accounts f o r t h e high 232U/U r a t i o , which w i l l be reduced by a f a c t o r o f 5 t o 8 i n t h e denatured f u e l manufactured from t h i s mate-

r i a l . Thus, denatured f u e l generated v i a t h e f a s t Pu/Th transmuter i s expected t o have approximately 150-750 ppm 23211 i n uranium.

L

FBR Core Region 232U Discharge Concentrations'

Table 3.1-3.

2321) i n

u

(PP~)

Case t = l y r b

t = 2 y r

t = 3 y r

t = 5 y r

982

1710

2380

3270

11% 239Pu i n Th

1106

2376

3670

3

10% 233U i n Th

288

830

1330

2210

4

10% 2331) i n

6.6

10.7

12.5

13.3

5

10% 233U i n Th

1820

2760

3260

6

10% 23311 i n

35

35

35

NO.

Fuel

1

10% 239Pu i n Th

2

II

No r e c y c l e

238U

Ll

With r e c y c l e

23%

I!

aCases 1, 3-6 are from r e f . 7; Case 2 i s from r e f . 8. bt = f u e l residence time f o r no r e c y c l e cases; t = burning time before r e c y c l e f o r r e c y c l e cases.


3-9

The l a s t two cases i n Table 3.1-3 g i v e t h e e q u i l i b r i u m 232U concentrations assumi n g r e c y c l e o f t h e 2 3 3 U and t h e associated 232U. I t should be noted t h a t these two cases For a 20% denatured f u e l represent t h e extremes regarding a1 lowable enrichment ( 2 3 3 U / U ) . i n which approximately h a l f t h e heavy metal i s 232Th, t h e expected 232U e q u i l i b r i u m conc e n t r a t i o n would be

4

1600 ppm (232U/U)

--

3.1.4.

f o r a 3-yr c y c l e residence time. Conclusions

hi The r e s u l t s presented i n t h i s section are, f o r t h e most part, p r e l i m i n a r y and/or approximate. T h i s i s l a r g e l y a consequence o f t h e u n c e r t a i n t i e s i n t h e a n t i c i p a t e d f u e l

i

compositions, denaturing l i m i t s , r e c y c l e modes, etc.,

as w e l l as t h e basic nuclear data.

Also, t h e r e s u l t s assumed zero o r near-zero 230Th concentrations, which can approach s i g n i f i c a n t l e v e l s depending cn t h e source o f t h e thorium stock, p a r t i c u l a r l y i n thermal sys-

-

t

tems. Because o f t h e r e l e v a n t cross sections, t h e presence o f even small amounts o f 230Th can r e s u l t i n considerably higher 232U concentrations. It i s p o s s i b l e t o conclude, how-

L

ever, t h a t 232U concentrations w i l l be highest f o r 23%-producing fuel recycle, and decrease w i t h f i s s i l e denaturing.

FBRs, increase w i t h

References f o r Section 3.1

L

*

L

e

1.

Summary Minutes o f t h e Cross Section Evaluation Working Group Meeting, May 25-26, 1977.

2.

N. L. Shapiro, J. R. Rec, R. A. Mattie, "Assessment of Thorium Fuel Cycles i n Pressurized-Water Reactors ,I' ERR1 NP-359, Combustion Engineering (1977).

3.

P r i v a t e comnunication from Combustion Engineering t o A. Frankel, Oak Ridge National Laboratory, 1977.

4.

P r i v a t e comnunication from W. August 11, 1977.

5.

J. E. Rushton J. D. Jenkins, and S. R. McNeany, "Nondestructive Assay Techniques f o r Recycled $3311 Fuel f o r High-Temperature Gas-Cooled Reactors," J. Institute Nuclear Materials Management I V (1 ) (1975).

6.

M. H. M e r r i l l and R. K. Lane, " A l t e r n a t e Fuel Cycles i n HTGRs," D r a f t f o r J o i n t Power Generation Conference, Long Beach, Calif., 1977.

7.

F. )rl. Mann and R. E. Schenter, "Production o f 232U i n a 1200 MWe LMFBR," Hanford Engineering Development Laboratory (June 10, 1977).

a.

T. J . Burns, p r i v a t e communication.

6. A r t h u r t o J. W. Parks, Oak Ridge National Laboratory,


3-1 0

3.2.

RADIOLOGICAL HAZARDS OF DENATURED FUEL ISOTOPES

H. R. Meyer and J. E. T i l l Oak Ridge National Laboratory Consideration o f the denatured 2 3 3 U c y c l e has created the need t o determine t h e r a d i o l o g i c a l hazards associated w i t h extensive use o f 2 3 3 U as a nuclear f u e l . These hazards w i l l be determined by the t o x i c i t y o f the various isotopes present i n t h e f u e l and i n thorium ore, which i n t u r n i s influenced by t h e path through which t h e isotopes enter the body--that i s , by i n h a l a t i o n o r ingestion.

I n a d d i t i o n , the gamma rays e m i t t e d

from the denatured f u e l present a p o t e n t i a l hazard. 3.2.1.

T o x i c i t y o f 233U and 2 3 2 U

Only l i m i t e d experimental data are a v a i l a b l e on t h e t o x i c i t y o f h i g h s p e c i f i c a c t i v i t y uranium isotopes such as 2 3 3 U and 232U.

Chemical t o x i c i t y , as opposed t o r a d i o l o g i c a l

hazard, i s t h e l i m i t i n g c r i t e r i o n f o r t h e l o n g - l i v e d isotopes o f uranium ( 2 3 J U and 2 3 8 U ) which are o f primary concern i n the l i g h t - w a t e r r e a c t o r uranium f u e l cycle.' I n order t o e s t a b l i s h t h e r e l a t i v e r a d i o t o x i c i t y o f denatured 2 3 3 U f u e l , i t i s h e l p f u l t o consider s p e c i f i c metabolic and dosimetric parameters o f uranium and plutonium isotopes. Table 3.2-1 l i s t s several important parameters used i n r a d i o l o g i c a l dose c a l c u l a t i o n s . The e f f e c t i v e h a l f l i f e f o r 239Pu i n bone i s approximately 240 times t h a t o f uranium.

How-

ever, t h e e f f e c t i v e energy p e r d i s i n t e g r a t i o n f o r 2 3 2 U i s about t h r e e times g r e a t e r than t h a t f o r any of t h e plutonium isotopes. I n general, the time-integrated dose from plutonium isotopes would be s i g n i f i c a n t l y greater than the dose from uranium isotopes f o r t h e i n h a l a t i o n pathway, assuming i n h a l a t i o n o f equal a c t i v i t i e s o f each radionuclide. Doses v i a t h e i n g e s t i o n pathway, again on a per V C i basis, are much lower than those e s t i mated f o r t h e i n h a l a t i o n pathway. It i s c u r r e n t l y assumed t h a t a l l bone-seeking radionuclides are f i v e times more

e f f e c t i v e i n inducing bone tumors than 226Ra. However, the l i m i t e d number o f studies t h a t have been conducted w i t h 2 3 3 U ( r e f . 2) and 2 3 2 U ( r e f s . 3-5) suggest a reduced e f f e c t i v e n e s s i n inducing bone tumors f o r these isotopes and may r e s u l t i n use o f exposure l i m i t s t h a t are l e s s r e s t r i c t i v e than c u r r e n t l i m i t s . The l a s t two columns i n Table 3.2-1 represent dose conversion f a c t o r s (DCFs) f o r uranium 'and plutonium isotopes c a l c u l a t e d on t h e basis o f mass r a t h e r than a c t i v i t y . may be seen t h a t t h e

It

232U "Mass DCFs" a r e more than f o u r orders o f magnitude g r e a t e r than

those f o r f i s s i o n a b l e 233U, due l a r g e l y t o , t h e high s p e c i f i c a c t i v i t y o f 232U. T h i s f a c t o r c o n t r i b u t e s t o t h e o v e r r i d i n g importance o f

232U content when considering t h e r a d i o t o x i c i t y

of denatured uranium f u e l s . Figure 3.2-1 i l l u s t r a t e s the importance o f 232U content w i t h respect t o p o t e n t i a l This f i g u r e presents the estimated dose commitment t o bone calcut o x i c i t y of 2 3 3 U f u e l .


3-1 1

hd Table 3.2-1.

Isotope

23 2u

a

Specific A c t i v i t y (Ci/g)

E f f e cint i v e H a l f Life (Days 1

3.00 x

21.42

23 3 u

9.48 x 10-3

3.00 x

235u

2.14 x

3.00 x

2 3 8 ~

3.33 x 10-7

3.00 x

238Pu

I,

Metabolic Data and Dose Conversion Factors (DCFS) f o r Bone f o r Selected Uranium and Plutonium Isotope

17.4

lo2 lo2 lo2 lo2

A c t i v i t - v Dose Conversion Factor ngestion (r e m s h C i ) (rems/pCi)

w loo

Mass Dose Conversion Factor f n h a l a t i o n c Ingestionc (rems/ug) ( rems /pg ) 2.4 x

lo3

1.1 x 102

4.1 x

2.2 x 10'

8.6 x 10''

2.1 x 10''

8.2 x 10-3

2.0 x 10'

8.0 x 10''

4.3 x

1.7 x 10'6

1.9 x 101

7.6 x 10''

6.3 x

2.5 x

6.8 x 10-1

9.9 x

1.2 x 101

4.0

4.8 x

2.3

x 104

5.7 x 103

23 9Pu

6.13 x

7.2

x 104

6.6 x

24OPu

2.27 x 10'1

7.1

x 104

6.6 x

lo3 lo3

7.9 x

lo-'

7.9 x 10''

1.5

lo4 x lo2 x lo3

8.8 x 10'

1.8 x 10''

I n t e r n a t i o n a l Commission on Radiological Protection, "Report o f Committee I 1 on Permissible Dose f o r I n t e r n a l Radiation,'' ICRP P u b l i c a t i o n 2, Pergamon Press, New York, 1959.

bKillough, 6. G., and L. R. McKay, "A Methodology f o r C a l c u l a t i n g Radiation Doses from R a d i o a c t i v i t y Released t o t h e Environment ,'I ORNL-4992, 1976.

-

C

Product o f s p e c i f i c a c t i v i t y and a c t i v i t y dose conversion f a c t o r .

1

Y

I: L

lated for inhalation o f

lo-''

g o f u n i r r a d i a t e d 2 3 3 U HTGR f u e l (@3% 233U/U) as a f u n c t i o n

o f the 2 3 2 U i m p u r i t y content f o r two d i f f e r e n t times f o l l o w i n g separation a t a reprocessing f a c i l i t y . The upper curve i s t h e dose commitment a t 10 years a f t e r separation. Two basic conclusions can be drawn from these data. F i r s t as recycle progresses and concentrations o f 232U become greater, the o v e r a l l r a d i o t o x i c i t y o f z 3 3 U f u e l w i l l increase s i g n i f i c a n t l y . Second, t h e ingrowth o f "*U daughters i n 2 3 3 U f u e l increases f u e l r a d i o t o x i c i t y s i g n i f i c a n t l y f o r a given concentration o f 232U. Although t h e data g r a p h i c a l l y i l l u s t r a t e d i n Fig. 3.2-1 were n o t s p e c i f i c a l l y c a l c u l a t e d f o r denatured 2 3 3 U f u e l , the required data n o t being available, the r e l a t i v e shape o f t h e curves would remain the same. A l l e l s e being equal, t h e estimated r a d i o t o x i c i t y o f denatured f u e l would be reduced due t o d i l u t i o n o f 2 3 3 U and 2 3 2 U w i t h 238U, which has a low r a d i o l o g i c a l hazard.

A comparison o f the dose commitment t o bone r e s u l t i n g from i n h a l a t i o n o f g o f three types o f f u e l , HTGR 2 3 3 U f u e l , LWR 235U f u e l , and FBR plutonium fuel, i s given i n Fig. 3.2-2. T h i s analysis evaluates u n i r r a d i a t e d HTGR f u e l containing 1000 ppm 232U and

I

z

does n o t consider f i s s i o n products, a c t i v a t i o n products, transplutonium radionuclides, o r environmental transport. As shown i n Table 3.2-1, the i n h a l a t i o n pathway would be by f a r t h e most s i g n i f i c a n t f o r environmentally dispersed fuels. pathways o f exposure are n o t considered i n t h i s b r i e f analysis.

lo-'

Therefore, other p o t e n t i a l


3-1 2

ORNL - D W G

232u IN Effect of Fig. 3.2-1. Commi tment t o Bone.

23%

75-3472R3

RECYCLED HTGR FUEL (ppm) Concentrations i n HTGR Fuel (93% 233U/U) on Dose


3-1 3

ORNL-DWG 75-2938R3

L

L i-

ki

TIME AFTER SEPARATION ( years) Fig. 3.2-2. R e l a t i v e R a d i o t o x i c i t y of FBR Plutonium Fuel, HTGR Fuel (93% 233U/U) and LWR Uranium Fuel as a Function o f t h e Time a f t e r Separation a t Reprocessing Plant.

I t i s noted t h a t Fig. 3.2-2 a p p l l e s t o f r e s h f u e l as a function o f time a f t e r separation, presuming i t has been released t o t h e environment. I n h a l a t i o n long a f t e r release could r e s u l t from t h e resuspension o f r a d i o a c t i v e m a t e r i a l s deposited on t e r r e s t r i a l surfaces. A dose commitment curve f o r denatured 2 3 3 U fuel would be expected t o l i e s l i g h t l y below t h e given curves f o r HTGR f u e l ; however, t h e denatured f u e l would remain s i g n i f i c a n t l y more hazardous

t bd

from a r a d i o l o g i c a l standpoint than

LWR

uranium f u e l .


3-14

3.2.2

T o x i c i t y o f 232Th

Given t h e p o t e n t i a l f o r r a d i o l o g i c a l hazard v i a t h e mining o f western U.S.

thorium

deposits as a r e s u l t o f implementation o f 232Th-based f u e l cycles, c u r r e n t d i f f i c u l t i e s i n e s t i m a t i o n o f 232Th DCFs must a l s o be considered here. As i s evident i n Fig. 3.0-1 (see Section 3.0),

both 2 3 2 U and 232Th decay t o 228Th, 232U decays t o 232Th

and then through t h e remainder o f t h e decay chain t o s t a b l e ‘OePb.

v i a a s i n g l e 5.3-MeV alpha emission; 232Thdecays v i a three steps, a 4.01-MeV alpha emission t o 228Ra, followed by s e r i a l beta decays t o 228Th. The t o t a l energy released i n t h e convergent decay chains i s obviously n e a r l y equal.

I;

The ICRP’ l i s t s e f f e c t i v e energies ( t o bone, per d i s i n t e g r a t i o n ) as 270 MeV f o r 232Th and 1200 MeV f o r 232U; these e f f e c t i v e energies are c r i t i c a l i n the determination of dose conversion factors t o be used i n e s t i m a t i o n o f long-term dose commitments. The l a r g e d i f f e r e n c e between t h e e f f e c t i v e energies c a l c u l a t e d f o r the two radionuclides i s based on t h e ICRP assumption ( r e f . 7) t h a t radium atoms produced by decay i n bone o f a thorium parent should be assumed t o be released from bone t o blood, and then r e d i s t r i b u t e d as though t h e radium were i n j e c t e d intravenously.

I;

As a r e s u l t , t h e presence o f 228Ra i n

t h e 232Th decay chain implies, under t h i s ICRP assumption, t h a t 90% o f the 228Ra created w i t h i n bone i s e l i m i n a t e d from the body.

Therefore, most o f the p o t e n t i a l dose from the

remaining chain alpha decay events i s n o t accrued w i t h i n t h e body, and t h e t o t a l e f f e c t i v e energy f o r t h e 232Th chain i s a f a c t o r o f 4.4 lower than t h a t f o r 232U, as noted. Continuation and reevaluation o f t h e e a r l y research”

l e a d i n g t o the above d i s -

s i m i l a r i t y i n d i c a t e d t h a t t h e presumption o f a major t r a n s l o c a t i o n o f 228Ra o u t o f bone was suspect ( r e f s . 10-14)

, and

o f 97% o f 228Ra i n bone.

t h a t s u f f i c i e n t evidence e x i s t e d t o s u b s t a n t i a t e r e t e n t i o n

Recalculation o f e f f e c t i v e energies f o r t h e 232Th chain on t h i s

basis r e s u l t s i n a value of 1681 MeV as l i s t e d i n ERDA 1451 ( r e f . 15), a s u b s t a n t i a l increase i m p l y i n g t h e need f o r more r e s t r i c t i v e l i m i t s w i t h respect t o 232Th exposures.

B 1: L

I n con-

t r a s t t o t h i s argument, t h e 1972 r e p o r t o f an ICRP Task Group o f Committee 2 ( r e f . 1 6 ) presents a newly developed whole-body r e t e n t i o n f u n c t i o n f o r elements i n c l u d i n g radium which e f f e c t i v e l y relaxes 232Th exposure l i m i t s .

3.2.3

Hazards Related t o Gamma-Ray Emissions

I;

While f u e l f a b r i c a t e d from f r e s h l y separated 2 3 3 U emits no s i g n i f i c a n t gamma r a d i a t i o n , ingrowth o f

232U daughters leads t o buildup o f 208T1 2.6-MeV gamma r a d i a t i o n , as

w e l l as o t h e r gama and x-ray emissions.

As discussed elsewhere i n t h i s report, i t i s

a n t i c i p a t e d t h a t occupational gamma exposures d u r i n g f u e l f a b r i c a t i o n can be minimized by such techniques as remote handling and increased shielding.

I‘


3-1 5

L

Gama exposure r e s u l t i n g from t h e t r a n s p o r t a t i o n o f i r r a d i a t e d f u e l elements cont a i n i n g 232U w i l l n o t be s i g n i f i c a n t l y d i f f e r e n t from t h a t due t o other f u e l s . Shielded Fasks would be used i n shipment t o c o n t r o l exposures t o t h e p u b l i c along t r a n s p o r t a t i o n routes.

F

L

Gama exposure from

232U daughters would be i n s i g n i f i c a n t compared t o exposure

from f i s s i o n products i n t h e spent f u e l . Refabricated f u e l assemblies containing 232U would r e q u i r e greater r a d i a t i o n s h i e l d i n g than LWR f u e l . However, t h i s problem can be minimized by shipping f r e s h assemb l i e s i n a c o n t a i n e r s i m i l a r i n design t o a spent f u e l cask. Gamma doses t o workers and t o t h e general p u b l i c due t o t r a n s p o r t o f f u e l m a t e r i a l s between f a c i l i t i e s are t h e r e f o r e expected t o be e a s i l y c o n t r o l l e d , and have been estimated t o be low, perhaps one man-rem p e r 1000 m ( e ) reactor-plant-year. l 5 The estimated gama hazard o f environmentally dispersed 232U, w h i l e a s i g n i f i c a n t c o n t r i b u t o r t o e x t e r n a l l y derived doses, i s overshadowed as a hazard by the e f f i c i e n c i e s o f i n t e r n a l l y deposited alpha e m i t t e r s i n d e l i v e r i n g r a d i o l o g i c a l doses t o s e n s i t i v e

id

tissues

. 3.2.4.

Conclusions

Several conclusions can be made from t h i s assessment.

I t appears t h a t a d d i t i o n a l metabolic and t o x i c o l o g i c a l data, both human and animal-derived, focusing on h i g h s p e c i f i c

a c t i v i t y uranium, would be h e l p f u l i n assessing t h e r a d i o l o g i c a l hazards associated w i t h

[I

denatured 233U f u e l .

S p e c i f i c a l l y , data on t h e b i o l o g i c a l e f f e c t i v e n e s s o f 232U and 233U

could modify exposure standards f o r these radionuclides.

In terms o f r e l a t i v e t o x i c i t i e s based on t h e dose c o m i t m e n t r e s u l t i n g from inhalat i o n o f equal masses o f f u e l , plutonium f u e l i s s i g n i f i c a n t l y more hazardous than HTGR 2 3 3 U f u e l o r denatured

233U fuel.

However, denatured 2 3 3 U f u e l would be s i g n i f i c a n t l y more

hazardous than LWR uranium fuel. As t h e range o f f u e l c y c l e options i s narrowed, more comprehensive research should be d i r e c t e d a t d e r i v a t i o n o f t o x i c i t y data s p e c i f i c t o f a c i l i t i e s and f u e l compositions

of choice.

Research i n v e s t i g a t i n g p o t e n t i a l environmental hazards r e s u l t i n g from d e l i b e r a t e

7-

t

i n t r o d u c t i o n ( f o r safeguards purposes) o f gama e m i t t e r s i n t o f u e l s p r i o r t o r e f a b r i c a t i o n i s necessary, as i s a thorough i n v e s t i g a t i o n of t h e hazards r e l a t e d t o repeated i r r a d i a t i o n o f r e c y c l e materials, w i t h consequent b u i l d u p of low cross-section transmutation products.


3-16

References f o r Section 3.2

1.

M. R. Ford, "Comnents on Intake Guides f o r Various Isotopes and I s o t o p i c Mixtures o f Uranium," ORNL 3697, 1964.

2.

H. C. Hodge, J. N. Stannard, and J. B. Hursh, uranium, Plutonium, and the Transptutoniwn EZements, S p r i nger-Verl ag , Hei del berg, 1973.

3.

M. P. Finkel , "Relative B i o l o g i c a l Effectiveness o f Radium and Other Alpha Emitters i n CF No. 1 Female Mice," Proceedings of the Society for Exp. BioZ. and Med. No. 3, p. 83, J u l y 1953.

4. J .

E. B a l l o u and R. A. Gies, "Early d l s p o s i t i o n of inhaled uranyl n i t r a t e (232U and in rats," p. 91 i n BNllL-2100 (Part 1 ) : Annual Report f o r 1976.

23311)

5.

J. E. B a l l o u and N. A. Wogman, "Nondestructive Analysis f o r 232U and Decay Progeny i n Animal Tissues," BNWL-2100 ( P a r t 1 ) : Annual Report f o r 1976.

6.

J. E. T i l l , "Assessment of t h e Radiological Impact of 232U and Daughters i n Recycled 233U HTGR Fuel ,I1 ORNL-TM-5049, February, 1976.

7.

I n t e r n a t i o n a l Commission on Radiological Protection, "Report o f Comnittee I 1 on Permissible Dose f o r I n t e r n a l Radiation," ICRP Pub1i c a t i o n 2, Pergamon Press, New York, 1959.

8.

J. C. Reynolds, P. F. G h t a f s o n , and L. D. N a r i n e l l i , "Retention and E l i m i n a t i o n of Radium Isotopes Produced by Decay of Thorium Parents w i t h i n t h e Body," USAEC Report ANL-5689, November 1957, p. 4.

9.

M. A. Van D i l l a and B. J. Stover, "On t h e Role o f Radiothorium (Th228) i n Radium Poisoning,'I RadiobioZogy 66: 400-401 , 1956.

10.

B. J. Stover, D. R. Atherton, D. 5 , Buster, and N. K e l l e r , "The Th228 Decay Series i n A d u l t Beagles: Ra224, Pb212 and B i 2 1 2 i n Blood and Excreta," Radiation Research 26: 226-243, 1965.

11.

C. W. Mays l e t t e r t o G. C. B u t l e r , 25 August 1967.

12.

A. Kaul, "Tissue D i s t r i b u t i o n and Steady State A c t i v i t y Ratios o f Th232 and Daughters i n Man Following I n t r a v a s c u l a r I n j e c t i o n o f Thorotrast," i n : Proceedings of the Third InternationaZ Meeting on the Toxicity of Thorotrast, Copenhagen, A p r i l 1973, M. Faber, ed., RISO Report No. 294.

13.

S. R. Bernard and W. S. Snyder, p r i v a t e communication e n t i t l e d "Memorandum on Thorium Daughters," A p r i l 1968.

14.

B. J. Stover, D. R. Atherton, D. S . Buster, and F. W. Bruenger, "The Thorium Decay Series i n A d u l t Beagles: Ra224, Pb212, and B i 2 1 2 i n Selected Bones and S o f t Tissues," , Radiation Research 26: 132-145, 1965.

15.

U.S. Energy Research and Development Administration, F i m Z EnvironmentaZ Statement, Light Water Breeder Reactor Progrm, ERDA-1541 , Vol s 1-5, 1976.

16.

I n t e r n a t i o n a l Council on Radiological P r o t e c t i o n Report 20, 1972.

.


3-1 7

3.3.

L.

ISOTOPICS IMPACTING FUEL SAFEGUARDS CONSIDERATIONS

3.3.1.

Enrichment C r i t e r i a o f Denatured Fuel C. M. Newstead Brookhaven National Laboratory

A very important problem i n t h e determination o f t h e c h a r a c t e r i s t i c s o f denatured f u e l i s t h e i s o t o p i c composition o f t h e uranium, t h a t i s t o say, t h e percent o f 233U

present i n t h e m i x t u r e o f 233U plus 238U.

The guidelines provided by c u r r e n t r e g u l a t i o n s

concerning t h e d i s t i n c t i o n between low-enriched uranium (LEU) and high-enriched uranium (HEU) are a p p l i c a b l e t o 235U, t h e l i m i t being s e t a t 20% 235U i n 238U. Anything above *-

-

1

t h a t c o n s t i t u t e s HEU and anything below t h a t c o n s t i t u t e s LEU. LEU i s considered t o be unsuitable f o r c o n s t r u c t i n g a nuclear explosive device.

-

The r a t i o n a l e f o r making t h i s statement i s based upon t h e f a c t t h a t t h e c r i t i c a l mass o f

ibi

20% 235U-enriched uranium i s 850 kg, and i n a weapon t h i s amount o f m a t e r i a l must be brought together s u f f i c i e n t l y r a p i d l y t o achieve an explosive e f f e c t . enrichment could be lower and s t i l l achieve prompt c r i t i c a l i t y .

Theoretically the

However, t h e amount o f

L

m a t e r i a l becomes so enormous and the d i f f i c u l t y o f b r i n g i n g i t together so great t h a t i t

'-

would be i m p r a c t i c a l t o attempt t o produce an explosive device w i t h l e s s than 20% enrich-

i

bi

ment.

It i s c l e a r t h a t t h e d i s t i n c t i o n i s somewhat o f a gray area and t h e enrichment

could be changed a few percent, b u t t h i s should be done extremely c a u t i o u s l y since t h e

i -

i

235U enrichment vs. c r i t i c a l mass curve i s r a t h e r steep and increasing t h e enrichment o n l y s l i g h t l y could reduce the c r i t i c a l mass s u b s t a n t i a l l y . consider i n s t i t u t i o n a l arrangements.

Also, i t i s necessary t o

A number o f domestic and i n t e r n a t i o n a l r e g u l a t i o n s

r e v o l v e about t h e 2 0 % ' f i g u r e and i t would be no easy m a t t e r t o change a l l these s t i p u l a t i o n s . This sets t h e background against which t h e enrichment considerations f o r denatured f u e l must be addressed. The m a t t e r o f a r r i v i n g a t a p r a c t i c a l c r i t e r i o n i s complicated and i s c u r r e n t l y under study by t h e Special P r o j e c t s D i v i s i o n o f Lawrence Livermore Laboratory, where an in-depth a n a l y s i s o f t h e weapons u t i l i t y o f f i s s i l e m a t e r i a l ( i n c l u d i n g 233U w i t h various enrichments) f o r t h e Non-Pro1 i f e r a t i o n A1 t e r n a t e Systems Assessment Program (NASAP) i s

L

being conducted i n accordance w i t h a work scope developed by t h e I n t e r n a t i o n a l S e c u r i t y A f f a i r s D i v i s i o n ( I S A ) and t h e management o f t h e NASAP Program. Unfortunately, t h e r e s u l t s o f t h e LLL study are n o t y e t a v a i l a b l e .

Because o f t h e considerable impact o f enrichment

considerations on t h e u t i l i t y o f p a r t i c u l a r r e a c t o r s and p a r t i c u l a r symbiotic systems, i t seems best a t t h i s p o i n t t o discuss t h e several approaches f o r determining t h e guidel i n e s f o r t h e enrichment o f 233U-238U mixtures and t o make a determination based on t h e LLL study a t a l a t e r time.


3.18

There are three approaches which can be employed t o estimate allowable enrichnent c r i t e r i a f o r 233U i n 238U corresponding t o t h e s t a t u t o r y 20% l i m i t s e t f o r 235U i n 238U. These three c r i t e r i a are: ( 1 ) c r i t i c a l mass, (2) i n f i n i t e m u l t i p l i c a t i o n f a c t o r , and These can be employed s i n g u l a r l y o r i n combination as discussed below.

(3) yield.

C r i t i c a l Mass As s t a t e d above, the bare-sphere c r i t i c a l mass o f m e t a l l i c 20% 235U and 80% 238U i s about 850 kg.

This amount can be reduced by a f a c t o r o f two t o three by t h e use o f a

neutron r e f l e c t o r .

However, the s i z e and weight o f the combination o f r e f l e c t o r and

f i s s i l e m a t e r i a l w i l l n o t be s u b s t a n t i a l l y l e s s than t h a t o f the bare sphere, and may even be greater.

I n a d d i t i o n , f o r a nuclear explosive, an assembly scheme must be added

which w i l l increase t h e s i z e and weight s u b s t a n t i a l l y .

Concentrations o f

235U,

233U, o r

plutonium i n mixtures w i t h 238U such t h a t they have bare-sphere m e t a l l i c c r i t i c a l masses o f about 850 kg represent one possible reasonably conservative c r i t e r i o n f o r a r r i v i n g a t concentrations below which t h e m a t e r i a l i s n o t usable i n p r a c t i c a l nuclear weapons. This 850 kg bare-sphere c r i t i c a l mass c r i t e r i o n can a l s o be used f o r o t h e r m a t e r i a l s which are o r might be i n nuclear f u e l cycles. Although t h i s c r i t e r i o n provides a basis f o r cons i s t e n t safeguards requirements f o r 2 3 3 U o r 235U embedded i n 238U, i t leans t o r a t h e r low l i m i t s . I n f i n i t e Mu1t i p 1 i c a t i o n Factor h o t h e r possible c r i t e r i o n i s the one associated w i t h the i n f i n i t e m u l t i p l i c a t i o n f a c t o r .,k attained. takes ,k

For a weapon t o be successful, a c e r t a i n degree o f s u p e r c r i t i c a l i t y must be D. P. Smith o f Los Alamos S c i e n t i f i c Laboratory has adopted t h i s approach. He = 1.658 f o r 20% 235U-enriched uranium, which i m p l i e s ,k

= 1.5346 f o r t h e oxide.

He then performs a search c a l c u l a t i o n on enrichment f o r the o t h e r systems so as t o o b t a i n We note t h a t f o r 2 3 3 U t h e l i m i t s t h e same ,k value. His r e s u l t s are shown i n Table 3.3-1. are 11.65% 2 3 3 U f o r t h e oxide and 11.12% 2 3 3 U f o r t h e metal. Table 3.3-1

Equivalent Enrichment L i m i t s

Fuel

H a t e r i a1

kco

Metal

20% 235U, 80% 238U

1.658

Oxide

11.12% 233U, 88.88% 238U

1.658

11.11% 239Pu, 88.89% 238U

1.658

(20% 235U, 80% 238U)02

(11.65% 233U, 88.35% 238U)02

1 5346 1.5346

(13.76% 2 3 9 P ~ , 86.24% 238U)02 (14.5% 239Pu, 1.5% 240Pu, 85% 238u)02

1.5346 1.5344

~

~

~~~

~~~

~

These numbers were obtained by D. P. Smith of Los Alamos S c i e n t i f i c Laboratory from DTF I V c a l c u l a t i o n s using Hansen-Roach cross sections.


3-19

Yield It may a l s o be p o s s i b l e t o s e t a minimum y i e l d f o r a p r a c t i c a l nuclear explosive

device.

An obvious consideration here i s t h a t i n attempting t o achieve s u p e r c r i t i c a l i t y

w i t h i n c r e a s i n g amounts o f f i s s i l e m a t e r i a l o f decreasing enrichment, a p o i n t i s reached where t h e y i e l d o f an equivalent mass o f chemical h i g h explosive exceeds the nuclear explosive y i e l d .

The LLL Special Projects D i v i s i o n i s c u r r e n t l y i n v e s t i g a t i n g

the p o s s i b i l i t y o f e s t a b l i s h i n g such a l i m i t .

t


3-20

3.3.2.

F a b r i c a t i o n and Handling o f Denatured Fuel J. D. Jenkins R. E. Brooksbank Oak Ridge National Laboratory

The techniques r e q u i r e d f o r f a b r i c a t i n g and handling 233U-containing f u e l s encount1

ered i n t h e denatured f u e l c y c l e d i f f e r from those employed f o r 235U f u e l s because o f t h e high gamma-ray and a l p h a - p a r t i c l e a c t i v i t i e s present i n t h e 233U fuels. Some idea of t h e r a d i a t i o n l e v e l s t h a t w i l l be encountered can be deduced from recent r a d i a t i o n measurements f o r a can t h a t contains 500 g o f 233U w i t h a 232U content o f 250 ppm and has been aged 12 years since p u r i f i c a t i o n .

The r e s u l t s were as follows:

Distance

Radiation (mr/hr)

Contact 1 ft 3 ft

250,000 20,000 2,000

I;

These r a d i a t i o n l e v e l s are equivalent t o those t h a t could be expected a t t h e same distances i

from 500 g o f 2 3 3 U containing % 1250 ppm 2 3 2 U and aged s i x months, which i s comparable w i t h 233U t h a t has undergone several cycles i n a f a s t breeder r e a c t o r . With such h i g h a c t i v i t i e s , complete alpha containment o f t h e f u e l w i l l be required, and a l l personnel must be protected from t h e f u e l w i t h t h i c k b i o l o g i c a l s h i e l d i n g (several f e e t o f concrete o r t h e equivalent).

L II

This, o f course, necessitates remote-handling operations, which c o n s t i t u t e s an i n h e r e n t safeguard against t h e d i v e r s i o n o f t h e f u e l w h i l e i t i s being f a b r i c a t e d and/or handled.

The requirement for remote operation i s f u r t h e r borne o u t by experience gained i n two e a r l i e r programs i n which 233U-containing f u e l s were fabricated.

I n these two programs, t h e "Kilorod" program' and t h e L i g h t Water Breeder Reactor (LWBR) program,2 (233U,Th)02 p e l l e t s could be f a b r i c a t e d i n glove boxes, b u t o n l y because t h e 233U used contained extremely low ( < l o ppm) amounts o f 232U. Even so, t h e time frame f o r f u e l fabr i c a t i o n was severely r e s t r i c t e d and e x t r a o r d i n a r y e f f o r t s were r e q u i r e d t o keep the con-

G

tamination l e v e l o f aged 233U s u f f i c i e n t l y low t o permit continued glove box operation. i

Based on experience a t ORNL i n t h e preparation o f n e a r l y two tons o f 233U02 f o r t h e LWBR program, i t was determined t h a t t h e handling o f kilogram q u a n t i t i e s o f 233U containing 10 ppm o f 232U and processed i n unshielded glove boxes 25 days a f t e r p u r i f i c a t i o n (complete daughter removal) t o produce 233U02 powder r e s u l t e d i n personnel r a d i a t i o n exposures o f 50 mr/man-week.

The techniques used i n preparing K i l o r o d and LWBR f u e l would n o t be f e a s i -

b l e i n a large-scale f a b r i c a t i o n p l a n t using expected i n recycled

23%.

233U

containing t h e 100 t o 2000 ppm 232U

Therefore, one must conclude t h a t remote f a b r i c a t i o n , behind

several f e e t o f concrete shielding, w i l l be r e q u i r e d f o r 233U-bearing LWR and FBR f u e l s . Remote operation w i l l impact t h e f a b r i c a t i o n process and t h e f u e l form.

For ex-

ample, LWR and LMFBR f u e l s can be manufactured e i t h e r as oxide p e l l e t s o r as sol-gel microspheres.

The many powder-handling operations r e q u i r e d i n f a b r i c a t i n g p e l l e t s w i t h

1'

L


3-21

..

L J

L

L

t h e i r i n h e r e n t d u s t i n g problems and t h e many mechanical operations r e q u i r e d i n blending powder, pressing, s i n t e r i n g , and g r i n d i n g p e l l e t s make remotely operating and maintaining a 233Ubearing p e l l e t f a b r i c a t i o n l i n e d i f f i c u l t . A l t e r n a t i v e l y , t h e r e l a t i v e ease o f handling l i q u i d s and microspheres remotely makes t h e sol-gel spherepac process appear more amenable t o remote operation and maintenance than powder preparation and p e l l e t i z i n g processes, a1though t h e process i s l e s s f u l l y developed. D e t a i l e d analyses o f s p e c i f i c f l o w sheets and process l a y o u t s f o r a p a r t i c u l a r f u e l form would be r e q u i r e d t o q u a n t i t a t i v e l y determine t h e r e l a t i v e safeguards m e r i t s o f one process versus another.

.- --

L

I n general, however, batch processes where c o n t r o l o f special

nuclear m a t e r i a l s can be e f f e c t e d by i t e m a c c o u n t a b i l i t y are e a s i e r than continuous processes i n which t h e m a t e r i a l i s contained i n l i q u i d form. Thus, i n our example above, an assessment might conclude t h a t some s a c r i f i c e s must be made i n m a t e r i a l a c c o u n t a b i l i t y i n order t o achieve-remote f u e l f a b r i c a t i o n .

,i--

L r-

ia

t

The o v e r r i d i n g safeguards consideration i n denatured f u e l f a b r i c a t i o n however i s t h e remote n a t u r e o f t h e process i t s e l f , which l i m i t s personnel access t o t h e f i s s i l e m a t e r i a l . Access i s n o t impossible, however, f o r two reasons. F i r s t , f o r m a t e r i a l and equipment t r a n s f e r , t h e processing c e l l s w i l l be l i n k e d t o o t h e r c e l l s o r t o o u t - o f - c e l l mechanisms. Second, some p o r t i o n s o f t h e processing equipment may be maintained by persons who e n t e r t h e c e l l s a f t e r appropriate source shdelding o r source removal. Thus, some c e l l s may be designed f o r personnel access, b u t a l l access p o i n t s w i l l be c o n t r o l l e d because o f t h e requirement f o r a l p h a - a c t i v i t y containment. Health physics r a d i a t i o n monitors would provide an i n d i c a t i o n o f breach o f containment and o f p o s s i b l e diversion. Because t h e ingress p o i n t s from t h e c e l l s w i l l be l i m i t e d , p o r t a l monitors may a l s o provide , a d d i t i o n a l

t t -

safeguards assurance. It should be noted t h a t although kilogram q u a n t i t i e s o f m a t e r i a l represent highr a d i a t i o n l e v e l s from t h e standpoint o f occupational exposures, t h e l e v e l s o f r e c e n t l y

p u r i f i e d 23% a r e low enough t h a t d i r e c t handling o f t h e m a t e r i a l f o r several days would not r e s u l t i n noticeable health effects. The remote nature of t h e r e f a b r i c a t i o n process r e q u i r e s h i g h l y automated machinery

L

f o r most of t h e fabrication. Elaborate c o n t r o l and monitoring instrumentation w i l l be r e q u i r e d f o r automatic operation and process c o n t r o l and can provide a d d i t i o n a l data f o r m a t e r i a l a c c o u n t a b i l i t y and m a t e r i a l balance consistency checks. The remote nature o f t h e process has t h e p o t e n t i a l of s u b s t a n t i a l l y improving t h e safeguarding o f t h e r e c y c l e f u e l d u r i n g r e f a b r i c a t i o n . The e x t e n t of t h i s improvement w i l l depend on t h e s p e c i f i c f a c i l i t y design and on t h e degree t o which t h e a d d i t i o n a l r e a l - t i m e process i n f o r m a t i o n can enhance t h e safeguards system.


3-22

3.3.3

Detection and Assay o f Denatured Fuel D. T. I n g e r s o l l Oak Ridge National Laboratory

The r e l a t i v e l y high gamma-ray a c t i v i t y o f 2 3 3 U f u e l s , enriched o r denatured, has opposite e f f e c t s on d e t e c t i o n and assay: i t increases t h e d e t e c t a b i l i t y o f t h e fuels b u t i t a so increases t h e d i f f i c u l t y o f passive gamma assay. That t h i s s i t u a t i o n e x i s t s \

i s apparent from Fig. 3.3-2, which presents a Ge(Li)-measured gamma-ray spectrum3 from a 233U sample containing 250 ppm 232U. A l l major peaks i n the spectrum are from t h e decay products o f 232U, which i s near secular e q u i l i b r i u m w i t h t h e products. The presence

G

o f the 2.6-MeV gamma r a y emitted by 208T1 provides a useful handle f o r t h e d e t e c t i o n of m a t e r i a l s t h a t contain even small q u a n t i t i e s o f 232U, thus p r o v i d i n g a basis f o r preventing f u e l d i v e r s i o n and/or f o r recovering d i v e r t e d f u e l . On t h e o t h e r hand, t h e presence of numerous gamma rays i n t h e spectrum eliminates the p o s s i b i l i t y o f d i r e c t gamma-ray assay o f t h e f i s s i l e isotope. I n d i r e c t assay using t h e 2 3 2 U gamma rays would be i m p r a c t i c a l , since i t would r e q u i r e a d e t a i l e d knowledge o f the h i s t o r y o f t h e sample. Detection systems are already a v a i l a b l e .

A Los Alamos S c i e n t i f i c Laboratory (LASL) r e p o r t describes a doorway monitor system" t h a t employs a 12.7- x 2.5-cm NaI(T1) d e t e c t o r and has been used t o measure a dose r a t e o f about 2.5 mr/hr a t a distance o f 30 cm from a

fl

c

20-9 sample o f Pu02.

Approximately t h e same dose r a t e would be measured f o r a s i m i l a r sample o f 233U containing 100 ppm o f 232U o n l y 12 days f o l l o w i n g t h e separation o f daught e r products. The dose r a t e would increase by a f a c t o r o f 10 a f t e r 90 days and by an a d d i t i o n a l f a c t o r o f 4 a f t e r one year.5 Also, t h e gamma-ray dose r a t e scales l i n e a r l y w i t h 232U content and i s n e a r l y independent o f t h e type o f b u l k m a t e r i a l , i.e., 233U, ,35U, or 2 3 8 ~ . The n e t counting r a t e f o r t h e PuO, sample (shielded w i t h 0.635 cm o f lead) was 1000 cps. The observed background was 1800 cps, r e s u l t i n g i n a signal-to-noise r a t i o o f o n l y 0.6.

S i m i l a r samples o f 232U-contaminated uranium n o t o n l y would y i e l d higher count-

i n g rates, b u t could a l s o y i e l d considerably b e t t e r signal-to-noise r a t i o s i f t h e d e t e c t o r window were s e t t o cover o n l y t h e 2.6-MeV gamma r a y present i n t h e spectrum. Although t h e denaturing o f uranium f u e l s tends t o d i l u t e t h e 232U content, t h e a n t i c i p a t e d 232U l e v e l s i n most denatured fuels i s s t i l l s u f f i c i e n t l y high f o r r e l a t i v e l y easy detection, except immediately a f t e r complete daughter removal. The d i f f i c u l t y i n performing nondestructive assays (NDA) o f denatured f u e l s r e l a t i v e t o h i g h l y enriched f u e l s i s a t t r i b u t a b l e t o two e f f e c t s : (a) t h e desired s i g n a l (emitted neutrons o r gama rays, heat generation, etc.) i s reduced because o f t h e m a t e r i a l d i l u t i o n , and (b) t h e s i g n a l i s mostly obscured by t h e presence o f 232U.

The l a t t e r problem

e x i s t s because although denaturing reduces t h e t o t a l concentration o f 232U, t h e r e l a t i v e p r o p o r t i o n o f 232U t o f i s s i l e m a t e r i a l remains t h e same. This i s an e s p e c i a l l y s i g n i f i cant problem w i t h passive NDA techniques.

As i s shown i n Fig. 3.3-2,

t h e gamma-ray

L L


3-23

spectrum from a 232U sample containing 250 ppm o f 232U i s t o t a l l y dominated by t h e decay gamma rays, thus e l i m i n a t i n g t h e p o s s i b i l i t y o f d i r e c t gamma-ray assay.

232U

Passive

techniques employing c a l o r i m e t r y are a l s o complicated since 232U decay p a r t i c l e s can cont r i b u t e s i g n i f i c a n t l y t o t h e heat generation i n a f u e l sample. I t has been calculated,3,6 t h a t f o r a fresh sample of 233U containing 400 ppm 232U, n e a r l y 50% o f t h e thermal heat generation can be a t t r i b u t e d t o 232U decay, which increases t o 75% a f t e r o n l y one year.

I t i s , therefore, apparent t h a t f i s s i l e content assay f o r denatured uranium f u e l s w i l l r e q u i r e more s o p h i s t i c a t e d a c t i v e NDA techniques which must overcome t h e obstacles of m a t e r i a l d i l u t i o n and 2 3 2 U - a ~ t i v i t ycontamination.

CHANNEL NUMBER

Gama-Ray S ectrum from a 233U Sample Containing 250 ppm 232U. A l l Fig. 3.3-2. major peaks a r e a t t r i b u t e d t o 2U decay products. Gamma-ray energies i n d i c a t e d i n MeV. (From ref. 3.)

!


3-24

3.3.4.

P o t e n t i a l Circumvention o f t h e I s o t o p i c B a r r i e r o f Denatured Fuel

E. H, G i f t and W. B. A r t h u r Oak Ridae Gaseous D i f f u s i o n P l a n t I f a large-scale denatured-uranium r e c y c l e program i s f u l l y implemented ( w i t h secure

energy centers), many types o f both f r e s h ( u n i r r a d i a t e d ) and spent f u e l may be i n t r a n s i t I n order t o ensure t h a t these f u e l s are p r o l i f e r a t i o n r e s i s t a n t , they

throughout t h e world.

must meet t h e basic c r i t e r i o n t h a t a s u f f i c i e n t q u a n t i t y o f f i s s i l e m a t e r i a l cannot be chemically extracted from seized elements f o r d i r e c t use i n t h e f a b r i c a t i o n o f a nuclear weapon.

L

As pointed o u t i n previous sections o f t h i s r e p o r t , t h e a d d i t i o n o f t h e denaturant

238U t o t h e f i s s i l e isotope 233U w i l l prevent t h e d i r e c t use o f t h e uranium i n weapons manufacture p r o v i d i n g t h e 2 3 % content o f t h e uranium remains below a s p e c i f i e d l i m i t , which f o r t h i s study has been s e t a t 12% (see Section 3.3.1).

Thus, even i f t h e uranium were

chemically separated from t h e thorium f e r t i l e m a t e r i a l included i n t h e elements, i t could n o t be used f o r a weapon. S i m i l a r l y , i f t h e 235U content o f uranium i s kept below 20%, t h e uranium would n o t be d i r e c t l y usable.

For t h e discussion presented here, i t i s f u r t h e r

assumed t h a t fuels containing both 233U and average l i e s between these l i m i t s .

235U

w i l l meet t h i s c r i t e r i o n i f t h e i r weighted

With t h e chemical i s o l a t i o n o f t h e primary f i s s i l e isotopes thus precluded, two potent i a l means e x i s t f o r e x t r a c t i n g f i s s i o n a b l e . m a t e r i a l f o r t h e denatured f u e l :

(1) i s o t o o i c

separation o f t h e f r e s h f u e l i n t o i t s 2 3 % ( o r 235U) and 238U components; and (2) chemical e x t r a c t i o n from t h e spent f u e l o f t h e 239Pu bred i n t h e 238U denaturant o r chemical e x t r a c t i o n of t h e intermediate isotope 233Pa t h a t would subsequently decay t o

233U.

I n t h i s examination

of t h e p o t e n t i a l circumvention o f t h e i s o t o p i c b a r r i e r o f denatured f u e l b o t h these p o s s i b i l i t i e s a r e discussed; however, t h e p r o b a b i l i t y o f the second one a c t u a l l y being c a r r i e d o u t i s e s s e n t i a l l y discounted.

Thus t h e emphasis here i s on t h e p o s s i b i l i t y t h a t would-be p r o l i f e r a t o r s would o p t f o r producing weapons-grade uranium through t h e c l a n d e s t i n e o p e r a t i o n o f an i s o t o p e separation f a c i l i t y .

For t h e purposes o f t h i s study i t i s assumed t h a t t h e seized f u e l i s i n

t h e form of f r e s h L W R elements o f one o f t h e f o l l o w i n g f u e l types:

A.

Approximately 3% 2 3 5 b e n r i c h e d uranium (same as c u r r e n t l y used L W R f u e l ) .

B.

Recycle uranium from a thorium breeder blanket, denatured t o %12% 233U w i t h depleted uranium.

C.

F i f t h - g e n e r a t i o n r e c y c l e of f u e l type B w i t h

233U

f i s s i l e makeup from a thorium

breeder blanket. D.

F i r s t c y c l e o f 235U-238U-Th f u e l assuming no 233U i s a v a i l a b l e from an external source. I n t h i s f u e l scheme t h e 235U concentration i n uranium can be as h i g h as 20% (see above).

E.

F i r s t r e c y c l e o f f u e l type D w i t h 93% 235U i n uranium makeup. I n t h i s f u e l i n g option, n o t a l l o f t h e f u e l i n a r e l o a d batch w i l l contain r e c y c l e uranium. Some p o r t i o n o f t h e r e l o a d batch w i l l contain f u e l type D. This o p t i o n i s analogous t o t h e " t r a d i t i o n a l " concept envisioned f o r plutonium r e c y c l e fuels.

It allows some o f t h e fuel

h: Id: L

b'

c


3-25

This f u e l i n g o p t i o n w i l l be r e f e r r e d

t o be f a b r i c a t e d i n nonradioactive f a c i l i t i e s .

t o i n t h e remainder o f t h e t e x t as f u e l r e c y c l e Option 1. id i

L

F.

E

F i f t h - g e n e r a t i o n r e c y c l e o f f u e l types 0 and

with 93%

235U makeup (Option 1).

6. F i r s t r e c y c l e of' f u e l type 0, w i t h r e c y c l e uranium i n a l l f u e l assemblies o f a r e l o a d batch. Makeup uranium i s 20% and 93% 235U as needed t o maintain r e a c t i v i t y . I n t h i s o p t i o n a l l f u e l would probably r e q u i r e remote f a b r i c a t i o n f a c i l i t i e s .

This f u e l i n g

o p t i o n w i l l be r e f e r r e d t o i n t h e remainder o f t h e t e x t as fuel r e c y c l e Option 2.

H.

1

F i f t h r e c y c l e o f f u e l type G w i t h

The uranium compositions o f these f u e l s are shown i n Table 3.3-2.

h id

hi

I

I n a d d i t i o n t o these, i t

should be assumed t h a t n a t u r a l uranium i s a l s o available. Table 3.3-2.

L i

235U makeup (Option 2).

Isotope

A

23211

0

23311

0

Z3"U

1.2

lo-'

C

6

5.02

x

Uranium Fuel Mixtures That May Be A v a i l a b l e (Weight F r a c t i o n i n Urani um)

x

lo-'

6.565

x

0

10-4 0

E 1.2363

F x

10-4 2.445

x

H

G

lo-'

1.134

x

lo-'

2.331

x

10''

0.118611

0.11498

0

0.047004

0.0591 4

0.04310

0.05638

0.008523

0.035108

0.001754

0.005430

0.02115

0.005125

0.020245

235U

0.032

0.002317

0.01255

0.2000

0.13201

0.11 3457

0.1 3765

0.11749

236U

0

0.000036

0.005327

0

0.02303

0.056496

0.021119

0.05386

0.870011

0.831228

0.798246

0.792389

0.749522

0.793021

0.75188

0.96788 238U

Description of Fuel Type: A 3.2 w t X z35U from natural uranium. 6 Thorium breeder blanket fuel denatured w i t h depleted uranium. C Flfth generation recycle of B w i t h thorium breeder blanket makeup. 0 20 w t % 23511 from natural uranium. E First recycle of D w i t h 93 w t X 235U i n uranium makeup (Option 1. see note). F Flfth generation recycle of 0 w i t h 93 w t % 235U i n uranium makeup (Option 1. see note). G First recycle of 0 with 93 w t X 235U makeup (Option 2, see note H Fifth recycle of D w i t h 93 w t X 235U makeup (Option 2. see notel.

------

NOTE:

.

Fuel types E and F are deslgned so that not a l l of the fuel i n a reload batch I s recyclf fuel; some of the reload batch w i l l contain fuel type D. T h i s situation Is analogous t o the "traditional concept envisioned for plutonium recycle fuels. This concept allows some of the fuel t o be fabricated i n non-radioactive f a c i l l t i e s , and is referred t o In the text as fuel recycle Option 1. Fuel types G and H result i f every assembly i n the reload batch contains recycle fuel. The fueling mde 1s referred t o as Option 2.

I s o t o p i c Separation o f Fresh Fuel 1

O f t h e various uranium isotope separation processes gaseous d i f f u s i o n ,

h

which have been conceived, o n l y t h e c u r r e n t technology processes (i.e.,

L

gas c e n t r i f u g e , t h e Becker nozzle and t h e South A f r i c a n f i x e d w a l l c e n t r i f u g e ) and p o s s i b l y t h e c a l u t r o n process could be considered as near-term candidates f o r a clandestine f a c i l i t y capable o f e n r i c h i n g divered r e a c t o r f u e l . O f these, t h e gas c e n t r i f u g e may be the preferred

i

S e l e c t i o n o f Separation F a c i l i t y .

technology.

This conclusion i s d i r e c t l y r e l a t e d t o t h e proven advantages o f t h e process,

which i n c l u d e a h i g h separation f a c t o r per machine, low e l e c t r i c a l power needs, and t h e a d a p t a b i l i t y t o small low-capacity b u t high-enrichment plants. Further, more n a t i o n a l groups (i.e.,

t h e U.S.,

England, Holland, Germany, Japan, A u s t r a l i a , and France) have operated


3-26

-

e i t h e r l a r g e c e n t r i f u g e p i l o t p l a n t s o r small commercial-sized plants, more so than f o r any o t h e r enrichment process, so i t i s apparent t h a t t h i s technology i s widely understood and applied. A b r i e f d e s c r i p t i o n o f t h e c e n t r i f u g e process, as w e l l as d e s c r i p t i o n s o f other c u r r e n t and f u t u r e separation technologies, i s given i n Appendix A. The a p p l i c a t i o n o f c e n t r i f u g e technology t o a small p l a n t capable o f producing a couple o f hundred kilograms o f uranium enriched t o 90% 235U has n o t proved t o be i n o r d i n a t e l y expensive. Two examples can be provided. An a r t i c l e appearing i n two journals7'8 presents This p l a n t , which could be operational i n 1980,, i s designed t o produce 50 MT SWU/yr i n a 7000-machine f a c i l i t y . The t o t a l c o s t o f t h e f a c i l i t y was estimated by t h e Japanese t o be $166.7 m i l l i o n . Simple a r i t h m e t i c y i e l d s information on a proposed Japanese c e n t r i f u g e p l a n t .

the i n d i v i d u a l c e n t r i f u g e separation capacity o f 7 kg SWU/yr and a c e n t r i f u g e c o s t o f approximately $24,000

(which includes i t s share o f a l l p l a n t f a c i l i t i e s ) .

An upper l i m i t f o r t h e cost o f developing a small gas c e n t r i f u g e enrichment f a c i l i t y can be estimated from published costs from t h e United States uranium gas c e n t r i f u g e program.

A paper by Kiserq provides a convenient summary o f t h e s t a t u s and cumulative costs f o r t h e U.S. program. The Component Test F a c i l i t y , a p l a n t which i s expected t o have a separative capacity o f 50 MT SWU/yr (see Appendix A), was operational i n January o f 1977. To t h a t date, t h e cumulative cost o f t h e e n t i r e U.S.

gas c e n t r i f u g e program was given as about $310 m i l l i o n . Of t h i s t o t a l , about $190 m i l l i o n was i d e n t i f i e d as development costs. The remaini n g $120 m i l l i o n was i d e n t i f i e d as equipment and f a c i l i t y expense. Further, o n l y about $30 m i l l i o n was i d e n t i f i e d as being technology i n v e s t i g a t i o n . Even more i n t r i g u i n g i s t h a t w i t h i n t h e i n i t i a l 3-year development program (beginning i n 1960 and budgeted a t $6 m i l l i o n ) , t h e f o l l o w i n g accomplishments were recorded. a.

The operating performance o f t h e gas centrifuge was g r e a t l y improved.

b.

Small machines were successfully cascaded i n 1961 (one year a f t e r i n i t i a t i o n o f t h e contract).

c.

When the l a s t of these u n i t s was shut down i n 1972, some machines had run continuously f o r about e i g h t years.

That these c e n t r i f u g e s were n o t commercially competitive w i t h gaseous d i f f u s i o n may be i r r e l e v a n t when they are considered as a candidate f o r a clandestine enrichment f a c i l i t y . Thus, as s t a t e d above, o f t h e c u r r e n t technologies, t h e c e n t r i f u g e process would probably be selected.

The u t i l i z a t i o n o f t h e developing technologies (laser, plasma, etc,) f o r a clandestine enrichment f a c i l i t y i s n o t c u r r e n t l y feasible. Successful development o f these technologies by any o f t h e numerous n a t i o n a l research groups would make them candidates f o r such a f a c i l i t y , however, and they would o f f e r t h e decided advantages o f a h i g h separat i o n factor,

low-power requirement and modular construction.

E f f e c t o f 232U on the Enrichment Process and Product.

A1

fuels containing

23311

also

As mentioned e a r l i e r i n t h s r e p o r t , t h e daughter products from Z3*U (t+= 72 y r ) release h i g h l y energetic gama rays and alpha p a r t i c l e s t h a t can complicate both t h e enrichment process and t h e subsequent weapon f a b r i c a t i o n . contain s u b s t a n t i a l amounts o f

232U.


3-27

As a f i r s t step i n evaluating the e f f e c t of 232U on the enrichment process and t h e en-

L L. L

r i c h e d product, consider f u e l types B and C from Table 3.3-2 as feed t o an enrichment plant. For making an acceptable weapon a f i s s i l e content o f 90% 233U + 235U i n the product should be satisfactory.

An acceptable product f l o w r a t e from such a p l a n t might be 100 kg U/yr.

Based on these assumptions, the product concentrations shown i n Table 3.3-3 were obt a i n e d from multicomponent enrichment c a l c u l a t i o n a l methods.lo This t a b l e i l l u s t r a t e s t h a t w h i l e a s u f f i c i e n t l y f i s s i l e uranium i s produced, a t a r e l a t i v e l y low feed rate, t h e product has a l s o concentrated the h i g h l y gama a c t i v e (through i t s decay daughters) 232U by about a f a c t o r o f 10. Greater than 99% o f t h e 232U i n the enrichment p l a n t feed w i l l be present i n the product.

I 1

I n t h e enrichment p l a n t t h e 232U concentration gradient from t h e feed p o i n t w i l l drop r a p i d l y i n the s t r i p p i n g section. I n t h e t a i l s the 232U concentration w i l l be reduced by about a f a c t o r o f 150 from the feed concentration. As a r e s u l t , the gamna

I,

Calculations have been made f o r a t y p i c a l c e n t r i f u g e enrichment p l a n t t o i l l u s t r a t e the gama r a d i a t i o n l e v e l t h a t could be expected a t e q u i l i b r i u m as a f u n c t i o n of the

r a d i a t i o n l e v e l s i n t h e enrichment p l a n t can be expected t o vary by a f a c t o r of greater than 1000 from the t a i l s t o t h e product.

2 3 2 U concentration.11

These r e s u l t s a r e shown i n Table 3.3-4.

I m p l i c i t i n these estimates

i s the assumption t h a t t h e daughter products o f 232U are a l l deposited w i t h i n t h e enrichment f a c i l i t y . This assumption seems j u s t i f i e d since t h e f l u o r i d e compound o f t h e f i r s t daughter product, 228Th ($ = 1.9 years), i s nonvolatile. With the exception o f 224Ra

(t+ * 3.6 d ) , a l l o f t h e other daughters have very s h o r t l i v e s . Experimentally, l i t t l e evidence e x i s t s t o determine the t r u e f r a c t i o n a l deposition Current evidence i s incorporated i n the e x i s t i n g s p e c i f i c a t i o n s f o r UF6 feed t o the gaseous d i f f u s i o n plants.12 These s p e c i f i c a t i o n s c a l l f o r a maximum 232U

of 232U daughters.

L

concentration o f 110 p a r t s o f 232U per b i l l i o n p a r t s of 2 3 5 1 ) i n t h e feed.

.

A t t h i s concentra-

t i o n , t h e r a d i a t i o n l e v e l s would be s i g n i f i c a n t i n a h i g h l y enriched product (*270 mr/hr a t 1 ft and 3 mr/hr on the p l a n t equipment). Based on Tables 3.3-3 and 3.3-4,

the maximum gamma r a d i a t i o n l e v e l i n a p l a n t

e n r i c h i n g 233U t o 90%would be about 2 r / h r a t equilibrium. A t t h i s r a d i a t i o n l e v e l , l i t t l e decomposition o f e i t h e r l u b r i c a t i o n o i l s o r the UF6 gas would occur. Some evidenceâ&#x20AC;? e x i s t s t o show t h a t a t t h i s r a d i a t i o n l e v e l the v i s c o s i t y o f the l u b r i c a t i n g o i l s would be 1

ili

i;â&#x20AC;?

unaffected over a 20-year p l a n t l i f e . Thus, there should be no bearing problem. I t i s a l s o expected t h a t the UF6 would be f a i r l y s t a b l e t o t h e combined alpha and gama r a d i a t i o n l e v e l s . A t the 2 - r l h r l e v e l , l e s s than one-tenth o f the mean inventory o f the machine would be decomposed per year. This m a t e r i a l would be expected t o be d i s t r i b u t e d f a i r l y uniformly throughout the machine w i t h perhaps s l i g h t l y higher accumulation on the withdrawal scoops. Since t h e i n d i v i d u a l machine inventory would be very low, t h i s should n o t be a s i g n i f i c a n t l o s s o f material.


3-28

Table 3.3-3. Enriched Proauct Compositions (Weight F r a c t i o n i n Uranium) Isotope

Feed 5.02

2321)

!

lom4

x

Fuel Type c Feed Product

Fuel Type B Product 4.1545

x

10-3

6.565

x

loe4

5.626

x

233u

0.11861 1

0.90

0.11498

0.90

2341)

0.008523

0.03757

0.0351 08

0.0901

2351)

0.00231 7

0.00376

0.01255

0.00379

3.6

23611

0.87001 1

2381)

2s3U

10-5

10-5

0.05450

0.005327

1.73

0.831 228

3.124

0.01

i n Tails

Feed Flow, kg U/yr

1.98

10-4 x

lo-'+

0.01 859

832

Product F1ow, kg U/yr

100

100

When removed from the plant, the UF,

product would be condensed and probably stored i n

I f i t i s assumed t h a t t h e c y l i n d e r s were sized t o hold 16 kg o f UF6, t h e

monel cylinders.

gamma dose r a t e s t h a t could be expected from t h e unshielded c y l i n d e r s are as shown i n Table 3.3-5. To reduce these product dose r a t e s t o acceptable l e v e l s would r e q u i r e substant i a l shielding.

As an example, Table 3.3-6

shows the s h i e l d i n g r e q u i r e d t o reduce t h e dose

r a t e a t 1 f t t o 1.0 and 50 mr/hr.

Table 3.3-4.

Gamma Radiation Level i n an Enrichment P l a n t as a Function o f 232U Concentration

Tjzu Concentration (wt X )

*

Radiation Level ( r / h r ) a t Equi 'I ib r i urn*

2.0

6.8

1.o

3.4

0.5

1.7

0.1

.34

0.001

.0034

0.0001

.00034

Within an i n f i n i t e a r r a y o f centrifuges.

bi


3-.29

krd

i

Table 3.3-5.

-

232U-Induced Gamna-Ray Dose Rates from Unshlelded Monel Cylinders Containing 16 kg o f UF,

Dose Rate ( r / h r ) Distance from Cy1i n d e r

Decay Time* (days)

0.1 w t % 232U

Contact

10

40.2 194

1,166

90

654

3,922

.

42,300

7,046

10

4.2

30

20.4

122

90

68.6

412

Equi 1.

1 Meter

242

30

Equi 1

1 Foot

0.6 w t % 2320

25.4

740

4,440

10

0.85

5.1

30

4.1

24.6

90

13.8

82.9

Equi 1.

149

894

*Time measured from chemical separation from thorium.

Shielding Required t o Reduce 232U-Induced Gamna-Ray Table 3.3-6. Dose Rates from Monel Cylinders Containing 16 kg o f UF6* Concrete Thickness Icm) . . Design Dose Rate (mr/hr)

1.o

Decay Time** (days)

50

LSZU

0.6 w t %

2~

101

120

90

114

132

138

157

30

62

80

90

74

92

98

116

Equi 1

t

wt %

30

Equil

1

D.1

. .

*Distance from source t o s h i e l d = 1 ft. **Time measured from chemical separation from thorium.

U


3-30

The high alpha a c t i v i t y of uranium containing

1.

I n t h e UF, there w i l l be a strong (u,n)

232U

reaction.

w i l l present two problems: A crude estimate o f t h e neutron

emission from a 16-kg UF, product c y l i n d e r containing 0.6 w t % neutrons/sec a t 10 days decay, 2.5 x decay.

2.

lo8

23%

i s 5.7 x

a t 30 days decay, and 8.7 x

lo8

io7

a t 90 days

The 232U w i l l provide a strong heat source i n t h e UF, and t h e metal products.

A

crude estimate o f t h e heat generation r a t e from pure 232U as a f u n c t i o n o f time a f t e r p u r i f i c a t i o n i s : 0.03 W/g a t 10 days, 0.13 W/g a t 30 days, and 0.46 W/g a t 90 days. The degree t o which these p r o p e r t i e s w i l l a f f e c t weapon manufacture o r d e l i v e r y i s unknown. A l t e r n a t i v e Enrichment Arrangements t o Reduce 232U Content i n t h e Product.

I n con-

s i d e r i n g t h e complications introduced t o t h e f i n a l uranium metal product, i.e., the radiat i o n l e v e l and heat generation r e s u l t i n g from 232U, i t i s apparent t h a t removal o f t h e 232U would be b e n e f i c i a l . Enrichment cascades can be designed t o accomplish t h i s . The most eff i c i e n t arrangement would be t o f i r s t design a cascade t o s t r i p 232U from a l l other uranium isotopes and then t o feed t h e t a i l s from t h e f i r s t cascade t o a second cascade where the f i s s i l e isotopes can be enriched. i s i l l u s t r a t e d i n Fig. 3.3-2.

Pmduct Containin Nearly All the 23%

This

Such an enrichment arrangement can be independent o f t h e s p e c i f i c e n r i c h i n g device. Based on t h e discussion o f t h e gas c e n t r i f u g e process i n Appendix A and a t t h e beginning o f t h i s section, a small, low separative work capacity machine may be w i t h i n t h e t e c h n i c a l c a p a b i l i t i e s o f a would-be d i v e r t e r (see Appendix A ) .

A1 though no i n f o r m a t i o n e x i s t s on t h e separative work capacity o f a Zippe machine

i n a cascade, a reasonable estimate of i t s separative capacity i s about 0.3 k g SWU/yr

cUaste Fig. 3.3-2. I l l u s t r a t i o n o f Enrichment Arrangement t o Produce Low 232U Content Urani urn.

when separating 235U from 238U.


Y

3-31

tpr) To f u r t h e r s p e c i f y t h e plant, i t can be assumed t h a t t h e d i v e r t e r would l i k e to:

Lj

1.

Minimize t h e feed and waste stream flows i n t h e f i r s t and second cascades consistent w i t h l i m i t i n g t h e number of c e n t r i f u g e s required.

b

L

2.

Achieve a s i g n i f i c a n t weapons-grade product f l o w rate.

(A f l o w r a t e o f 100 kg U/yr

having a f i s s i l e content o f 90% 233U t 235U was chosen.) 3.

Reduce t h e 232U content i n t h e metal product so t h a t contact manufacture can be achieved w i t h o u t serious r a d i a t i o n hazard. Based on these assumptions and considering t h e f u e l types l i s t e d i n Table 3.3-2,

L

a s e r i e s o f enrichment cascades, flows and selected i s o t o p i c parameters are presented

i n Table 3.3-7.

The basic c r i t e r i o n chosen f o r t h e f i n a l uranium product was t h a t t h e

232U concentration was about 1 ppm 232U i n t o t a l uranium. A t t h i s l e v e l t h e gama emission r a t e from t h e f i n a l metal product i s s u f f i c i e n t l y low t h a t most f a b r i c a t i o n and subsequent handling operations can be c a r r i e d o u t i n unshielded f a c i l i t i e s using

L 1 L

contact methods. The f i r s t enrichment cascade t o perform t h e separation o f 232U from t h e remaining uranium w i l l be very r a d i o a c t i v e . But i t w i l l be o n l y s l i g h t l y more r a d i o a c t i v e than i f o n l y one cascade were used and t h e z32U n o t separated f r o m t h e f i n a l product.

The t a b l e

shows t h a t a f a c t o r o f two increase i n Z32U product concentration w i l l provide s u f f i c i e n t decontamination w i t h o u t a p r o h i b i t i v e increase i n t h e number o f centrifuges. I f much g r e a t e r (by a f a c t o r o f 20) concentrations o f 232U can be t o l e r a t e d i n t h e cascade, some

t

b

r e d u c t i o n (420 t o 30%) can be made i n t h e necessary number o f centrifuges. Table 3.3-7

L L

d i il

a l s o shows a s t r i k i n g d i f f e r e n c e i n t h e number o f c e n t r i f u g e s r e q u i r e d

t o decontaminate t h e uranium product when t h e uranium makeup t o t h e thorium cycles i s 93% 235U r a t h e r than 2331) from t h e thorium breeder blanket. This r e s u l t s because w i t h t h e 235U r e c y c l e f u e l i t i s more advantageous, both i n c e n t r i f u g e s and i n annual feed requirements, t o design t h e separation t o throw away i n t h e f i r s t cascade waste stream much of t h e 233U and 2341) i n a d d i t i o n t o t h e 232U. Thus, t h e f i s s i l e content i n t h e f i n a l product from these f u e l mixtures i s n e a r l y a l l 235U. As a b e t t e r means o f measuring t h e p r o l i f e r a t i o n p o t e n t i a l o f t h e d i f f e r e n t f u e l mixtures, t h e data presented i n Table 3.3-7 have been r e c a s t i n Table 3.3-8

as a

function of t h r e e parameters: (1) t h e number of c e n t r i f u g e s needed, (2) the uranium feed requirements t o produce 100 k g l y r of 90% f i s s i l e uranium and (3) t h e number o f standard Westinghouse

PWR fuel assemblies t h a t must be d i v e r t e d .

Based on these c r i t e r i a , t h e f o l l o w i n g conclusions can be drawn w i t h respect t o

i,

d e s i r a b i l i t y o f f u e l s f o r diversion:


t;

3- 32

Table 3.3-7.

S u m r y o f Results o f Centrifuge Enrichment Survey o f Potential Fuel Mixture" ~~

Content (wt. Fraction) O f 1st O f 2nd Cascade Cascade I n i t i a l Product Product

232U

$T :

8

N A ~

0

A

5.02(-4)' 5.02 -4

5.021-41 5.02(-4)

C

6.564 -4) 6.5641-4) 6.564(-4) 6.564(-4) 0

0

PA 0.005

0.01 0.10

0

4.15(-3) 2.7 6 1.31161 8.1(-7)

Fissile Content (wt. Fraction) O f 2nd O f 2nd Cascard Cascade Tails Product 0.002

0.90

2993

0

29220

29220

0.005 0.01

0.90 0.90 0.90 0.90

832 3180 1302 817

0

82410 50600 41653

5468 10880 9981 7257

5468 93290 60581 48910

0.005

0.90 0.90 0.90 0.90

860 3000 1749 853

86227 61277 45483

9191 18302 18802 11277

9191 104529 80079 56760

0.01

0.90

0.005

0.005

0.01

5.626(-3) 0.005 2.68(-6) 0.01 1.63(-6) 0.005

0.1

8.5(-7)

NA

0.0065

NA

0

468

0

4991

4991

25244 15459 9292

7002

5921 13635

32246 21380 22927

0.90 0.90

3001 860

33033 11872

14398 20982

47431 32854

0.005

0.90

664

8758

13033

21791

0.0715

0.90 0.90

3000 805

32136

0.005

11889

12419 19477

44555 31366

0.002

0.90

17575

0

7791.8

77916

1.236 -4) 1.2361-4) 1.236 (-4)

0.001236 0.00235 0.00235

2.4!-6) 1.14(-6) 6.67(-7)

0.06 0.0

F

2.445(-4) 2.445(-4)

0.002445 0.003

2.63(-6) 7.87(-6)

0.. 15 0.005

G

1.134(-4)

0.003

6.42(-7)

H

2.331(-4) 2.331(-4)

0.0023 0.003

2.5(-6) 7.44(-7)

0

NA

0

0

3000 1210 704

E

Natural Uranium

Annual Feed (kg U/yr)

Number of Centrifuges Required (0.3 kg SWU/yr Zippes) I n 232U I n Fissile Stripping Enriching Cascade Cascade Total

0,065

"Feed and centrifuges needed t o produce 100 kg U/yr o f 90% f i s s i l e product. bSee Table 3.3-2 f o r description of fuel types. %cad: 5.02 x 'NA = not applicable.

T a b l e 3.3-8.

Enrichment Resistance o f Fuel M i x t u r e s I n v e s t i g a t e d *

Fuel Type

Centrifuges

Number o f

Feed Requirements (kg Ulyr)

Approximate Number o f PWR Fuel Assemblies Needed t o Supply Feed

2,993

6.7

A

3.2 w t % 235U

D

20 w t % 235U w i t h t h o r i u m

4,991

468

N a t u r a l uranium (0.711 w t % 235U)

77,918

17,575

1 s t g e n e r a t i o n 233U r e c y c l e w i t h t h o r i u m No 232U removal W i t h 232U removal

5,469 48,910

832 81 7

7.1 6.9

5 t h g e n e r a t i o n 233U r e c y c l e w i t h t h o r i u m No 232U removal With 232U removal

9,191 80,079

860 1,750

7.0 14.2

235U r e c y c l e w i t h t h o r i u m ( O p t i o n 1 ) With 232U removal

22,927

704

6.8

5 t h g e n e r a t i o n 235U r e c y c l e w i t h t h o r i u m ( O p t i o n 1 ) w i t h 232U removal

32,854

860

7.4

1 s t generation 235U r e c y c l e w i t h t h o r i u m (Option 2) With 232U removal

21,791

664

6.6

5 t h g e n e r a t i o n 235U r e c y c l e w i t h t h o r i u m ( O p t i o n 2) With 232U removal

31 ,366

805

7.0

29,220

4.8 Not Applicable

1 s t generation

-

*Feed and c e n t r i f u g e s needed t o produce 100

kg U/yr

o f 90% f i s s i l e p r o d u c t .

II

L*

c


L U

3-33

1.

O f t h e f u e l mixtures t h a t may be i n commerce i n a thorium-based f u e l cycle, 20% 2 3 % mixed with thorium i s t h e most d e s i r a b l e both i n ease o f enrichment and because i t r e q u i r e s d i v e r s i o n o f t h e fewest f u e l assemblies t o produce a given q u a n t i t y o f

1

h i g h l y enriched uranium.

h

2.

Enrichment o f 233U r e c y c l e fuels, w i t h o u t 232U removal, i s an enrichment task comparable (with respect t o t h e number o f c e n t r i f u g e s ) t o e n r i c h i n g 20% 235U. The product, however, w i l l be h i g h l y radioactive.

i

t

3.

Ifwould-be p r o l i f e r a t o r s must remove t h e 232U, t h e 235U makeup f u e l s are l e s s p r o l i f e r a t i o n r e s i s t a n t than t h e 233U makeup fuels.

!

id

L

4.

The 235U r e c y c l e f u e l s w i t h thorium and 232U removal are equivalent t o 3.2 w t % s l i g h t l y enriched uranium f u e l s w i t h respect t o both t h e number o f c e n t r i f u g e s and t h e number o f f u e l assemblies t o be diverted.

5.

The 233U r e c y c l e f u e l s w i t h thorium and 232U removal are equivalent t o n a t u r a l uranium enrichment with respect t o t h e number o f centrifuges.

id

L 1

i 1;

t L

1

t

6.

I f 232U removal i s necessary f o r ease o f weapon manufacture and r e l i a b i l i t y o f d e l i v e r y ,

then a d i v e r t e r would probably p r e f e r t o d i v e r t e i t h e r s l i g h t l y enriched uranium f u e l o r e n r i c h n a t u r a l uranium than t o e n r i c h e i t h e r 235U o r 233U r e c y c l e f u e l from thorium cycles. This conclusion r e s u l t s from t h e f a c t t h a t f o r each r e c y c l e f u e l , t h e corresponding s l i g h t l y enriched o r n a t u r a l uranium f u e l enrichment p l a n t requires approximately t h e same number o f c e n t r i f u g e s b u t has t h e decided advantage o f a nonradioactive f a c i l i t y .

As a f i n a l item, t h e average c e n t r i f u g e f a i l u r e r a t e and i t s impact on t h e m a i n t a i n a b i l i t y and production r a t e o f a c e n t r i f u g e enrichment p l a n t must be considered. Information on t h e r e l i a b i l i t y and operating l i f e o f c e n t r i f u g e s i s scarce. The URENCO-CENTEC organization has over t h e years made claims o f very R e l i a b i l i t y o f Centrifuge Enrichment Plants.

long average operating l i f e and correspondingly low f a i l u r e rates.

Typical examples o f

these claims can be found i n some o f t h e i r sales brochures.13 These c l a i m an average 10-year operating l i f e and a f a i l u r e r a t e o f l e s s than 0.5%/year. It i s n o t c l e a r how much p e r i o d i c maintenance (e.g.,

o i l changes and bearing inspection) i s r e q u i r e d t o achieve these low

f a i l u r e rates.

I f these claims are accepted as a goal o f a long-term development p r o j e c t , then i t can be assumed t h a t i n t h e e a r l y p a r t of t h e development somewhat higher f a i l u r e r a t e s would occur, perhaps g r e a t e r by a f a c t o r o f 10. This f a c t o r might be f u r t h e r j u s t i f i e d i n a h i g h l y r a d i o a c t i v e p l a n t since p e r i o d i c maintenance would n o t be p r a c t i c a l . The e f f e c t o f c e n t r i f u g e f a i l u r e s on t h e production r a t e i n a r a d i o a c t i v e p l a n t

1 bi

has n o t been determined; however, some q u a l i t a t i v e statements can be made.

i'

i s o l a t e d from t h e r e s t o f t h e p l a n t .

L

All c e n t r i -

fuge p l a n t s must be designed so t h a t f a i l e d u n i t s o r groups o f u n i t s can be immediately

I t should a l s o be possible, f o r a s p e c i f i c cascade layout, an assumed f a i l u r e rate, and a s p e c i f i e d p l a n t operating l i f e , t o provide


3-34

s t a t i s t i c a l redundancy throughout t h e p l a n t , so t h a t as u n i t s f a i l a new u n i t i s a v a i l able t o be started. Thus, t h e production r a t e could be maintained f o r t h e chosen time p e r i o d w i t h i n t h e assumed s t a t i s t i c a l r e l i a b i l i t y .

I n order t o achieve t h i s r e l i a b i l i t y , g r e a t e r numbers o f c e n t r i f u g e s than l i s t e d i n Table 3.3-9 would be required. The exact number would be determinable when t h e above parameters are s p e c i f i e d . Chemical Extractions from Spent Fuel

As pointed o u t i n t h e i n t r o d u c t i o n t o t h i s section, another p o s s i b i l i t y f o r o b t a i n i n g f i s s i o n a b l e m a t e r i a l from d i v e r t e d denatured 233U f u e l i s through t h e chemical e x t r a c t i o n of protactinium o r plutonium from spent f u e l elements. 233Pa i s an intermediate i s o t o p e i n t h e decay chain leading from 232Th t o 233U t h a t would be chemically separable from t h e uranium p r i o r t o i t s decay. The plutonium a v a i l a b l e i n t h e f u e l elements would be t h a t produced i n t h e 238U denaturant o f t h e f u e l elements. The technical p o s s i b i l i t y o f producing pure 233U v i a chemical e x t r a c t i o n o f 233Pa (t = 27.4 days) from spent denatured f u e l was suggested by Wymer.14

4

Subsequent decay o f

t h e p r o t a c t i n i u m would produce pure 233U. While such a process i s t e c h n i c a l l y f e a s i b l e , c e r t a i n p r a c t i c a l c o n s t r a i n t s must be considered. It i s estimated15 t h a t t h e e q u i l i b r i u m c y c l e discharge o f a denatured LWR would c o n t a i n %34 k g o f 233Pa [approximately 1 kg/metric However, due t o i t s 27.4-day h a l f - l i f e , a 1-MT/day reprocessing capt o n of heavy metal]. a b i l i t y could recover o n l y s23 kg o f 233Pa (beginning immediately upon discharge w i t h a 100% 233pa e f f i c i e n c y ) . Presumably a d i v e r t e r group/nation choosing t h i s r o u t e would have access t o a r e processing f a c i l i t y . Under r o u t i n e operations, spent f u e l elements are u s u a l l y allowed a cool-down p e r i o d o f a t l e a s t 120 days t o permit t h e decay o f s h o r t - l i v e d f i s s i o n products, b u t i n order t o o b t a i n t h e maximum q u a n t i t y o f 233Pa from t h e denatured f u e l s i t would be necessary t o process t h e f u e l s h o r t l y a f t e r i t s discharge from t h e reactor. This would i n v o l v e handling m a t e r i a l s g i v i n g o f f intense r a d i a t i o n s and would probably i n v o l v e an On t h e o t h e r hand, con-

upgrading o f t h e reprocessing f a c i l i t y , e s p e c i a l l y i t s shielding.

ventional reprocessing p l a n t s i n general already have high-performance s h i e l d s and i n c r e mental increases i n t h e dose r a t e s would n o t be unmangeable, e s p e c i a l l y f o r dedicated groups who were n o t averse t o r e c e i v i n g r e l a t i v e l y h i g h exposures. Other problems r e q u i r i n g a t t e n t i o n b u t nevertheless solvable would be associated w i t h upgrading t h e system f o r c o n t r o l l i n g r a d i o a c t i v e off-gases, making allowances f o r some degradation o f t h e organic solvent due t o t h e h i g h r a d i a t i o n l e v e l , and o b t a i n i n g shipping casks w i t h p r o v i s i o n s f o r r e c i r c u l a t i o n of t h e coolant t o a r a d i a t o r . While from t h e above i t would appear t h a t e x t r a c t i o n o f 233Pa would be possible, considerably more f i s s i l e m a t e r i a l could be obtained by e x t r a c t i n g plutonium from t h e spent denatured elements.

Moreover, t h e usual cool-down p e r i o d probably could be allowed, which would r e q u i r e l e s s upgrading o f t h e reprocessing f a c i l i t y . On t h e o t h e r hand, t h e amount o f plutonium obtained from t h e denatured elements would be considerably l e s s (approximately a f a c t o r o f 3 l e s s ) than t h e amount t h a t could be obtained by s e i z i n g and reprocessing spent LEU elements which are already stored i n numerous countries. Thus i t seems u n l i k e l y t h a t a nation/ group would choose t o e x t r a c t e i t h e r 233Pa o r Pu from seized spent denatured f u e l elements.


3-35

3.3.5.

Deterrence .Value o f 232U Contamination i n Denatured Fuel

L

C. M. Newstead Brookhaven National Laboratory

e

The preceding sections have emphasized t h a t unless 232U i s i s o t o p i c a l l y separated from 233U, b o t h i t and i t s daughter products w i l l always e x i s t as a contaminant of t h e f i s s i l e fuel.

And since as 232U decays t o s t a b l e 208Pb i t h e daughter products emit several

h i g h - i n t e n s i t y gamma rays (see Fig. 3.0-l), a l l recent p u r i f i c a t i o n , w i l l be h i g h l y radioactive.

f u e l , except t h a t which has undergone While t h e gama rays, and t o a l e s s e r

23%

e x t e n t t h e decay alpha and beta p a r t i c l e s and t h e neutrons from a,n reactions, w i l l i n t r o duce complications i n t o t h e f u e l cycle, they w i l l a i s 0 serve as a d e t e r r e n t t o t h e seizure o f t h e f u e l and i t s subsequent use i n t h e f a b r i c a t i o n o f a clandestine nuclear explosive. Consider, f o r example, t h e steps t h a t would have t o be followed i n producing and using such a device: 1.

D i v e r t i n g o r s e i z i n g t h e f i s s i l e m a t e r i a l (as r e a c t o r f u e l elements o r as b u l k material).

2.

a.

Chemically reprocessing t h e spent f u e l t o separate o u t t h e bred f i s s i l e p l u tonium ( o r 233Pa) o r

b.

I s o t o p i c a l l y e n r i c h i n g t h e f r e s h f u e l o r b u l k m a t e r i a l t o increase t h e 233U conc e n t r a t i o n i n uranium s u f f i c i e n t l y f o r i t s use i n a weapon.

I

3.

b

t

F a b r i c a t i n g t h e f i s s i l e m a t e r i a l i n t o a c o n f i g u r a t i o n s u i t a b l e f o r an explosive device.

4.

Arming and d e l i v e r i n g t h e device. As indicated, a t Step 2 a decision must be made as t o which f i s s i l e m a t e r i a l i s t o be

employed, 239Pu o r 233U.

E x t r a c t i n g t h e plutonium present i n spent denatured f u e l would

r e q u i r e a chemical separation c a p a b i l i t y analogous t o t h a t r e q u i r e d f o r c u r r e n t LEU spent kilograms o f heavy metal) t h a t f u e l ; however, t h e quantity o f spent denatured f u e l (i.e., would have t o be processed t o o b t a i n a s u f f i c i e n t amount o f 239Pu would be increased by a f a c t o r o f 2 t o 3 over t h e amount o f LEU f u e l t h a t would have t o be processed. Moreover, f o r some r e a c t o r systems, t h e quaZity (i,e.,

ki I

t t

t h e f r a c t i o n o f t h e m a t e r i a l which i s f i s s i l e )

o f t h e plutonium recovered from denatured f u e l would be somewhat degraded r e l a t i v e t o t h e LEU cycle. The s e l e c t i o n o f 233U as t h e weapons f i s s i l e m a t e r i a l means, o f course, t h a t t h e m a t e r i a l being processed through a l l t h e operations l i s t e d above would be radioactive.

While

both n a t i o n a l and subnational groups would be i n h i b i t e d t o some degree by t h e r a d i a t i o n f i e l d , i t i s c l e a r t h a t a n a t i o n a l group would be more l i k e l y t o have t h e resources and technological base necessary t o overcome t h e r a d i a t i o n hazard v i a remote hand1 ing, shielding, and various cleanup techniques. Thus, t h e r a d i a t i o n f i e l d due t o t h e 232U contamination

&d

would be e f f e c t i v e i n l i m i t i n g p r o l i f e r a t i o n by a n a t i o n t o t h e e x t e n t t h a t i t would com-

iâ&#x20AC;?

p l i c a t e t h e procedures which t h e n a t i o n would have t o f o l l o w i n employing t h i s path and


L

3-36

introduce time, c o s t and v i s i b i l i t y considerations.

These f a c t o r s would f o r c e a

t r a d e - o f f between t h e d e s i r a b i l i t y o f u t i l i z i n g m a t e r i a l from t h e denatured f u e l c y c l e and o b t a i n i n g f i s s i l e m a t e r i a l by some o t h e r means, such as i s o t o p i c a l l y e n r i c h i n g n a t u r a l uranium o r producing plutonium i n a research reactor. A subnational group, on t h e o t h e r hand, would n o t i n general possess t h e r e q u i s i t e

technological c a p a b i l i t y .

I n addition, w h i l e a n a t i o n could, i f they chose to, c a r r y o u t

L

c

these processes o v e r t l y , a subnational group would have t o f u n c t i o n c o v e r t l y . Thus the r a d i a t i o n b a r r i e r interposed by t h e s e l f - s p i k i n g e f f e c t o f t h e 232U contaminant i n the denatured f u e l would c o n t r i b u t e i n some measure t o t h e s a f e g u a r d a b i l i t y o f t h e denatured f u e l c y c l e i n s o f a r as t h e subnational t h r e a t i s concerned.

The degree o f p r o t e c t i o n provided by t h e s e l f - s p i k i n g o f denatured fuel v a r i e s accordi n g t o t h e r a d i a t i o n l e v e l . The r a d i a t i o n l e v e l i n t u r n depends on both t h e 232U concentrat i o n and t h e time elapsed a f t e r t h e decay daughters have been chemically separated. As i n d i c a t e d i n o t h e r sections of t h i s chapter, i n denatured f u e l t h e expected concentrations o f 232U i n uranium are expected t o range from $100 t o 300 ppm f o r thermal systems up t o %1600 ppm f o r recycled f a s t r e a c t o r f u e l . I t should be noted t h a t i f t h e l a t t e r denatured f u e l ( t y p i c a l l y 10-20% 233U i n 238U) i s processed i n an enrichment f a c i l i t y t o o b t a i n h i g h l y enriched (~90%)uranium, t h e r e s u l t i n g m a t e r i a l would have a 232U content t h a t i s proport i o n a l l y higher, i n t h i s case $7000 t o 8000 ppm maximum.

Table 3.3-9 shows t h e r a d i a t i o n l e v e l s t o be expected from various concentrations o f 232U a t a number o f times a f t e r t h e uranium has been separated from o t h e r elements i n a chemical processing plant. For a 5-kg sphere o f 233U w i t h 5000 ppm o f 232U t h e r a d i a t i o n l e v e l 232 days a f t e r chemical separation i s 67 r per hour a t l m. The h i g h e s t l e v e l of deterrence, o f course, i s provided when the r a d i a t i o n l e v e l i s i n c a p a c i t a t f n g . Table 3.3-10 describes t h e e f f e c t s on i n d i v i d u a l s o f various t o t a l body doses o f gamma rays. Complete i n c a p a c i t a t i o n r e q u i r e s a t l e a s t 10,000 rem. Beginning a t about 5000 rem t h e dose i s s u f f i c i e n t t o cause death w i t h i n about 48 hr. I n t h e 1000-rem range, death i s p r a c t i c a l l y c e r t a i n w i t h i n a week o r two. A dose causing 50% o f those exposed t o d i e w i t h i n several weeks (an LD-50) i s around 500 rem.

Below 100 rem i t i s u n l i k e l y t h a t any

s i d e e f f e c t s w i l l appear i n t h e s h o r t term b u t delayed e f f e c t s may occur i n t h e long term. I n general, t h e gamma-ray t o t a l dose l e v e l s r e q u i r e d t o ensure t h a t an i n d i v i d u a l i s d i s abled w i t h i n an hour o r so are a t l e a s t on t h e order o f a magnitude higher than those l i k e l y to'cause e v e W death. There may be i n d i v i d u a l s who are w i l l i n g t o accept doses i n excess o f several hundred rem and thus e v e n t u a l l y s a c r i f i c e t h e i r l i v e s . As i n d i c a t e d above, t o stop persons o f s u i c i d a l dedication from completing t h e operations would r e q u i r e doses i n t h e 10,000-rem range. Apart from the dedicated few, however, most i n d i v i d u a l s would be deterred by t h e prospect o f long-term effects from 100-rem levels. However, i t

L I]'

i s a l s o important t o note t h a t t h e i n d i v i d u a l s i n v o l v e d i n t h e actual physical operations may n o t be informed as t o t h e presence o f o r t h e e f f e c t s o f t h e r a d i a t i o n f i e l d .

C'

CJ


3-37

Table 3.3-9.

of

Gama-Ray Dose Rates a t a Distance of 1 m from a 5-kg Sphere 23% Containing Various Concentrations of 2 s U a

Timeb (days)

100 ppmC 0 1.6~10-~ 4.3~10' 3. 5x101 1. 1x102 2. 6x102 5. 5x102 1. 3x103

0 0.116 3.5 10 23 46 93 232

1

b

Dose Rate a t 1 m (mr/hr) 500 ppm 1000 ppm 0 8x10e4 2. l x l o l 1.8x102 5. 7x102 1.3x103 2.8~10~ 6. 7x103

0 1.6~10-~ 4. 3x101 3. 5x102 l.lxl0 3 2. 6x103 5. 5x103 1.3~10~

5000 ppm 0

8~10-~ 2. 1x102 1.8~10~ 5. 7x103 1.3x104 2. 8x104 6. 7x104

aFrom Ref. 16. bTime a f t e r separation. 'Concentration of 2320.

i Table 3.3-10.

L

Total Body Dose (rem) <

i:

25 25-100 100-200 200-600

600- 1,000 1,000-5,000 5,000- 10,000

t

10,000-50,000

t

LJ E

b

a

From Ref. 17.

Effects of Various Total Body Doses of Gamma Rays on Individualsa

.

Effects

No 1i kely acute health effects. No acute e f f e c t s other than temporary blood changes. Some discomfort and fatigue, b u t no major disabling effects; chances of recovery excellent. Entering lethal range (LD-50 ? 500 rads); death may occur w i t h i n several weeks; some sporadic, perhaps temporary disab1 i n g effects wi 11 occur (nausea, vomiting, diarrhea) w i t h i n hour or two a f t e r exposure; however, effects are unlikely t o be completely disabling i n f i r s t few hours. Same as above, except t h a t death within 4-6 weeks i s highly probable. Death within week or two is practically certain; disabling effects within few hours of exposure will be more severe than above, but only sporadically disabling. Death will occur w i t h i n about 48 hr; even i f delivered i n less than one hour, dose will not cause h i g h d i s a b i l i t y f o r several hours, except f o r sporadic intense vomiting and diarrhea; convulsing and ataxia will be likely a f t e r several hours. Death will occur w i t h i n a few hours o r l e s s , with complete Incapacitation w i t h i n minutes i f dose is dellvered within that short period.


L

3-38

An a d d i t i o n a l f a c t o r r e l a t i v e t o t h e d e t e r r e n t e f f e c t i s t h e time r e q u i r e d t o c a r r y o u t the necessary operations. This i s i l l u s t r a t e d by Table 3.3-11, which gives t h e dose r a t e s ( i n rem/hr) r e q u i r e d t o acquire each o f t h r e e t o t a l doses w i t h i n various times, varying from a t o t a l l y i n c a p a c i t a t i n g 20,000 rem t o a prudent i n d i v i d u a l ' s dose o f 100 rem. Thus, t o d i v e r t a small amount o f f i s s i l e m a t e r i a l t o a portable, shielded container might take l e s s than 10 seconds, i n which case a dose r a t e o f l o 7 rem/hr would be r e q u i r e d t o prevent completion o f t h e t r a n s f e r .

Only 200 rem/hr would be required, on t h e o t h e r

hand, t o d e l i v e r a l e t h a l dose t o someone who spends f i v e hours c l o s e t o unshielded 233U w h i l e performing t h e complex operations r e q u i r e d t o f a b r i c a t e components f o r an explosive device.

The maximum a n t i c i p a t e d concentration o f 232U as p r o j e c t e d f o r denatured f u e l does n o t provide s u f f i c i e n t i n t e n s i t y t o reach t o t a l l y d i s a b l i n g l e v e l s . Fast-reactor bred m a t e r i a l (depending on time a f t e r separation and q u a n t i t y as w e l l as 232U concentrat i o n ) can come w i t h i n t h e lOO-rem/hr range. Table 3.3-11.

Gama-Ray Dose Rates f o r Three Levels o f Total Dose vs. Exposure Timea Dose Rate (rem/hr) Required t o D e l i v e r T o t a l Dose o f 100 rem

1000 rem

20,000 rem

10 sec 1 min

36,000 6,000

360,000 60 ,000

7,400,000

5 min

1,200

12,000

1,200,000 240,000

30 min

200

2,000

40,000

1 hr

100

1,000

20,000

5 hr

20

200 83

4,000

Time o f Exposure

8.3

12 h r aFrom Ref.

1,660

18.

The f a c t t h a t t h e l e v e l o f r a d i a t i o n o f 232U-contaminated 233U increases w i t h time i s a major disadvantage f o r a 233U-based nuclear explosive device.

There i s a window o f

10 t o 20 days immediately f o l l o w i n g chemical separation when the m a t e r i a l i s comparatively i n a c t i v e due t o t h e removal of 228Th and i t s daughters. Having t o d e l i v e r a device l e s s than t e n days a f t e r f a b r i c a t i n g i t would be undesirable. While t h e tamper would provide some shielding, t h i s s h o r t time schedule would complicate the s i t u a t i o n considerably.

For a n a t i o n a l program i t i s l i k e l y t h a t the m i l i t a r y would want a clean 233U weapon. This could be accomplished t o a l a r g e degree by separating t h e 232U from t h e 233U using gas c e n t r i f u g a t i o n . However, because t h e masses are o n l y 1 amu a p a r t t h i s requires several thousand c e n t r i f u g e s t o make 100 kg o f clean m a t e r i a l per y e a r (see Sect i o n 3.4.4).

A n a t i o n possessing t h i s i s o t o p i c separation c a p a b i l i t y would t h e r e f o r e prob-

a b l y choose t o e n r i c h n a t u r a l uranium r a t h e r than t o u t i l i z e denatured f u e l , thus e l i m i n a t i n g t h e 232U-induced complications.

L


3-39

I n summary, f o r t h e case o f n a t i o n a l p r o l i f e r a t i o n , t h e intense gamma-ray f i e l d associated w i t h t h e 232U i m p u r i t y would n o t provide any absolute protection. However, t h e presence o f 233U and i t s decay daughters would complicate weapons production s u f f i c i e n t l y

so t h a t t h e n a t i o n might w e l l p r e f e r an a l t e r n a t e source o f f i s s i l e material.

L

For t h e case

o f subnational p r o l i f e r a t i o n , t h e intense gamna-ray f i e l d i s expected t o be a major deterrent. References f o r Section 3.3

i-

1.

R. E. Brooksbank, J. P. Nichols, and A. L. L o t t s , "The Impact o f K i l o r o d Operational Experience on t h e Design o f F a b r i c a t i o n Plants f o r 233U-Th Fuels," pp. 321-340 i n Proceedings o f Second I n t e r n a t i o n a l Thorium Fuel Cycle Symposium, Gat1 inburg, Tennessee, May 3-6, 1966.

2.

D r a f t Environmental Statement, "Light-Water Breeder Reactor Program" ERDA-1541.

3.

J. E. Rushton J. D. Jenkins, and S. R. McNeany, 'Nondestructive Assay Techniques f o r Recycled 133U Fuel f o r High-Temperature Gas-Cooled Reactors ,I' J. Institute Nuclear Materials Nanagement Iv( 1 ) (1975).

4.

T. E. Sampson and P. E. Fehlan, "Sodium Iodide and P l a s t i c S c i n t i l l a t o r Doorway Monitor Response t o Shielded Reactor Grade Plutonium" UC-15, Los Alamos S c i e n t i f i c Laboratory (1 976).

5.

N. L. Shapiro, J. R. Rec, R. A. Matzie, "Assessment o f Thorium Fuel Cycles i n Pressurized-Water Reactors ,'I ERR1 NP-359, Combustion Engineering (1 977).

L

6.

P r i v a t e comnunication from T. J. Burns t o P. R. Kasten, Oak Ridge National Laboratory, September 2, 1977.

i-

7.

Nuclear Engineering I n t e r n a t i o n a l , p. 10 (June 1977).

8.

Nucleonics Week, p. 10 (June 16, 1977).

9.

E. B. Kiser, Jr., "Review o f U.S. Gas Centrifuge Program," AIF Fuel Cycle Conf. '77, Kansas City, Mo. ( A p r i l 1977).

10.

A. de l a Garza. "A Generalization o f t h e Matched Abundance-Ratio Cascade f o r Multicomponent Isotope Separation,' Chemical Eng. Sci. 18, pp. 73-83 (1963).

11.

E. D. Arnold, ORGDP, p r i v a t e communication (August 5, 1977).

12.

USAEC, "Uranium Hexafluoride S p e c i f i c a t i o n Studies,"

13.

URENCO-CENTEC, "Organization and Services" (June 1976).

14.

"Report t o t h e LMFBR Steering Committee on Resources Fuel, and Fuel Cycles, and P r o l i f e r a t i o n Aspects," ERDA-72-60 ( A p r i l 1977).

1'5.

N. L. Shapiro, J. R. Rec, and R. A. Matzie (Combustion Engineering), "Assessment o f Thorium Fuel Cycles i n Pressurized-Water Reactors," EPRI NP-359 (February 1977).

16.

From c a l c u l a t i o n s by E. D. Arnold i n e a r l y 1990's a t Oak Ridge National Laboratory.

17.

From The E f f e c t s o f Nuclear Weapons, Revised E d i t i o n , p. 592, Samuel Glasstone, E d i t o r , prepared by U.S. Department o f Defense, published by U.S. Atomic Energy Commission ( A p r i l , 1962; r e p r i n t e d February, 1964).

18.

E. 0. Weinstock, "A Study on t h e E f f e c t o f Spiking on Special Nuclear Materials," Submitted t o Nuclear Regulatory Commission by Brookhaven National Laboratory (1976).

L

I

bi

b

L 5 ij

Y

0R0-656 ( J u l y 12, 1967).


CHAPTER 4 IMPACT OF DENATURED 233U FUEL ON REACTOR PERFORMANCE

I

b

t Chapter Out1i n e

I

t

L 1;

L t

ki

4.0. 4.1.

Introduction, L.

Abbott, T . J . Burns, and J . C . C l e v e l a n d , ORNL

Light-Water Reactors, J . 4.1.1. 4.1.2.

4.2.

S.

c.

C l e v e l a n d , ORNL

Pressurized Water Reactors B o i l i n g Water Reactors

Spectral-Shift-Controlled Reactors,

N . L . S h a p i r o , CE

4.3.

Heavy-Water Reactors, Y. I. chang, ANL

4.4.

Gas-Cooled Thermal Reactors, J . c . C l e v e l a n d , OWL 4.4.1. High-Temperature Gas-Cooled Reactors 4.4.2. Pebble-Bed High-Temperature Reactors

4.5.

Liquid-Metal Fast Breeder Reactors, T . J . Burns, ORNL

4.6.

A l t e r n a t e Fast Reactors 4.6.1. 4.6.2. 4.6.3.

Advanced Oxide-Fueled LMFBRs, T . J . Burns, ORNL Carbide- and Metal-Fueled LMFBRs, D. L . S e l b y , P . M . Haas, and H. E . Knee, ORNL Gas-Cooled Fast Breeder Reactors, T . 3 . Burns, ORNL


c

I

'


4-3

4.0.

L L __ {

i

J _\

I

b

INTRODUCTION

L. S. Abbott, T. J. Burns, and J. C. Cleveland Oak Ridge National Laboratory The t h r e e preceding chapters have introduced the concept o f 233U f u e l and i t s use i n nuclear power systems t h a t i n c l u d e secure (guarded) energy centers supporting dispersed power reactors, t h e r a t i o n a l e f o r such systems being t h a t they would a l l o w f o r t h e production and use o f f i s s i l e m a t e r i a l i n a manner t h a t would reduce weapons p r o l i f e r a t i o n r i s k s r e l a t i v e t o power systems t h a t a r e i n c r e a s i n g l y based on plutonium-fueled reactors.

Throughout t h e

discussion i t has been assumed t h a t t h e use o f denatured 233U f u e l i n power reactors i s feasible; however, up t o t h i s p o i n t the v a l i d i t y o f t h a t assumption has n o t been addressed.

B

u

A number o f c a l c u l a t i o n s have been performed by various organizations t o estimate the impact t h a t conversion t o t h e denatured c y c l e (and a l s o t o o t h e r " a l t e r n a t e " f u e l cycles) would have on power reactors, using as models both e x i s t i n g reactors and reactors whose

il

c

_.

L i

designs have progressed t o t h e extent t h a t they could be deployed before o r s h o r t l y a f t e r t h e t u r n o f t h e century.

This chapter presents p e r t i n e n t r e s u l t s from these c a l c u l a t i o n s

which, together w i t h t h e p r e d i c t i o n s given i n Chapter 5 on t h e a v a i l a b i l i t y o f t h e various reactors and t h e i r associated f u e l cycles, have been used t o p o s t u l a t e s p e c i f i c symbiotic nuclear power systems u t i l i z i n g denatured f u e l . The adequacy o f such systems f o r meeting p r o j e c t e d e l e c t r i c a l energy demands i s then the subject o f Chapter 6. The impact o f an a l t e r n a t e f u e l c y c l e on t h e performance o f a r e a c t o r w i l l , o f course, be r e a c t o r s p e c i f i c and w i l l l a r g e l y be determined by the d i f f e r e n c e s between t h e

b

neutronic p r o p e r t i e s of t h e f i s s i l e and f e r t i l e nuclides included i n t h e a l t e r n a t e c y c l e and those included i n the r e a c t o r ' s reference cycle. I n t h e case o f t h e proposed denatured

L

f u e l , t h e f i s s i l e n u c l i d e i s 233U and t h e primary f e r t i l e n u c l i d e i s 232Th, w i t h f e r t i l e

Li

238U i n c l u d e d as t h e 233U denaturant. I f LWRs such as those c u r r e n t l y p r o v i d i n g nuclear power i n t h e United States were t o be t h e reactors i n which t h e denatured f u e l i s deployed, then t h e performance of t h e reactors using t h e denatured f u e l must be compared with t h e i r performance using a fuel comprised of t h e f i s s i l e n u c l i d e 23% and t h e f e r t i l e isotope 238U. And since t h e use of 233U assumes recycle, then t h e performance o f t h e LWRs using denatured f u e l must a l s o be compared w i t h LWRs i n which Pu i s recycled. S i m i l a r l y , i f FBRs were t o be t h e r e a c t o r s i n which the denatured f u e l i s deployed, then t h e performance o f FBRs operating on 233U/238U o r 233U/238U/232Th and i n c l u d i n g 232Th i n t h e i r blankets must be compared w i t h t h e performance o f FBRs operating on surrounded by a 238U blanket.

.-

i

br

I L

A s i g n i f i c a n t p o h t i n these two examples i s t h a t they represent t h e two generic thema2 and fast and t h a t t h e neutronic p r o p e r t i e s o f t h e types o f power r e a c t o r s f i s s i l e and f e r t i l e nuclides i n a thermal-neutron environment d i f f e r from t h e i r p r o p e r t i e s

--

i n a fast-neutron environment.

--

Thus w h i l e one f i s s i l e m a t e r i a l may be t h e optimum f u e l i n

a' r e a c t o r operating on thermal neutrons (e,g., f o r a r e a c t o r operating on f a s t neutrons (e.g.,

LWRS) i t may be t h e l e a s t d e s i r a b l e f u e l FBRs).


4-4

Table 4.0-1 gives some o f t h e p e r t i n e n t neutronic p r o p e r t i e s o f t h e d i f f e r e n t f i s s i l e nuclides f o r a s p e c i f i c thermal-neutron energy. I n discussing these properties,* i t i s necessary t o d i s t i n g u i s h between the two functions o f a f i s s i l e m a t e r i a l : t h e production o f energy (i.e., power) and the production o f ezcess neutrons which when absorbed by f e r t i l e m a t e r i a l w i l l produce a d d i t i o n a l f i s s i l e f u e l .

Table 4.0-1.

Nuclear Parameters o f t h e P r i n c i p a l F i s s i l e Nuclides

233U, 235U, 239Pu, and 241P~a9b a t Thermal Energy (Neutron Energy = 0.0252 eV, v e l o c i t y = 2200 m/sec) Cross Section (barns) Nuclide 23311 2351)

239Pu 241Pu

â&#x20AC;&#x2DC;a

22 22 1013 2 4 1375 2 9 578 678

uf

742

22 22 23

1007

7

531 580

a

uc

47 98

21

1

271 368

3 8

2

0.089 0.169 0.366 0.365

2 0.002 2 0.002 2 0.004 2 0.009

n

V

2.487 2.423 2.880 2.934

a

2 0.007 2 0.007 5 0.009 2 0.012

2.284 2.072 2.109 2.149

2 0.006 2 0.006 2 0.007 2 0.014

~~

G. C. Hanna e t al., A t o m i c Energ. Rev. 7, 3-92 (1969); f i g u r e s i n t h e referenced a r t i c l e were a l l given t o one a d d i t i o n a l s i g n i f i c a n t f i g u r e ,

b ~ =a uf + uc; a = uc/uf; v = neutrons produced per f i s s i o n ; 11 = neutrons produced per atom

destroyed = v / ( l + a).

The energy-production e f f i c i e n c y o f a f i s s i l e m a t e r i a l i s d i r e c t l y r e l a t e d t o i t s neutron c a p t u r e - t o - f i s s i o n r a t i o (a), t h e smaller the r a t i o t h e g r e a t e r t h e f r a c t i o n of neutron-nuclide i n t e r a c t i o n s t h a t are energy-producing f i s s i o n s . As i n d i c a t e d by Table 4.0-1,

a t thermal energy the value o f a i s s i g n i f i c a n t l y smaller f o r 233U than f o r t h e

o t h e r isotopes, and thus 233U has a g r e a t e r energy-production e f f i c i e n c y than t h e o t h e r isotopes. ( l h e energy released per f i s s i o n d i f f e r s o n l y s l i g h t l y f o r t h e above isotopes.) The neutron-production e f f i c i e n c y o f a f i s s i l e m a t e r i a l i s determined by t h e number o f neutrons produced per atom o f f i s s i l e m a t e r i a l destroyed (n), t h e higher t h e number t h e more t h e neutrons t h a t w i l l be a v a i l a b l e f o r absorption i n f e r t i l e m a t e r i a l . Table 4.0-1 shows t h a t t h e n value f o r 233U i s higher than t h a t f o r any o f t h e o t h e r nuclides, although plutonium would a t f i r s t appear t o be superior since i t produces more neutrons per f i s s i o n The s u p e r i o r i t y o f 233U r e s u l t s from t h e f a c t t h a t a i s lower f o r 233U and 0 = v / ( l + a). (v). Thus a t thermal energies 2 3 3 U both y i e l d s more energy and produces more neutrons p e r atom destroyed than any o f the o t h e r f i s s i l e nuclides.

-

I n the energy range o f i n t e r e s t f o r f a s t reactors ( ~ 0 . 0 5 4.0 MeV), t h e s i t u a t i o n i s n o t q u i t e so straightforward. Here again, the a value f o r 233U i s s i g n i f i c a n t l y lower than t h e values f o r t h e other f i s s i l e nuclides, and, moreover, t h e microscopic cross secThe energy release per f i s s i o n o f 233U i s t i o n f o r f i s s i o n i s higher (see Fig. 4.0-1). somewhat l e s s than t h a t of the plutonium nuclides, b u t the energy release per atom o f 233U destroyed i s s i g n i f i c a n t l y higher than f o r th, o t h c r nuclides. Thus, from t h s standpoint *Much o f t h i s discussion on the neutronic p r o p e r t i e s o f nuclides i s based on r e f s . 1

-

3.


Y 4-5

ORNL-DWG 78-13

I ORNL-DWG 76-17705

3.0 2.8

,

2.6 2.4

9

2.2

2.0 1.8 I ID0

I

aro

I .

I .6 0 I

10.00

I

I

0.I

1.0

-

E(MeV)

ORNL-DNG 7647702

ORNL-DWG76-47704

m

0.04

I

J

aro

+Do

E(MeV)

0

2

6

4

E(MeV)

Fig. 4.0-1. Nuclear Parameters o f the Principal Fissile and Fertile Nuclides a t High Neutron Energies. a = ac/uf; rl = v / ( l + a ) .

10


4-6

..

o f energy-production e f f i c i e n c y , 233U i s c l e a r l y superior f o r f a s t systems as w e l l as f o r thermal systems. However, w i t h t h e h i s t o r i c a l emphasis on f i s s i l e production i n f a s t systems, t h e o v e r r i d i n g consideration i s t h e neutron-production e f f i c i e n c y o f t h e system, and f o r neutron production 239Pu i s superior. This can be deduced from t h e values f o r q given i n Fig. 4.0-1.

The

rl

value f o r 239Pu i s much higher than t h a t f o r the o t h e r nuclides, es-

p e c i a l l y a t t h e higher neutron energies, owing t o the f a c t t h a t 239Pu produces more neutrons p e r f i s s i o n than t h e o t h e r isotopes; t h a t i s , i t has a higher v value, and t h a t value i s ess e n t i a l l y energy-independent.

L

As a r e s u l t , more neutrons are a v a i l a b l e f o r absorption i n

f e r t i l e m a t e r i a l s and 239Pu was o r i g i n a l l y chosen as the f i s s i l e f u e l f o r f a s t breeder reactors. The f i s s i o n p r o p e r t i e s o f the f e r t i l e nuclides are a l s o important since f i s s i o n s i n t h e f e r t i l e elements increase both the energy production and t h e excess neutron production and thereby reduce f u e l demands. A t higher energies, f e r t i l e f i s s i o n s c o n t r i b u t e s i g n i f i cdntly, t h e degree o f t h e c o n t r i b u t i o n depending g r e a t l y on t h e n u c l i d e being used. As shown i n Fig. 4.0-1,

t h e f i s s i o n cross section f o r 232Th i s s i g n i f i c a n t l y lower (by a f a c t o r

o f approximately 4) than t h e f i s s i o n cross s e c t i o n o f 238U. I n a f a s t reactor, t h i s means t h a t w h i l e 15 t o 20% o f the f i s s i o n s i n the system would occur i n 238U, o n l y 4 t o 5% would

Li L

occur i n 232Th. Thus the p a i r e d use o f 233U and 232Th i n a f a s t system would i n c u r a double penalty w i t h respect t o i t s breeding performance. It should be noted, however, t h a t since denatured 233U f u e l would a l s o c o n t a i n 238U (and e v e n t u a l l y 239Pu), the penalty would be somewhat m i t i g a t e d as compared w i t h a system operating on a nondenatured 233U/232Th fuel. I n a thermal system, t h e f a s t f i s s i o n e f f e c t i s l e s s s i g n i f i c a n t due t o t h e smaller f r a c t i o n o f neutrons above the f e r t i l e f a s t f i s s i o n threshold. I n considering t h e impact o f the f e r t i l e nuclides on r e a c t o r performance, i t i s a l s o

L

c

necessary t o compare t h e i r n u c l i d e production chains. Figure 4.0-2 shows t h a t t h e chains are very s i m i l a r i n structure. The f e r t i l e species 232Th and 234U i n t h e thorium chain .corresponding t o 238U and 240Pu i n t'he uranium chain, w h i l e t h e f i s s i l e components 233U and

235U a r e p a i r e d w i t h 239Pu and 241Pu, and f i n a l l y , t h e p a r a s i t i c nuclides 236U and 242Pu complete t h e r e s p e c t i v e chains. A s i g n i f i c a n t d i f f e r e n c e i n t h e two chains l i e s i n t h e nuclear c h a r a c t e r i s t i c s o f the intermediate nuclides 233Pa and 237Np. Because 233Pa has a longer h a l f - l i f e (i .e.

,a

smaller decay constant), intermediate-nuclide captures are more

probable i n t h e thorium cycle.

Such captures are doubly s i g n i f i c a n t since they n o t o n l y

u t i l i z e a neutron t h a t could be used f o r breeding, b u t i n a d d i t i o n e l i m i n a t e a p o t e n t i a l f i s s i l e atom.

L]

A f u r t h e r consideration associated w i t h t h e d i f f e r e n t intermediate nuclides

i s t h e r e a c t i v i t y a d d i t i o n associated w i t h t h e i r decay t o f i s s i l e isotopes f o l l o w i n g r e a c t o r shutdown. Owing t o t h e longer h a l f - l i f e (and correspondingly higher e q u i l i b r i u m i s o t o p i c concentration) o f 233Pa, t h e r e a c t i v i t y a d d i t i o n f o l l o w i n g r e a c t o r shutdown i s h i g h e r f o r thorium-based fuels. Proper consideration o f t h i s e f f e c t i s r e q u i r e d i n t h e design o f t h e r e a c t i v i t y c o n t r o l and shutdown systems. The actual e f f e c t o f a l l these f a c t o r s , o f course, depends on t h e neutron energy spectrum o f t h e p a r t i c u l a r r e a c t o r type and must be addressed on an i n d i v i d u a l r e a c t o r basis. S i g n i f i c a n t differences a l s o e x i s t i n t h e f i s s i o n - p r o d u c t y i e l d s o f 23311 versus 235U, and these, too, must be addressed on an i n d i v i d u a l r e a c t o r basis.

II


t

4-7

..

-.

F

4d

--

I '

L

0- 122, 23 2

Th(n,y)---t

233

Th

Fig. 4.0-2a.

Nuclide Production Chain f o r 232Th.

L _-

I f

hi

L -

I

239Np(n, N 0p42,)y

b

0- 123.5111

-..

1 '

L

L

23*U(n,y)-239U

Fig. 4.0-2b.

Nucl i e Production Chain f o r 23%.

Consideration o f many o f t h e above f a c t o r s i s inherent i n t h e "mass balance" calculat i o n s presented i n t h i s chapter f o r the various reactors operating on a l t e r n a t e f u e l cycles.

It i s emphasized, however, t h a t i f a d e f i n i t e decision were made t o employ a s p e c i f i c a l t e r n a t e fuel c y c l e i n a s p e c i f i c reactor, t h e next step would be t o optimize the r e a c t o r design f o r t h a t p a r t i c u l a r cycle, as i s discussed i n Chapter 5. Optimization o f each r e a c t o r f o r t h e

L I

many f u e l s considered was beyond the scope o f t h i s study, however, and instead t h e design used f o r each r e a c t o r was t h e design f o r t h a t r e a c t o r ' s reference f u e l , regardless o f the f u e l c y c l e under consideration.

-

! *

rw --

t

u

f"

The reactors analyzed i n t h e c a l c u l a t i o n s are l i g h t - w a t e r thermal reactors; spectrals h i f t - c o n t r o l l e d thermal reactors; heavy-water thermal reactors; high-temperature gascooled thermal reactors; l i q u i d - m e t a l f a s t breeder reactors; and f a s t breeder reactors o f advanced o r a l t e r n a t e designs.


4-8

Since w i t h the exception o f the F o r t S t . Vrain HTGR, t h e e x i s t i n g power reactors i n t h e United States are LWRs, i n i t i a l studies o f a l t e r n a t e f u e l cycles have assumed t h a t they would f i r s t be implemented i n LWRs.* Thus t h e c a l c u l a t i o n s f o r LWRs, sunnnarized i n Sect i o n 4.1 have considered a number o f fuels. For t h e purposes o f t h e present study t h e f u e l s have been categorized according t o t h e i r p o t e n t i a l usefulness i n t h e envisioned power system scenarios. Those fuel types t h a t meet t h e n o n p r o l i f e r a t i o n requirements s t a t e d e a r l i e r i n t h i s r e p o r t are c l a s s i f i e d as " d i s p e r s i b l e " f u e l s t h a t could be used i n LWRs operating outs i d e a secure energy center. The d i s p e r s i b l e f u e l s are f u r t h e r d i v i d e d i n t o denatured 233U f u e l s and 235U-based f u e l s .

The remaining f u e l s i n t h e power systems a r e then categorized as

"energy-center-constrained" fuels. F i n a l l y , a f o u r t h category i s used t o i d e n t i f y "reference" fuels. Reference f u e l s , which are n o t t o be confused w i t h an i n d i v i d u a l r e a c t o r ' s reference fuel

, are

b;

fuels t h a t would have no apparent usefulness

in

t h e energy-center,

tl;

dispersed-reactor

scenarios b u t are included as l i m i t i n g cases against which t h e o t h e r f u e l s can be compared. (Note: The r e a c t o r ' s reference fuel may o r may n o t be appropriate f o r use i n t h e reduced p r o l i f e r a t i o n r i s k scenarios.)

To t h e e x t e n t t h a t they apply, these f o u r categories have been used t o c l a s s i f y a l l t h e fuels presented here f o r the various reactors. Although t h e c o n t r i b u t i n g authors have used d i f f e r e n t notations, the f u e l s included are i n general as f o l l o w s : D i s p e r s i b l e Resource-Based Fuels A.

Natural uranium f u e l ( c o n t a i n i n g approximately 0.7% 2 3 5 U ) , CANDU heavy-water reactors.

B.

Low-enriched LWRs.

C.

235U

Notation:

Medium-enriched

Notation:

as c u r r e n t l y used i n

U5(NAT)/U.

f u e l (containing approximately 3% 235U), as c u r r e n t l y used i n LEU; U5(LE)/U.

235U

f u e l (containing approximately 20% 235U) mixed w i t h thorium

f e r t i l e m a t e r i a l ; could serve as a t r a n s i t i o n f u e l p r i o r t o f u l l - s c a l e implementat i o n o f t h e denatured

233U cycle.

Notation:

MEU(235)/Th; DUTH(235).

D i s p e r s i b l e Denatured Fuel

D.

Denatured 233U f u e l (nominally approximately 12% 233U i n U).

Notation:

233U; denatured uranium/thorium; denatured 233U02/Th02; MEU(233)/Th; DUTH(233); U3(DE)/U/Th. '

Denatured

233U/23*U;

*NOTE: The r e s u l t s presented i n t h i s chapter do n o t consider the p o t e n t i a l improvements i n the once-through LWR t h a t are c u r r e n t l y under study. I n general, t h i s i s a l s o t r u e f o r t h e resource-constrained nuclear power systems evaluated i n Chapter 6; however, Chapter 6 does i n c l u d e r e s u l t s from a few c a l c u l a t i o n s f o r an extended exposure (43,000-MWD/MTU) once-through LEU-LWR. The p a r t i c u l a r extended exposure design considered requires 6% l e s s U3O8 over t h e r e a c t o r ' s l i f e t i m e .

c i:

c


4-9

Energ-y-Center-Constrained Fuels

E.

LEU f u e l w i t h plutonium recycle.

F. P u - ~ ~mixed-oxide ~ T ~ fuel. G.

L

Notation:

Pu02/Th02; (Pu-Th)O,;.Pu/Th.

P U - ~ ~mixed-oxide U f u e l , as proposed f o r c u r r e n t l y designed LMFBRs. Pu02/UO,; P u p a u; PU/U.

Reference Fuels

H.

H i g h l y enriched 235U fuel (containing approximately 93% 235U) mixed w i t h thorium f e r t i l e m a t e r i a l , as c u r r e n t l y used i n HTGRs.

L;

Notation:

I.

Notation:

HEU(235)/Th; US(HE)/Th.

Highly enriched 233U f u e l (containing approximately 90% 233U) mixed w i t h thorium f e r t i l e material.

Notation:

HE(233)/Th;

U3/Th; US(HE)/Th.

I n c l u d i n g plutonium-fueled r e a c t o r s w i t h i n t h e energy centers serves a t w o - f o l d purpose: It provides a means f o r disposing o f the plutonium produced i n the dispersed reactors, and

i t provides f o r an exogeneous source o f 233U.

L

The discussion of LWRs operating on these various f u e l cycles presented i n Section

-

4.1 i s followed by s i m i l a r treatments o f t h e o t h e r reactors i n Sections 4.2 4.6. The f i r s t , t h e Spectral-Shift-Controlled Reactor (SSCR), i s a m o d i f i e d PWR whose operation on a LEU c y c l e has been under study by both t h e United States and Belgium f o r more than a

t - I

decade.

The primary goal of t h e system i s t o improve f u e l u t i l i z a t i o n through t h e i n f creased production and i n - s i t u consumption of f i s s i l e plutonium (Pu ). The capture o f neut r o n s i n t h e 238U i n c l u d e d i n t h e fuel elements i s increased by mixing heavy water w i t h

I '

b

t h e l i g h t - w a t e r moderator-coolant, thereby s h i f t i n g t h e neutron spectrum w i t h i n the core t o energies a t which neutron absorption i n 2 3 8 ~i s more l i k e l y t o occur. The heavy water content i n t h e moderator i s decreased d u r i n g t h e c y c l e as f u e l r e a c t i v i t y i s depleted. The

L

increased capture i s a l s o used as t h e r e a c t o r c o n t r o l mechanism. The SSCR i s one o f a c l a s s o f r e a c t o r s t h a t are i n c r e a s i n g l y being r e f e r r e d t o as advanced converters, a term a p p l i e d t o a thermal r e a c t o r whose design has been m o d i f i e d t o increase i t s production o f f i s s i l e material. Heavy-water-modi f i e d thermal r e a c t o r s a r e represented here by Canada's n a t u r a l uranium-fueled CANDUs. L i k e t h e SSCR, t h e CANDU has been under study i n t h e U.S. as an advanced converter, and scoping c a l c u l a t i o n s have been performed f o r several f u e l cycles,

-t i b --

I;

i n c l u d i n g a s l i g h t l y enriched 235U c y c l e t h a t i s considered t o be t h e r e a c t o r ' s reference c y c l e f o r implementation i n t h e United States. The high-temperature gas-cooled thermal reactors considered a r e t h e U.S. HTGR and t h e West German Pebble Bed Reactor (PBR), t h e PBR d i f f e r i n g from t h e HTGR i n t h a t i t u t i l i z e s spherical f u e l elements r a t h e r than p r i s m a t i c f u e l elements and employs o n - l i n e refueling.

b

For both reactors t h e reference c y c l e [HEU(233U)/Th]

includes thorium, and s h i f t i n g


4-1 0

t o t h e denatured c y c l e would c o n s i s t i n i t i a l l y i n r e p l a c i n g t h e 93% 235U i n 238U w i t h 15 t o 20% 235U i n 238U.

The HTGR has reached t h e prototype stage a t t h e F o r t V r a i n p l a n t i n

Colorado and a PBR-type r e a c t o r has been generating e l e c t r i c i t y i n West Germany since 1967.

tl

While t h e above thermal reactors show promise as power-producing advanced converters, they w i l l n o t be s e l f - s u f f i c i e n t on any o f t h e proposed a l t e r n a t e f u e l cycles and w i l l req u i r e an exogenous source o f 233U. An e a r l y b u t l i m i t e d q u a n t i t y o f 233U could be provided by i n t r o d u c i n g thorium w i t h i n the cores o f 235U-fueled LWRs, but, as has already been pointed o u t i n t h i s report, f o r t h e long-term, reactors dedicated t o 233U production w i l l be required. f

.

I n the I n t h e envisioned scenarios those reactors p r i m a r i l y w i l l be f u e l e d w i t h Pu c a l c u l a t i o n s presented here a p r i n c i p a l 2 3 % production r e a c t o r i s t h e mixed-oxide-fueled LMFBR containing thorium i n i t s blanket.

I n addition, "advanced LMFBRs" t h a t have

blanket assemblies intermixed w i t h f u e l assemblies are examined.

L

The p o s s i b l e advantages

and disadvantages o f using metal- o r carbide-based LMFBR f u e l assemblies are a l s o discussed. F i n a l l y , some p r e l i m i n a r y c a l c u l a t i o n s f o r a helium-cooled f a s t breeder r e a c t o r (GCFBR) are presented. The consideration o f f a s t reactors t h a t burn one f i s s i l e m a t e r i a l t o produce another has introduced considerable confusion i n r e a c t o r terminology which, unfortunately, been resolved i n t h i s report. I n t h e past, t h e term f a s t breeder has been a p p l i e d f a s t r e a c t o r t h a t breeds enough o f i t s own f u e l t o s u s t a i n i t s e l f . Thus, t h e f a s t t h a t burn 239Pu t o produce 233U are n o t "breeders" i n the t r a d i t i o n a l sense. They

has n o t to a reactors are,

however, producing f u e l a t a r a t e i n excess o f consumption, which i s t o be contrasted w i t h t h e advanced thermal converters whose primary f u n c t i o n i s t o s t r e t c h b u t n o t increase t h e f u e l supply. I n order t o d i s t i n g u i s h t h e P u - ~ o - ~ ~f a~s Ut reactors from others, the term

transmuters was coined a t ORNL. Immediately, however, the word began t o be a p p l i e d t o any r e a c t o r t h a t burns one f u e l and produces another. Moreover, i t soon became obvious t h a t t h e words f a s t and breeder a r e used synonymously.

Thus i n t h i s r e p o r t and elsewhere

we f i n d various combinations o f terms, such as LMFBR transmuter and converter transmuter. The s i t u a t i o n becomes even more complicated when t h e f a s t r e a c t o r design uses both 238U and 232Th i n the blanket, so t h a t i n e f f e c t i t takes on the c h a r a c t e r i s t i c s o f both a transmuter and a breeder. F i n a l l y , t h e reader i s cautioned n o t t o i n f e r t h a t o n l y those reactors discussed i n t h i s chapter dre candidates f o r t h e enerTy-center, dispersed-reactor scenarios. I n f a c t , t h e scenarios discussed i n Chapter 6 do n o t even use a l l these reactors and they c o u l d e a s i l y consider o t h e r r e a c t o r types. The s e l e c t i o n o f reactors f o r t h i s p r e l i m i n a r y assessment o f t h e denatured 233U f u e l c y c l e was based p r i m a r i l y on t h e a v a i l a b i l i t y o f data a t t h e time t h e study was i n i t i a t e d (December, 1977).

L I: L


_-

I

-L

4-1 1 i

d

References f o r Section 4.0

.--

L

1.

P. R. Kasten, F. J. Homan e t al., "Assessment o f t h e Thorium Fuel Cycle i n Power Reactors," ORNL/TM-5665, Oak Ridge National Laboratory (January 1977).

2.

P. R. Kasten, "The Role o f Thorium i n Power-Reactor Development," view, Vol. 111, No. 3.

3.

"The Use o f Thorium i n Nuclear Power Reactors," prepared by D i v i s i o n o f Reactor Development and Technology, U.S. Atomic Energy Conmission, w i t h assistance o f ANL, B&W, BNL, GEA, ORNL, and PNL, WASH 1097 (June, 1969).

Atomic Energy Re-


4-1 2

4.1.

LIGHT-WATER REACTORS

J. C. Cleveland Oak Ridge National Laboratory I f an a l t e r n a t e c y c l e such as t h e denatured c y c l e i s t o have a s i g n i f i c a n t e a r l y impact, i t must be implemented i n LGIRs already operating i n t h e U n i t e d States o r soon t o be operating. The c u r r e n t n a t i o n a l LWR capacity i s about 48 GWe and LWRs t h a t w i l l provide a t o t a l capacity o f 150 t o 200 GWe by 1990 are e i t h e r under c o n s t r u c t i o n o r on order. Much o f t h e i n i t i a l analyses o f t h e denatured

tl

233U f u e l c y c l e has t h e r e f o r e been performed f o r c u r r e n t LWR core

and fuel assembly designs under t h e assumption t h a t subsequent t o t h e r e q u i r e d f u e l s development and demonstration phase f o r t h o r i a f u e l s these f u e l s could be used as r e l o a d f u e l s f o r operating LWRs.

It should be noted, however, t h a t these c u r r e n t LWR designs were optimized t o minimize

power costs w i t h LEU f u e l s and plutonium recycle, and t h e r e f o r e they do n o t represent optimum designs f o r t h e denatured cycle.

Also excluded from t h i s study a r e any improvements i n reac-

t o r design and operating s t r a t e g i e s t h a t would improve i n - s i t u u t i l i z a t i o n o f bred f u e l and reduce t h e nonproductive loss o f neutrons in LWRs operating on t h e once-through cycle. Studies t o consider such improvements have r e c e n t l y been undertaken as p a r t o f NASAP (Nonproliferation

L tj

A1 t e r n a t i v e Systems Assessment Program).

4.1.1.

Pressurized Water Reactors

Mass f l o w c a l c u l a t i o n s f o r PWRs presented i n t h i s chapter were performed p r i m a r i l y by Combustion Engineering, w i t h some a d d i t i o n a l r e s u l t s presented from ORNL c a l c u l a t i o n s . The Combu'stion Engineering System 80TM (PWR) design was used i n a l l o f these analyses.

A

d e s c r i p t i o n o f t h e core and f u e l assembly design i s presented i n t h e Combustion Engineering The f o l l o w i n g cases have been Standard Safety Analysis Report (CESSAR)

.

D i s p e r s i b l e Resource-Based Fuels A.

LEU (i.e.,

B.

MEU/Th (i.e.,

low enriched uranium, -3% 235U i n 238U), no recycle. medium-enriched uranium, 20% 235U i n

238U,

mixed w i t h 232Th),

no recycle. C.

LEU, r e c y c l e o f uranium only, 235U makeup.

D.

MEU/Th, r e c y c l e of uranium (235U + 233U), 20% 235U makeup.*

D i s p e r s i b l e Denatured Fuel E.

Denatured 233U (i,e.,

-12%

233U

in

238U,

mixed w i t h 232Th), r e c y c l e o f uranium,

233U makeup.

L]: *An a l t e r n a t e case u t i l i z i n g 93% 235U as a f i s s i l e topping f o r recovered r e c y c l e uranium and u t i l i z i n g 20% 235U as f r e s h makeup i s a l s o discussed by Combustion Engineering.


_-

4-1 3

id Enerqy-Center-Constrained Fuels

L

F.

LEU, r e c y c l e o f uranium and self-generated plutonium, 235U makeup.

6. H.

P U / ~ ~ ~ rUe c, y c l e o f plutonium, plutonium makeup. P ~ / ~ 3 2 T hr,e c y c l e o f plutonium, plutonium makeup.

I.

P ~ / ~ 3 2 T hone-pass , p l uto-nium, p l u t o n i um. makeup.

Reference Fuel

J. 1

HEU/Th (i.e.,

, mixed w i t h

h i g h l y enriched uranium, 93.15 w/o 235U i n 238U

232Th),

r e c y c l e o f uranium (235U + 233U), 235U makeup. Case A represents t h e c u r r e n t mode o f LWR operation i n t h e absence o f reprocessing.

t I ,

B involves t h e use o f MEU/Th f u e l i n which t h e i n i t i a l uranium enrichment i s l i m i t e d With reprocessing again disallowed, Case B r e f l e c t s a "stowaway" o p t i o n t o 20% 235U/238U.

Case

i n which t h e 233U bred i n t h e f u e l and t h e unburned 235U are reserved f o r f u t u r e u t i l i z a t i o n .

4d

Case C represents one l o g i c a l extension o f Case A f o r t h e cases where t h e r e c y c l e o f c e r t a i n m a t e r i e l s i s allowed. However, consistent w i t h t h e reduced p r o l i f e r a t i o n r i s k

i,

ground r u l e , o n l y t h e uranium component i s recycled back i n t o t h e dispersed reactors.

Case D

s i m i l a r l y r e f l e c t s t h e extension o f Case B t o the r e c y c l e scenario. I n t h i s case, t h e bred plutonium i s assumed t o be separated from t h e spent f u e l b u t i s n o t recycled. MEU(2O% 235U/U)/Th f u e l i s used as makeup m a t e r i a l and i s assumed t o be f a b r i c a t e d i n separate assemblies from

L

t h e r e c y c l e material. Thus, o n l y t h e assemblies containing r e c y c l e m a t e r i a l r e q u i r e remote (It i s assumed t h a t t h e presence o f t h e 232U pref a b r i c a t i o n due t o t h e presence o f 232U. cludes t h e recovered uranium being reenriched by i s o t o p i c separation. )

The recovered uranium

from both t h e r e c y c l e and t h e makeup f u e l f r a c t i o n s a r e mixed together p r i o r t o t h e next recycle. T h i s a d d i t i o n o f a r e l a t i v e l y h i g h q u a l i t y f i s s i l e m a t e r i a l (uranium recovered from t h e makeup f u e l ) t o t h e r e c y c l e f u e l stream slows t h e decrease i n t h e f i s s i l e content of t h e r e c y c l e uranium. As i n t h e LEU cycle, t h e f i s s i l e component o f t h e r e c y c l e f u e l i n t h i s f u e l c y c l e scheme i s d i l u t e d w i t h 23eU which provides a p o t e n t i a l safeguards advantage over t h e conventional concept o f plutonium r e c y c l e i n LWRs w i t h about t h e same

U3O8

u t i 1i z a t i o n . Case E i s t h e denatured 233U'fuel.

I t u t i l i z e s an exogenous source o f 233U f o r both

t h e i n i t i a l core f i s s i l e requirements and t h e f i s s i l e makeup requirements.

id L

5

-

I represent p o s s i b l e f i s s i l e / f e r t i l e f u e l c y c l e systems allowable f o r use Cases F i n secure energy centers. Case F represents an extension o f Case C i n which a l l t h e f i s s i l e

bi

m a t e r i a l present i n t h e spent f u e l , i n c l u d i n g t h e plutonium, i s recycled.

t

which c o n t a i n t h e recycled plutonium i n a uranium d i l u e n t .

hd

t'

Under e q u i l i b r i u m

conditions, about 1/3 o f each r e l o a d f u e l batch consists o f mixed oxide IM02) f u e l assemblies c o n s i s t s o f f r e s h o r recycled uranium (235U) oxide f u e l .

The remaining 2/3 o f each r e l o a d


4-14

Case G allows one possible means for utilizing the plutonium bred in the dispersed reactors. Plutonium discharged from LEU-LWRs is used to provide the initial core fissile requirements as well as the fissile makeup requirements. This plutonium is blended in a U02 diluent consisting of natural or depleted uranium. The plutonium discharged from the UO,/PuO, reactor is continually recycled - with two years for reprocessing and refabrication through the reactor. In the equilibrium condition, plutonium discharged from about 2.7 LEU-fueled LWRs can provide the makeup fissile Pu requirement for one U02/Pu02 LUR.

-

In Case H the Pu02/Th02 LWR also utilizes plutonium discharged from LEU-LWRs to provide the initial core fissile requirements and the fissile makeup requirements. This plutonium is blended in a Tho, diluent. The isotopically degraded plutonium recovered from the PuO2/ThO2 LWR is blended with LEU-LWR discharge plutonium (of a higher fissile content) and recycled back into the Pu02/Th02 LWR. Not only does this case provide a means of eliminating the Pu bred in the dispersed reactors but, in addition, also provides for the production of 233U that can be denatured and used to fuel dispersed reactors. !

The Pu02/Th02 LWR of Case I is similar to that in Case H in that plutonium discharged from LEU-LWRs is used to provide the fissile requirements. However, the isotopically degraded plutonium recovered from the PuO,/ThO, LWR is not recycled into an LWR but is stored for later use in a breeder reactor. Case J involves the use of highly enriched uranium fuel enrichment. The uranium enrichment in HEU fuels was of information in Ref. 7. Initially all fuel consists of equilibrium recycle conditions are achieved, about 35% of makeup fuel, the remaining fuel assemblies in each reload not re-enriched) uranium oxide blended with fresh Tho2.

blended with Tho2 to the desired selected as 93.15 w/o on the basis fresh HEU/Th fuel assemblies. Once the fuel consists of this fresh batch containing the recycled (but

Table 4.1-1 provides a summary, obtained from the detailed mass balance information, of initial loading, equilibrium cycle loading, equilibrium cycle discharge, and 30-year cumulative U308 and separative work requirements. All recycle cases involve a two-year ex-reactor delay for reprocessing and refabrication. It is important to point out that for cases which involve recycle of recovered fissile material back into the same LWR, in "equi 1 ibrium" conditions the makeup requirement for a given recycle generation is greater than the difference between the charge and discharge quantities for the previous recycle generation because of the degradation of the isotopics. This is especially important i n Case H where, for example, the fissile content of the plutonium drops from about 71% to about 47% over an equilibrium cycle. Comparing Cases A and B of Table 4.1-1 indicates the penalties associated with implementation of the MEU/Th cycle relative to the LEU cycle under the restriction of no recycle. The MEU/Th case requires 40% more U308 and 214% more separative work than the LEU

c I: h:


4-15

case,

C l e a r l y t h e MEU/Th c y c l e would be p r o h i b i t i v e f o r "throwaway" options.

f i c a n t r e s u l t from Table 4.1-1

-- .

-

&

I

'

I

h

and Case F, LEU w i t h uranium and self-generated plutonium recycle, The U308 demand i n each case i s t h e same, although t h e MEU/Th c y c l e requires increased separative work. A d d i t i o n a l l y i t should be noted t h a t i n Case D t h e MEU/Th fuel a l s o produces s i g n i f i c a n t q u a n t i t i e s o f

plutonium, an a d d i t i o n a l f i s s i l e m a t e r i a l s t o c k p i l e which i s n o t recycled i n t h i s case.

I '

Table 4.1-1.

Fuel U t i l i z a t i o n Characteristics f o r PWRs Under Various Fuel Cycle Optionsenb

Y

61

A second s i g n i -

i s given by t h e comparison o f Case D, MEU/Th w i t h uranium r e c y c l e

Case

Fuel Type

Initial Fissile Inventory (kq/GWe)

Fissile Charge (kalGWe-vrl

Separative Work U 0 Requirement Requirement Equilibrium Cycle Fissile '(ST/GU;!j 1 0 3 kg SWU/GWe) Discharge Conversion Burnup 30-yr e (ka/GWe-vr) Ratio (MWD/kq HM) I n i t i a l ' Tot%dInitial e Total

Dispersible Resource-Based Fuels

L --

I(

A

LEU. no recycle

1693 235U

B

MEU/Th. no recycle

2538 235U

C D

LEU, U recycle MEU/Th, s e l f generated U recycle

1693 235U 2538 2 s U

794

235u

1079 235U

215 23511 174 Puf 260 23% 384 23% 71 Puf

0.60

30.4

392

5989f

203

3555

0.63

32.6

638

8360

580

7595

0.60 282 23S8 0.66 257 235U8 95 Puf0

30.4 32.6

392 638

4946 4090

203 580

3452 3632

-

313 2 3 3 U 8 675 235U8

Dispersible Denatured Fuel E

L

750 233U 29 235U

446 233U 43 235U 63 Puf

33.4

Energy-Center-Constrained Fuels F G H

L

Denatured 33U02/Th02. 1841 233U 27 235U U recycle (exogenous 33U makeup)

I

LEU. recycle o f U + self-generated Pu

1693 2sU

PuOp/UO2. Pu recycle

1568 Puf 546 235U PuOZ/ThOz. Pu recycle 2407 Puf Pu02/Th02. single Pu pass

2407 Puf

612 258 1153 173 1385

235U Puf Puf 235U Puf

1140 Puf

193 288 858 108 696 272 410 284

235U

Puf Puf 235U Puf

0.61

30.4

392

4089

233

2690

0.63

30.4

100

1053

0

0

597

3453

596

3436

33.0

233U

Puf 23%

33.0

Reference Fuel

J

HEWTh. self-generated 2375 235U U recycle

388 23% 504 235U

377 233U 172 23511

0.67

33.4

:All cases assume 0.2 w/o t a i l s and 75% capacity factor. %11 calculations were p e r f o n e d f o r the mwt, 13WMWe tombustion Engineering System BOm reactor design. 1.0% f a b r i c a t i o n loss and 0.5% conversion loss. o c r e d i t taken f o r end o f reactor l i f e f i s s i l e inventory. 4ssumes ssumes 1.0% f a b r i c a t i o n loss.

9An additional case i s considered I n Chapter 6 i n which an extended exposure (43 MWO/kg HM) LEU-PUR on a once-through cycle results i n a 6% reduction i n the 30-yr t o t a l U 08 requirements, while s t i l l requiring e s s e n t i a l l y the same enrichment (SUU) requirements. Scinewhat less plutonium i s disciarged from the reactor because o f a reduced conversion ratio. Values provided are representatlve o f years 19-23. 'Reference f u e l s are considered only as l i m i t i n g cases.

Differences i n t h e n u c l i d e concentrations of f e r t i l e isotopes from case t o case r e s u l t I n d i f f e r e n c e s i n t h e resonance i n t e g r a l s o f each f e r t i l e isotope due t o s e l f - s h i e l d i n g e f f e c t s , thus s i g n i f i c a n t l y affecting t h e conversion of f e r t i l e m a t e r i a l t o f i s s i l e material. 4.1-2

Table

gives t h e resonance i n t e g r a l s a t core operating temperatures f o r various f u e l combina-

tions. Although t h e value of t h e 238U resonance i n t e g r a l f o r an i n f i n i t e l y d i l u t e medium i s much l a r g e r than t h e corresponding value f o r 232Th, t h e resonance i n t e g r a l f o r *38U i n LEU f u e l i s o n l y 25% l a r g e r than t h a t f o r 232Th i n HEU/Th f u e l , i n d i c a t i n g t h e much l a r g e r amount o f s e l f - s h i e l d i n g o c c u r r i n g f o r 238U i n LEU f u e l .

These two cases represent extreme values,


L

4-16

i

I

I Ii

since i n each case t h e one f e r t i l e isotope i s n o t s i g n i f i c a n t l y d i l u t e d by t h e presence o f

I

t h e other.

i

i

~

1

i

iI

I j

For MEU(20% 235U/U)/Th f u e l , t h e

238U

d e n s i t y i s reduced by a f a c t o r o f s6

( r e l a t i v e t o LEU f u e l ) , causing t h e 238U resonance i n t e g r a l t o increase due t o t h e reduced The decrease i n t h e 232Th d e n s i t y f o r t h e MEU/Th f u e l ( r e l a t i v e t o t h e

self-shielding.

HEU/Th) f u e l i s o n l y a f a c t o r o f s0.8 - r e s u l t i n g i n a much smaller increase i n t h e 232Th resonance i n t e g r a l . Thus, although the 238U number d e n s i t y i s roughly s i x times l e s s i n MEU/Th f u e l than i n LEU f u e l , t h e f i s s i l e Pu production i n t h e MEU/Th f u e l i s s t i l l 40% o f t h a t f o r t h e LEU f u e l as shown i n Table 4.1-1 238U resonance i n t e g r a l :

(Cases A and B) due t o t h e increase i n t h e

L L

The presence i n denatured uranium-thorium f u e l s o f two f e r t i l e isotopes having resonances a t d i f f e r e n t energy l e v e l s has a s i g n i f i c a n t e f f e c t on t h e i n i t i a l loading requirement.

The i n i t i a l 235U requirement f o r t h e HEU/Th and MEU/Th cases i s 2375 and

2538 kg/GWe, respectively, r e f l e c t i n g t h e penalty associated w i t h t h e presence o f t h e two f e r t i l e isotopes i n t h e MEU/Th fuel. The l a r g e increase i n i n i t i a l i I

I ~

235(5

requirements shown i n Table 4.1-1

f o r t h e thorium-

based HEU/Th and MEU/Th f u e l s compared t o t h e LEU f u e l r e s u l t s p r i m a r i l y from t h e l a r g e r thermal-absorption cross section o f 232Th r e l a t i v e t o 2 3 8 U as shown i n Table 4.1-2. Also c o n t r i b u t i n g t o t h e increased 2 3 5 U requirements i s t h e lower value o f 11 o f 235U which r e s u l t s from t h e harder neutron energy spectrum i n thorium-based fuels.

~

i Table 4.1-2.

~

iI

Thermal Absorption Cross Sections and Resonance I n t e g r a l s f o r 232Th and 238U i n PWRs

I

Resonance I n t e g r a l a (barns) Isotope

u

(0.025 eV) a (barns)

‘For

I n LEU Fuel

I n HEU/Th Fuel

23 2Th

7.40

85.8

-

17

23 8 u

2.73

273.6

21-22

-

~~~

I

Infinitely D i 1Ute

~

I n irlEU(235U/U)/Th Fuel

c

19 59-54

~~

absorption from 0.625 eV t o 10 MeV; oxide f u e l s .

!

i ~

I

1

A f u r t h e r consideration regarding MEU(235U/U)/Th f u e l w i t h uranium r e c y c l e must a l s o be noted. Since t h e f i s s i l e enrichment o f t h e recovered uranium decreases w i t h each generat i o n o f r e c y c l e f u e l , t h e thorium loadings must c o n t i n u a l l y decrease. (As pointed o u t above, i t i s assumed t h a t t h e recovered uranium i s n o t reenriched by i s o t o p i c separation techniques.) The i n i t i a l core 232Th/238U r a t i o i s ~ 5 . 8 and t h e f i r s t r e l o a d 232Th/238U r a t i o i s 4.4, b u t by t h e f o u r t h r e c y c l e generation t h e 232Th/238U r a t i o has declined t o ~ 1 . 4 . ~An a l t e r n a t i v e i s t o use HEU (93.15 w/o 235U) as a f i s s i l e topping f o r t h e recovered uranium. I n t h i s way t h e recovered uranium could be reenriched t o an allowed denaturing l i m i t p r i o r t o recycle, thus minimizing t h e core 238U component and t h e r e f o r e minimizing t h e production o f plutonium.

I I

j

I

i

b’ t -q

(r

j !

,i !

i I

I;‘


4-1 7

u--

The use o f HEU as a f i s s i l e topping could be achieved by f i r s t t r a n s p o r t i n g uranium recovered from t h e discharged f u e l t o a secure enrichment f a c i l i t y capable o f producing HEU. Next, t h e

HEU f i s s i l e topping would be added t o t h e recovered uranium t o r a i s e t h e f i s s i l e content o f

x i

t h e product t o an allowable l i m i t f o r denatured uranium. The product (denatured) would then be returned t o t h e f a b r i c a t i o n plant. MEU(20% 235U)/Th would be used t o supply t h e remainder

I*i

o f t h e makeup requirements.

Mass flows f o r t h i s o p t i o n i n which HEU i s used as a f i s s i l e

topping are reported i n r e f s . 2 and 6. For Case D, i n which t h e r e c y c l e f u e l i s n o t reenriched by a d d i t i o n o f HEU f i s s i l e topping, about 35% more plutonium i s bred over 30 y r (%60%more i n e q u i l i b r i u m ) than when t h e HEU i s used as a f i s s i l e topping. The 30-yr cumulative U3O8 and SWU requirements f o r t h e case i n which HEU i s used as a f i s s i l e topping are 4120 ST U308/GWe

and 3940 x 103 SWU/GWe r e s p e c t i v e l y a t a 75% capacity f a c t o r and 0.20 w/o t a i l s . 2

Table 4.1-3. I s o t o p i c Fractions o f Plutonium i n Pu02/Th02 PWRs E q u i l i b r i u m Once-Through Cycle

=99pu

1bi

Charged

Discharged

0.5680

0.2482

I n a d d i t i o n t o t h e uranium f u e l cycles discussed above, two d i f f e r e n t Pu/Th cases were analyzed. As i n d i c a t e d i n Table 4.1-3, t h e degradation o f t h e f i s s i l e percentage o f t h e plutonium which occurs i n a s i n g l e pass (i.e., once-through) i s r a t h e r severe. Thus, i n addi-

240Pu

0.2384

0.3742

=41pu

0.1428

0.2207

t i o n t o t h e plutonium r e c y c l e case (Case H) a case was considered i n which t h e discharged

242Pu

0.0508

0.1568

plutonium (degraded i s o t o p i c a l l y by the. burnup)

Fissile P1utonium

0.7108

0.4689

i s n o t recycled b u t r a t h e r i s s t o c k p i l e d f o r l a t e r use i n breeder reactors (Case I ) .

Only l i m i t e d analyses o f s a f e t y parameters have been performed thus f a r f o r t h e a l t e r n a t e f u e l types.

Combustion Engineering has reported some core physics parameters f o r

thorium-based (Pu02/Th02) and uranium-based ( P u O ~ / ~ ~ ~ APRs,* U O ~ ) and t h e remaining discuss i o n i n t h i s s e c t i o n i s taken from t h e i r a n a l y ~ i s : ~ . I n general, t h e s a f e t y - r e l a t e d core physics parameters (Table 4.1-4)

o f t h e two

burner r e a c t o r s a r e q u i t e s i m i l a r , i n d i c a t i n g comparable behavior t o p o s t u l a t e d accidents and p l a n t transients. Nevertheless, t h e f o l l o w i n g d i f f e r e n c e s a r e noted. The e f f e c t i v e delayed neutron f r a c t i o n (Beff)

and t h e prompt neutron l i f e t i m e ( a * ) are smaller f o r t h e

thorium APR. These are t h e c o n t r o l l i n g parameters i n t h e r e a c t o r ' s response t o short-term (*seconds) power transients. However, t h e most l i m i t i n g accident f o r t h i s type t r a n s i e n t i s u s u a l l y t h e r o d e j e c t i o n accident and since t h e e j e c t e d r o d worth i s l e s s f o r t h e thorium APR, t h e consequences o f t h e smaller values o f these k i n e t i c s parameters a r e l a r g e l y mitigated. The moderator and f u e l temperature c o e f f i c i e n t s a r e parameters which a f f e c t t h e i n h e r e n t s a f e t y o f t h e core. I n t h e power operating range, t h e combined responses o f these r e a c t i v i t y feedback mechanisms t o an increase i n r e a c t o r thermal power must be a decrease i n core r e a c t i v i t y . easily satisfied.

Since both c o e f f i c i e n t s are negative, t h i s requirement i s

The fuel temperature c o e f f i c i e n t i s about 25% more negative f o r t h e

*All-plutonium reactors.


4-18

thorium APR, w h i l e t h e moderator temperature c o e f f i c i e n t i s approximately 20% l e s s negat i v e . These differences compensate, t o a l a r g e extent, such t h a t t h e consequences o f accidents which-involve a core temperature t r a n s i e n t would be comparable. For some accidents, however, i n d i v i d u a l temperature c o e f f i c i e n t s a r e t h e c o n t r o l l i n g parameters, and f o r these cases t h e consequences must be evaluated on a case-by-case basis.

Control r o d and s o l u b l e boron worths are s t r o n g l y dependent on t h e thermal-neutron d i f f u s i o n length. Because o f t h e l a r g e r thermal absorption cross s e c t i o n o f 232Th and t h e higher plutonium loadings of t h e thorium APR, t h e d i f f u s i o n l e n g t h and, consequently, t h e c o n t r o l r o d and soluble boron worths are smaller. Table 4.1-4.

O f primary concern i s t h e Kainte-

Safety-Related Core Physics Parameters f o r Pu-Fueled PWRs ~

Third-Cycle Uranium APR E f f e c t i v e Delayed Neutron F r a c t i o n BOC EOC Sec) Prompt Neutron L i f e t i m e (x BOC EOC Inverse Soluble Boron Worth (PPM/% A P ) BOC EOC Fuel Temperature C o e f f i c i e n t ( x 10'5~p/oF) BOC EOC Moderator Temperature C o e f f i c i e n t (x 10-4Ap/"F) BOC EOC Control Rod Worth (% o f U02 APR) BOC EOC

.00430 .00438 10.54 12.53 221 180

T h i r d -Cycle Thorium APR 0.00344 0.00367

9.03 11.30

L1 b;

270 217

-1.13 -1.15

-1.40 -1.42

-1.65 -3.32

-1.31 -2.60

-

tl'

b'

90 96

nance o f adequate shutdown margin t o compensate f o r t h e r e a c t i v i t y defects d u r i n g postul a t e d accidents, e.g.

,for

t h e r e a c t i v i t y increase associated w i t h moderator cooldown i n

t h e steam-line-break accident. The analysis of i n d i v i d u a l accidents o f t h i s type would have t o be performed t o f u l l y assess t h e consequences o f t h e 10% r e d u c t i o n i n c o n t r o l - r o d worth a t t h e beginning o f cycle. The o v e r a l l r e s u l t s o f t h e above comparison o f core physics parameters i n d i c a t e t h a t t h e consequences o f postulated accidents f o r t h e thorium APR are comparable t o those o f t h e uranium APR.

Furthermore, t h i s comparison i n d i c a t e s t h a t o t h e r than t h e possi-

b i l i t y o f r e q u i r i n g a d d i t i o n a l c o n t r o l rods, a thorium-based plutonium burner i s f e a s i b l e and major m o d i f i c a t i o n s t o a PWR (already designed t o accommodate a plutonium-fueled core) are probably n o t required, although some m o d i f i c a t i o n s might be d e s i r a b l e i f r e a c t o r s were s p e c i f i c a l l y designed f o r operation w i t h high-Th content fuels.

I; fi ki


4-1 9

4.1.2.

B o i l i n g Water Reactors

Mass f l o w c a l c u l a t i o n s f o r BWRs presented i n t h i s chapter were performed by

r la

i , I

General E l e c t r i c .

A d e s c r i p t i o n o f the f u e l assembly designs developed by General The f o l l o w i n g cases have

E l e c t r i c f o r t h e u t i l i z a t i o n o f thorium i s presented i n Ref. 8. been analyzed: D i s p e r s i b l e Resource-Based Fuels

br

-_

t

A.

LEU, no recycle.

6.

B'.

MEU/Th, K O rec.vcle. LEU/Th mixed l a t t i c e (LEU and Tho2 rods), no recycle.

R".

LEU/)?EU/Th mixed l a t t i c e (Lâ&#x201A;ŹU/Th, MEU/Th, and Tho2 rods), no recycle.

D.

LEU/MEU/Th mixed l a t t i c e , r e c y c l e o f uranium, 235U makeup.

D i s p e r s i b l e Denatured Fuel

E.

Denatured 233U, recycle o f uranium, a33U makeup.

Enerqu-Center-Constrained Fuels

c

c L

e

G.

LEU, r e c y c l e o f uranium and self-generated plutonium, 235U makeup. P U / ~ ~ ~ rUe c, y c l e o f plutonium, plutonium makeup.

H.

P u / ~ ~ ~r eTc ~y c,l e o f plutonium, plutonium makeup.

F.

Case A represents t h e c u r r e n t mode of BWR operation. Case B involves the replacement of t h e c u r r e n t LEU f u e l w i t h MEU/Th f u e l i n which the i n i t i a l uranium enrichment i s l i m i t e d t o 20% 235U/238U.

Cases B ' and B" represent p a r t i a l thorium loadings t h a t could be

u t i l i z e d as a l t e r n a t i v e stowaway options.

I n Case B ' a few o f t h e LEU p i n s i n a

conventional LEU l a t t i c e are replaced w i t h pure Tho2 pins, w h i l e i n Case B" some LEU p i n s i n a conventional l a t t i c e are replaced by MEU/Th p i n s and a few others are replaced w i t h t h e pure Tho, pins. These cases are i n c o n t r a s t w i t h Case B i n which a " f u l l " thorium loading i s used (U02/Th02 i n every pin). Case D represents t h e extension o f Case B" t o the r e c y c l e mode; however, o n l y t h e uranium recovered from t h e Th-bearing pins i s recycled. Cases F-H represent possible f i s s i l e / f e r t i l e combinations f o r use i n secure energy centers. Table 4.1-5

provides a summary o f c e r t a i n mass balance information f o r BWRs operating

on these fuel cycles. A l l r e c y c l e cases i n v o l v e a two-year ex-reactor delay f o r reprocessing and r e f a b r i cation.

As was shown i n Table 4.1-1 f o r PWRs, t h e i n t r o d u c t i o n o f thorium i n t o a BWR core i n f l i c t s a penalty w i t h respect t o the resource requirements o f the r e a c t o r (compare U308 and SWU requirements o f Cases A and 6). f o r a f u l l thorium loading.

L

However, as pointed o u t above, Case B i s

I n t h e two General E l e c t r i c f u e l assembly designs8

represented by Cases 6' and 6" a much smaller f i s s i l e inventory penalty r e s u l t s from t h e i n t r o d u c t i o n o f thorium i n the core. PWRs. )

( S i m i l a r schemes may a l s o be f e a s i b l e f o r


4-20

Table 4.1-5.

Fuel U t i l i z a t i o n Characteristics f o r BWRs Under Various Fuel Cycle Options'

Initial Fissile Fissile Inventory Charge (kg/GWe) (kg/G%yrJ

Case

Fuel Type

A

LEU. no recycle

8

MEU/Th. no recycle'

8'

LEU/Th mixed lattice, no recyclee

-

854 23511

B"

LEU/HEU/Th mixed lattice, no recyclee

-

917

235U

D

LEU/MEU/Th mixed lattice, self-generated U recyclee

147 742

233U 235U

770 15

233U 235U

Equilibrium Cycle Fissile Discharge Burnup (kg/GWe-yr) (MWO/kg HM)

U30s Requirement (ST/GWe 1 30-yr I n i t i a l Total'

Separative Work Requirement (10' kg SWU/GWe) Initial

30-yr Total'

Dispersible Resource-Based Fuels 220OC

799 23% 1132

235 2 3 % 150 Puf

28.4

244 233U 428 23sU 53 Puf

31.6

868d

7764

;:2

28.7

6201

3836

125 233U 277 23sU 92 Puf

30.0

6852

5100f

152 233U 245 23sU 98 Puf

30.5

55035

3895f

z;

496C'd

6051d

235C'd

3493'

138 Puf

Dispersible Denatured Fuel E

Denatured 233U02/Th02, U recycle (exogeneous 2S3U makeup)e

F

LEU, recycle of U + self-generated Pu

-

-

452 *))U 17 23sU 55 Puf

31.6

0

0

0

L L 1:

0

Energy-Center-Constrained Fuels

Gh PuO2/UO2. Pu recycle

H

Pu02/Th02.

PU

recyclee

2200'

-

71 23511 1178 Puf

-

1705 Puf

38

2%

28.4

496'

38699

235'

19809

27.7

i

i

i

i

808 puf

275 233U 954 Puf

29.8

0

0

0

0

a A l l cases assume 0.2 w/o t a i l s and 75% capacity. factor; blank columns included t o show no data corresponding t o that given f o r

PWRs (Table 4.1-1) are available. bNo credit taken for end-of-reactor-life f i s s i l e inventory. 'Initial cycle i s 1.47 y r i n length a t 75%capacity factor. dFrom ref. 9. Based on three-enrichment-zone i n i t i a l core, axial blankets and improved refueling patterns which are currently being retrofitted i n t o many BWRs. 30-yr Up, and SWU requirements supplied t o INFCE f o r a reference BWR not employing these improvements are 6443 ST UpalGWe and 3887 x 103 SWU/GWe respectively. 'Analyses performed f o r equilibrium cycle only. 'Approximated from equilibrium cycle requirements. gFrom r e f . 8. 'From r e f . 10; adjusted from 80% capacity factor t o 75%. $ails uranium used f o r plutonium diluent.

Case B ' i s a p e r t u r b a t i o n t o t h e reference U02 BWR assembly design i n t h a t t h e f o u r U02 corner p i n s i n each fuel assembly a r e replaced w i t h f o u r pure Tho2 pins. The remaining U02 p i n s are adjusted i n enrichment t o o b t a i n a d e s i r a b l e l o c a l power d i s t r i b u t i o n and t o achieve r e a c t i v i t y l i f e t i m e . I n t h e once-through mode t h i s design increases U308 r e q u i r e ments by o n l y 2% r e l a t i v e t o t h e reference design.

This o p t i o n could be extended by

removing the Tho2 corner p i n s from t h e spent f u e l assemblies, reassembling them i n t o new assemblies, and r e i n s e r t i n g them i n t o t h e reactor. This would p e r m i t the Tho2 p i n s t o achieve increased burnups (and a l s o increased U308 requirements f o r t h i s scheme (i.e.,

233U

production) w i t h o u t reprocessing.

re-use o f t h e Tho2 rods coupled w i t h UO:! stowaway)

are approximately 1.3% higher than f o r t h e reference U02 cycle.8

c


4-21

Case B" i s a m o d i f i c a t i o n o f Case B' i n t h a t i n a d d i t i o n t o t h e f o u r Tho2 corner pins, t h e o t h e r p e r i p h e r a l p i n s i n t h e assembly are composed o f MEU(235)/Th. remainder o f t h e p i n s c o n t a i n LEU.

The

I n t h e once-through mode t h i s design increases U308

requirements by 12% r e l a t i v e t o t h e reference BWR U02 design. Both Case B' and Case B" would o f f e r 'operational b e n e f i t s t o t h e BWR since they have a l e s s negative dynamic v o i d c o e f f i c i e n t than t h e reference U02 design.8 This i s d e s i r a b l e since t h e s e n s i t i v i t y t o pressure t r a n s i e n t s i s reduced. As shown i n Table

4 . 1 4 , i n e q u i l i b r i u m conditions a BWR employing t h e Tho2 corner p i n once-through des i g n would discharge 24 kg 233U/Gl.le annually w h i l e t h e BWR employing t h e p e r i p h e r a l Tho2 mixed l a t t i c e design would discharge 125 kg 233U/GWe annually. Use o f these ciptions i n t h e once-through mode n o t o n l y could improve t h e operational performance o f t h e BWR b u t a l s o would b u i l d up a supply o f 233U.

This supply would then

be a v a i l a b l e i f a denatured 233U c y c l e (together w i t h reprocessing) were adopted a t a l a t e r time. Furthermore, use of t h e mixed l a t t i c e designs could be used t o acquire experience on t h e performance o f thorium-based f u e l s i n BWRs. S i m i l a r schemes f o r t h e use o f thorium

i n t h e once-through mode may a l s o be f e a s i b l e i n PWRs.

Although o n l y l i m i t e d scoping a n a l y s i s o f t h e s a f e t y parameters i n v o l v e d i n t h e use o f a l t e r n a t e f u e l s i n BWRs has been performed,8 t h e BWR thorium f u e l designs appear t o o f f e r some advantageous trends over U02 designs r e l a t i v e t o BWR operations and safety. Uranium/thorium f u e l s have a l e s s negative steam v o i d r e a c t i v i t y c o e f f i c i e n t than t h e U02 reference design a t e q u i l i b r i u m . This e f f e c t tends t o reduce t h e s e v e r i t y o f overpressurization accidents and improve t h e r e a c t o r s t a b i l i t y . The l e s s negative v o i d r e a c t i v i t y c o e f f i c i e n t f o r t h e denatured 23%/Th f u e l i n d i c a t e s t h a t t h e core w i l l have a f l a t t e r axial. oower shape than t h e reference U02 design.

This could r e s u l t i n an increase i n kW/ft margin and increase t h e maximum average planar heat generation r a t i o (MAPLHGR). A l t e r n a t i v e l y , i f c u r r e n t margins a r e maintained, t h e f l a t t e r a x i a l power shape c o u l d be u t i l i z e d t o increase t h e power d e n s i t y o r t o a l l o w r e f u e l i n g p a t t e r n s aimed a t improved f u e l u t i l i z a t i o n .

References f o r Section 4.1

1.

N. L. Shapiro, J. R. Rec, and R. A. I l a t z i e (Combustion Engineering), "Assessment o f Thorium Fuel Cycles i n Pressurized Water Reactors," EPRI NP-359 (Feb. 1977).

2.

"Thorium Assessment Study Q u a r t e r l y Progress Report f o r Second Q u a r t e r F i s c a l 1977," ORNL/TEl-5949 (June 1977).

3.

R. A. Hatzie, J. R. Rec, and A. N. Terney, "An Evaluation o f Denatured Thorium Fuel Cycles i n Pressurized Water Reactors,' paper presented a t t h e Annual Meeting o f t h e American Nuclear Society, June 12-16, 1977, New York, Mew Yorl:.


4-22

Letter from P. A. Hatzie (Combustion Engineering) t o H. B. Stewart (Nuclear Technology Evaluations Company), ''U,O, Requirements i n LllRs and SSCRs," July 25, 1977. 5.

Letter from I?. A. Hatzie (Combustion Engineering) t o J. C. Cleveland (ORNL), Wass Balances for Various LWF? Fuel Cycles," Hay 1977. "Quarterly Progress Report f o r Fourth Quarter vi'-77, Thorium Assessment Program," Combustion Engineering.

7.

"P:uclear Power Growth 1974-2000,'' Office o f Planning and Analysis , U.S. Atonic Energy Comission, I!P,SH 1139( 74) , (February 1574).

8.

"Assessment of Uti1 ization of Thorium i n BWRs ,I' ORNL/SUB-4380/5, NEGD-24073 (January 1978).

9.

"Monthly Progress Report f o r August 1978, NASAP Preliminary BWR Uranium Utilization Improvement Evaluations," General Electric Co.

10.

"Appraisal of BWR Plutonium Burners f o r Energy Centers ,I' GEAP-11367 (January 1976).

L L E

c L


c

4-23

4.2.

SPECTRAL-SHIFT-CONTROLLED REACTORS

N. L. Shapiro Combustion Engineering, Inc. The Spectral -Shift-Control l e d Reactor (SSCR) i s an advanced thermal converter r e a c t o r t h a t i s based on PWR technology and o f f e r s improved resource u t i l i z a t i o n , p a r t i c u l a r l y on t h e denatured f u e l cycle. The SSCR d i f f e r s from t h e conventional PWR i n t h a t i t i s designed t o minimize t h e number of reactions i n c o n t r o l m a t e r i a l s throughout t h e p l a n t l i f e , u t i l i z i n g t o t h e e x t e n t p o s s i b l e captures o f excess neutrons i n f e r t i l e m a t e r i a l as a method o f r e a c t i v i t y c o n t r o l .

The r e s u l t i n g increase i n t h e production o f f i s s i l e

m a t e r i a l serves t o reduce f u e l makeup requirements.

L i a

I n t h e conventional PWR, long-term r e a c t i v i t y c o n t r o l i s achieved by varying t h e concentration o f s o l u b l e boron i n t h e coolant t o capture t h e excess neutrons generated throughout p l a n t l i f e .

The s o l u b l e boron concentration i s r e l a t i v e l y h i g h a t beginning

o f cycle, about 700 t o 1500 ppm, and i s g r a d u a l l y reduced during t h e operating c y c l e by the i n t r o d u c t i o n o f pure water t o compensate f o r t h e d e p l e t i o n of f i s s i l e i n v e n t o r y and t h e buildup o f f i s s i o n products. The SSCR consists b a s i c a l l y o f t h e standard PWR w i t h t h e conventional soluble boron r e a c t i v i t y c o n t r o l system replaced w i t h s p e c t r a l - s h i f t c o n t r o l . S p e c t r a l - s h i f t c o n t r o l i s achieved by t h e a d d i t i o n o f heavy water t o t h e r e a c t o r coolant, i n a manner analogous t o t h e use of s o l u b l e boron i n t h e conventional PWR.

Since heavy water i s a poorer moderator

of neutrons than l i g h t water, t h e i n t r o d u c t i o n o f heavy water s h i f t s t h e neutron spectrum

L

L

i n t h e r e a c t o r t o higher energies and r e s u l t s i n t h e p r e f e r e n t i a l absorption o f neutrons i n f e r t i l e materials. I n c o n t r a s t t o t h e conventional PWR, where absorption i n c o n t r o l absorbers i s unproductive, t h e absorption o f excess neutrons i n f e r t i l e m a t e r i a l breeds a d d i t i o n a l f i s s i l e m a t e r i a l , increasing t h e conversion r a t i o o f t h e system and decreasing t h e annual makeup requirements. A t beginning of cycle, a h i g h (approximately 50-70 mole X ) D20 concentration i s employed i n order t o increase t h e absorption o f neutrons i n f e r t i l e m a t e r i a l s u f f i c i e n t l y t o c o n t r o l excess r e a c t i v i t y . Over t h e cycle, t h e spectrum i s thermalized by decreasing t h e D20/H20 r a t i o i n t h e coolant t o compensate f o r f i s s i l e m a t e r i a l d e p l e t i o n and f i s s i o n - p r o d u c t buildup, u n t i l a t end o f c y c l e e s s e n t i a l l y pure l i g h t water (approximately 2 mole % D20) i s present i n the coolant. The b a s i c changes r e q u i r e d t o implement s p e c t r a l - s h i f t c o n t r o l i n a conventional PWR are i l l u s t r a t e d i n a s i m p l i f i e d and schematic form i n Fig. 4.2-1. I n t h e conventional PWR, pure water i s added and borated water i s removed during t h e c y c l e t o compensate f o r t h e d e p l e t i o n o f f i s s i l e m a t e r i a l and b u i l d u p o f fission-product poisons. The borated water removed from t h e r e a c t o r i s processed by t h e boron concentrator which separates t h e discharged coolant i n t o two streams, one containing pure unborated water and t h e second


4-24

ORNL-DWG 78- 15056 Conventional Poison Control

c

Spectral S h i f t Control

Control

b'

S

.r

E .r

E

L 0

0,

$n 0

m o o Burnup

O Burnup

N EOC

EOC

H20 Makeup Tank

H20

1'

H20 H20 Makeup Tank

>

/~

Charging i

Borated Boron D20

I

Storage o f Highly Concentrated Boric Acid

Fig. 4.2-1.

.

D20 Upgrader

-Mixture

-

Storage o f Highly Concentrated (80%) D20

Basic Spectral Shift Control Modifications.

containing boron at high concentrations. The latter stream is stored until the beginning of the subsequent cycle where it is used to provide the boron necessary to hold down the excess reactivity introduced by the loading of fresh fuel. The SSCR can consist of the identical nuclear steam supply system as employed in a conventional poison-control led PWR, except that the boron concentrator is reDlaced with a D20 upgrader. The function of this upgrader is to separate heavy and light water, so that concentrated heavy water is available for the next refuelina. The upgrader consists of a series of vacuum distillation columns which utilize the differences in volatility between liqht and heavy water to effect the separation. Although the boron concentrator and the upgrader perform analogous functions and operate usina similar processes, the D20 upgrader is much larger and more sophisticated, consisting of three or four towers each about 10 ft in diameter and 190 ft tall. Although Fig. 4.2-1 illustrates the basic changes required to implement the shift-control concept, numerous additional changes will be required to realize spectral-shift control in practice. These include modifications to minimize and recover D20 leakage, to facilitate refueling, and to remove boron from the coolant after refueling.

L L

L


4-25

I n i t i a l analyses o f spectral-shift-controlled reactors were c a r r i e d o u t i n t h e U.S. by M. C. Edlund i n t h e e a r l y 1960s and an experimental v e r i f i c a t i o n program was performed by Babcock & Wilcox both f o r LEU f u e l s and f o r HEU/Th fuels.'

Edlund's studies, which

were performed f o r r e a c t o r s designed s p e c i f i c a l l y f o r s p e c t r a l - s h i f t c o n t r o l , i n d i c a t e d t h a t t h e i n v e n t o r y and consumption o f f i s s i l e m a t e r i a l could be reduced by 25 and 50%, r e s p e c t i v e l y , r e l a t i v e t o poison c o n t r o l i n r e a c t o r s f u e l e d w i t h h i g h l y enriched 235U and thorium oxide, and t h a t a 25% reduction i n uranium ore requirements could be r e a l i z e d w i t h s p e c t r a l s h i f t c o n t r o l using t h e LEU cycle.2 The s p e c t r a l - s h i f t - c o n t r o l concept has been demonstrated by t h e Vulcain r e a c t o r experiment i n t h e BR3 nuclear p l a n t a t Mol, B e l g i ~ m . ~The BR3 p l a n t a f t e r two years o f operation as a conventional PWR was modified f o r spectral - s h i f t - c o n t r o l s u c c e s s f u l l y operated w i t h t h i s mode o f c o n t r o l between 1966 and 1968.

operation and The Vulcain core

operated t o a core average burnup o f 23,000 M W T (a peak burnup o f around 50,000 MWd/T) and achieved an average load f a c t o r and primary p l a n t a v a i l a b i l i t y f a c t o r o f 91.2 and 98.6,

r e ~ p e c t i v e l y . ~The leakage r a t e o f primary water from t h e high-pressure r e a c t o r

system t o t h e atmosphere was found t o be n e g l i g i b l e , about 30 kg o f D20-H20 m i x t u r e per year.3

AfFer t h e Vulcain experiment was completed, t h e BR3 was subsequently returned t o

conventional PWR operation. I n a d d i t i o n t o demonstrating t h e t e c h n i c a l f e a s i b i l i t y o f s p e c t r a l - s h i f t c o n t r o l , t h e Vulcain experiment served t o i d e n t i f y t h e p o t e n t i a l engineering problems i n h e r e n t i n converting e x i s t i n g p l a n t s t o t h e s p e c t r a l - s h i f t mode of control.

A t t h e time o f t h e major development work

on t h e SSCR concept, f u e l resource con-

s e r v a t i o n was n o t recognized as having t h e importance t h a t i t has today.

Both uranium

ore and separative work were r e l a t i v e l y inexpensive and t h e technology f o r D20 concent r a t i o n was not as f u l l y developed as i t i s now.

With t h e expectation t h a t t h e plutonium-

f u e l e d breeder r e a c t o r would be deployed i n t h e n o t t o o d i s t a n t f u t u r e , t h e r e appeared t o be l i t t l e i n c e n t i v e t o pursue t h e spectral-shift-controlled r e a c t o r concept. The d e c i s i o n t o d e f e r t h e commercial use o f plutonium and the commercial plutoniumf u e l e d breeder r e a c t o r i s , o f course, t h e primary m o t i v a t i o n f o r reevaluating advanced converters, and t h e p r i n c i p a l i n c e n t i v e f o r considering t h e spectral-shift-controlled r e a c t o r i s t h a t t h e p o t e n t i a l gains i n resource u t i l i z a t i o n p o s s i b l e w i t h t h e SSCR concept may be obtainable w i t h changes l a r g e l y l i m i t e d t o a n c i l l a r y components and subsystems i n e x i s t i n g PWR systems. The prospects o f r a p i d acceptance and deployment o f t h e SSCR a r e ' a l s o enhanced by t h e low r i s k i n h e r e n t i n the concept. Since t h e SSCR can always be ~

operated i n t h e conventional poison c o n t r o l mode, t h e r e would be a reduced r i s k t o s t a t i o n generating capacity i f t h e SSCR were deployed, and f i n a n c i s i r i s k would be l i m i t e d t o t h e c o s t o f t h e a d d i t i o n a l equipment r e q u i r e d t o r e a l i z e s p e c t r a l - s h i f t c o n t r o l , which i s estimated t o be o n l y a few percent o f t h e t o t a l c o s t o f t h e plant. The r i s k , w i t h respect both t o c a p i t a l and generating capacity, i s thus much lower than f o r o t h e r a l t e r n a t e r e a c t o r systems.


4-26

I t may a l s o prove f e a s i b l e t o b a c k f i t e x i s t i n g pressurized water reactors with

spectral-shift control.

Such b a c k f i t t i n g might p o s s i b l y be performed i n some completed

p l a n t s where t h e l a y o u t favors modifications. However, even when judged f e a s i b l e , t h e b e n e f i t s o f b a c k f i t t i n g would have t o be g r e a t t o j u s t i f y t h e c o s t o f replacement power during p l a n t modification. A second and p o t e n t i a l l y more a t t r a c t i v e a l t e r n a t i v e i s t h e p o s s i b i l i t y o f modifying p l a n t s s t i l l i n t h e e a r l y stage o f c o n s t r u c t i o n f o r s p e c t r a l s h i f t c o n t r o l , o r o f i n c o r p o r a t i n g features i n t o these p l a n t s which would a l l o w conversion t o s p e c t r a l - s h i f t c o n t r o l t o be e a s i l y accomplished a t a l a t e r date. I n order t o e s t a b l i s h the p o t e n t i a l gains i n resource u t i l i z a t i o n which might be r e a l i z e d w i t h s p e c t r a l - s h i f t c o n t r o l , scoping mass balance c a l c u l a t i o n s have been performed by Combustion Engineering f o r SSCRs operating on both t h e LEU c y c l e and on thorium-based cycles, i n c l u d i n g t h e denatured 233U cycle.5

The c a l c u l a t i o n s were performed f o r t h e C-E

system 80TM core and l a t t i c e design, w i t h t h e i n t e n t o f updating t h e e a r l i e r analyses reported by Edlund t o the r e a c t o r design and operating conditions o f modern PWRs using s t a t e o f - t h e - a r t a n a l y t i c methods and cross sections. a r e presented i n Table 4.2-1.

Preliminary r e s u l t s from t h i s e v a l u a t i o n

Note t h a t these r e s u l t s were obtained using t h e standard

b

h:

System 80 design and operating procedures, and no attempt has been made t o optimize e i t h e r t h e l a t t i c e design o r mode o f operation t o f u l l y take advantage o f s p e c t r a l - s h i f t c o n t r o l . For t h e LEU throwaway mode, Table 4.2-1

i n d i c a t e s a r e d u c t i o n o f roughly 10% both

i n ore requirements and i n separative work requirements r e l a t i v e t o t h e conventional PWR (compare w i t h Case

A o f Table 4.1-1).

I f uranium r e c y c l e i s allowed, t h e SSCR a l s o reduces

c

t h e ore demand (and separative work) f o r t h e MEU/Th case by about 20% (compare w i t h Case D i n Table 4.1-1).

,

Of p a r t i c u l a r i n t e r e s t t o t h i s study i s t h e reduced e q u i l i b r i i m c y c l e makeup re-

quirements f o r t h e s p e c t r a l - s h i f t r e a c t o r f u e l e d w i t h 233U. As indicated, t h e e q u i l i b r i u m 233U/GWe-yr as opposed t o 304 kg 233U/GWe-yr f o r t h e c y c l e makeup requirement i s 236 I .

conventional PWR (see Case E i n Table 4.1-1). The reduced 233U requirements, coupled w i t h t h e s l i g h t l y higher f i s s i l e plutonium production, would a l l o w a given complement o f energycenter breeder reactors t o provide makeup f i s s i l e m a t e r i a l f o r roughly 40% more dispersed denatured SSCRs than conventional denatured PWRs. A comparison o f t h e Pu/Th case w i t h Case H i n Table 4.1-1 shows t h a t t h e SSCR and PWR are comparable as transmuters. These r e s u l t s are, of course, p r e l i m i n a r y and are l i m i t e d t o t h e performance o f otherwise unmodified PWR systems, A more accurate assessment o f SSCR performance, i n c l u d i n g t h e performance of systems optimized f o r s p e c t r a l - s h i f t c o n t r o l , w i l l be performed as p a r t o f t h e NASAP program.6 The p r e l i m i n a r y studies performed t o date and t h e demonstration of s p e c t r a l s h i f t c o n t r o l i n t h e Vulcain core have served t o demonstrate t h e f e a s i b i l i t y of t h e concept and t o i d e n t i f y t h e resource u t i l i z a t i o n and economic i n c e n t i v e s f o r t h i s

L


h

4-27 Table 4.2-1.

Fuel U t i l i z a t i o n C h a r a c t e r i s t i c 2 f o r SSCRs Under Various Fuel Cycle Optionsa* E q u i l i b r i u m Cycle

Initial F i s s ll e Inventory (kg/GWe 1

Fuel Type

Fissile Makeup (kg/GWe-yr)

Fissile Discharge (kg/GWe-yr)

30-Yr Cumulative U30, Requirement (ST/GWe)

30-Yr Cumulative Separative Work Requirement' (lo3 kg SW/GWe)

D i s p e r s i b l e Resource-Based Fuels LEU, no r e c y c l e

1577 23511

713

235U

182 23% 196 Puf

5320

3010

MEU/Th. 23sU feed, U recycle

2540

371

2ssU

228 235U 371 233U 65 Puf

3220

3077

--

--

--

--

D i s p e r s i b l e Denatured Fuel

1 '

b

Denatured 233U02/Th02, U recycle

1663 233U

236

233U

4;

E; 72 Puf

Energy-Center-Constrained Fuel PuO2/ThOz, Pu r e c y c l e

t

c

'1290-MWe

2354 Puf

791 Puf

780 Puf 273 233U 3 235u

SSCR; IO-MWe a d d i t i o n a l power r e q u i r e d t o r u n r e a c t o r c o o l a n t pumps and 020 upgrader f a c i l i t y .

bAs~umes75% c a p a c i t y factor, annual r e f u e l i n g , and 0.2 w/o t a i l s assay.

mode o f operation.

Because t h e basic PWR NSSS* i s used, t h e u t i l i z a t i o n of t h e denatured

thorium f u e l cycles w i l l pose no a d d i t i o n a l problems o r R&D needs beyond those r e q u i r e d t o implement t h i s t y p e o f f u e l i n t h e conventional PWR. Although t h e general f e a s i b i l i t y

i--;

o f s p e c t r a l - s h i f t c o n t r o l appears r e l a t i v e l y w e l l established, nevertheless t h e r e are a

&

number o f aspects o f SSCR design which must be evaluated i n order t o f u l l y assess the c o n e r c i a l p r a c t i c a l i t y o f s p e c t r a l - s h i f t - c o n t r o l l e d reactors. The more s i g n i f i c a n t o f

I-

&

these are b r i e f l y discussed below.

1.

- A more accurate assessment o f resource u t i l i z a t i o n i s

Resource U t i l i z a t i o n

r e q u i r e d t o more d e f i n i t i v e l y e s t a b l i s h t h e economic i n c e n t i v e s f o r s p e c t r a l - s h i f t c o n t r o l on t h e LEU cycle. I f t h e concept i s t o be economically competitive w i t h conventional water reactors, t h e savings i n Uj08 and separative work f o r 235U-based systems must be demonstrated t o be s u f f i c i e n t l y l a r g e t o compensate f o r t h e a d d i t i o n a l c a p i t a l c o s t o f equipment r e q u i r e d t o implement s p e c t r a l - s h i f t c o n t r o l . A s i m i l a r assessment f o r denatured 233U f u e l i s a l s o required.

1:

-

2. P l a n t Modifications The p l a n t m o d i f i c a t i o n s necessary t o r e a l i z e s p e c t r a l s h i f t c o n t r o l must be i d e n t i f i e d , and t h e cost o f these m o d i f i c a t i o n s established, The p r a c t i c a l i t y and c o s t o f these modifications, o f course, bear d i r e c t l y on t h e economics and commercial f e a s i b i l i t y o f t h e concept.

i

f

O f p a r t i c u l a r concern are m o d i f i c a t i o n s which

may be r e q u i r e d t o l i m i t t h e leakage of primary coolant (from valve stems, seals, etc.) and t h e equipment r e q u i r e d t o recover unavoidable primary coolant leakage. Primary coolant leakage i s important both from the standpoint o f economics, because o f t h e h i g h c o s t o f D20, and from t h e standpoint of r a d i a t i o n hazard, because o f t h e problem o f occup a t i o n a l exposures t o t r i t i u m during r o u t i n e maintenance, Other p o s s i b l e m o d i f i c a t i o n s t o current'designs which r e s u l t from t h e presence o f D20, such as t h e increased f a s t fluence on t h e r e a c t o r vessel and p o s s i b l e changes i n pumping power, w i l l a l s o have t o be addressed. NSSS = Nuclear Steam supply 'System.


4-28

3.

Refueling System M o d i f i c a t i o n s

- At

t h e end o f each operating cycle, spent f u e l

must be discharged and f r e s h f u e l i n s e r t e d i n t o t h e r e a c t o r ( t y p i c a l l y 1 / 3 - o f t h e core l o a d i n g i s replaced each year), and t h e l i g h t water present a t end o f c y c l e must be replaced w i t h a D20-H20 m i x t u r e before t h e r e a c t o r can be returned t o power operation. Refueling procedures and equipment must be developed which w i l l a l l o w these operations t o be performed w i t h minimum D20 inventory requirements. Minimizing t h e D20 i n v e n t o r y i s important t o t h e economics and commercial f e a s i b i l i t y o f t h e SSCR, since t h e c o s t o f D20 represents roughly 75% o f t h e a d d i t i o n a l c a p i t a l expenditures r e q u i r e d t o r e a l i z e s p e c t r a l s h i f t control.

Care must a l s o be taken t o ensure t h a t r e f u e l i n g does n o t increase outage

times because o f t h e adverse e f f e c t on capacity f a c t o r and t h e r e s u l t i n g increase i n power cost.

The exposure o f personnel t o t r i t i u m qenerated i n t h e coolant must a l s o be m i n i -

mized d u r i n g r e f u e l i n g operations.

4. D20 Upgrader Design - Although D20 upgraders have y e t t o be employed i n conj u n c t i o n w i t h s p e c t r a l - s h i f t c o n t r o l , s i m i l a r u n i t s have operated on CANDU reactors, and vacuum d i s t i l l a t i o n columns are a l s o u t i l i z e d i n heavy-water production f a c i l i t i e s . Thus, t h e t e c h n i c a l f e a s i b i l i t y o f t h e D20 upgrader can be considered as demonstrated. However, a conceptual upgrader design optimized f o r t h e s p e c i f i c demands o f t h e SSCR must be developed so t h a t i t s c o s t can be determined. The upgrader i s probably t h e s i n g l e most s i g n i f i c a n t and c o s t l y piece o f equipment which must be added t o r e a l i z e s p e c t r a l - s h i f t control.

-

Although t h e s p e c t r a l - s h i f t - c o n t r o l l e d r e a c t o r i s 5. L i c e n s a b i l i t y and Safety n o t expected t o r a i s e any new safety, l i c e n s i n g o r environmental issues except t h e basic issue o f t r i t i u m production and containment, a number o f core physics parameters are changed s u f f i c i e n t l y t h a t t h e response t o postulated accidents must be evaluated.

The

most s i g n i f i c a n t o f these appears t o be t h e somewhat d i f f e r e n t moderator temperature coe f f i c i e n t o f r e a c t i v i t y , which could lead t o a number o f p o t e n t i a l l y more severe accidents e a r l y i n c y c l e when t h e D20 concentration i s r e l a t i v e l y high.

The D20 d i l u t i o n accident

must a l s o be addressed; t h i s accident i s analogous t o t h e boron d i l u t i o n accident i n the poison-controlled PWR, b u t t h e response t o DO, d i l u t i o n may be more r a p i d and hence t h e accident may be p o t e n t i a l l y more severe than i t s counterpart i n t h e PWR. F i n a l l y , i t should be pointed o u t t h a t w h i l e t h e r e l a t i o n s h i p o f t h e SSCR t o t h e

LWR

I

gives i t market advantages, i t a l s o gives i t some disadvantages r e l a t i v e t o o t h e r

alternatives.

Although t h e SSCR demand f o r U308 w i l l be l e s s than t h a t o f t h e conventional

LWR, t h e basic p r o p e r t i e s o f l i g h t water and t h e LWR design c h a r a c t e r i s t i c s inherent i n t h e SSCR w i l l l i m i t i t s f u e l u t i l i z a t i o n e f f i c i e n c y t o lower l e v e l s than those achievable w i t h other a l t e r n a t i v e s such as t h e HWR. On t h e o t h e r hand, t h e prospect f o r e a r l y and widespread deployment may mean t h a t i t could e f f e c t a more s i g n i f i c a n t r e d u c t i o n i n overa l l system U308 demand than might be achievable w i t h o t h e r a l t e r n a t i v e s , even though t h e inherent resource u t i l i z a t i o n o f an i n d i v i d u a l SSCR p l a n t may be l e s s than t h a t o f o t h e r systems.

Employing denatured SSCRs would a l l o w a d d i t i o n a l time t o develop e f f e c t i v e


4-29

c-

i

L

safeguards for breeder reactors which will eventually be required. These breeders might produce 233U, which, as pointed out above, could then be denatured and used in SSCRs.

L

e I

References for Section 4.2 1.

T. C. Engelder, et al., "Spectral Shift Control Reactor Basic Physics Program, BAW1253 (November 1963).

2.

M. C. Edlund, "Developments in Spectral Shift Reactors," Proceedings of the Third International Conference on the Peaceful Uses of Atomic Energy," V o l . 6, p. 314-320, Geneva, Switzerland (1964).

3.

J. Storrer, "Experience with the Construction and Operation of the BR3Dulcain Experiment," Symposium on Heavy Water Power Reactors, IAEA Vienna (September 11-15, 1967).

4. J. Storrer, et al., "Belgonucleaire and Siemens PWRs for Small and Medium Power Reactors," Proceedin s of a Symposium on Small and Medium Power Reactors, IAEA Oslo (October 12-16, 1970f. 5.

b

.

Letter, R. A. Matzie (Combustion Engineering) to J. C. Cleveland (ORNL), "Mass Balances for SSCRs and LWRs," (May 10, 1977).

6. Nonproliferation Alternative Systems Assessment Program May 1977.

- Preliminary Plan (draft)


4-30

4.3.

HEAVY-WATER REACTORS

Y. I. Chang Argonne National Laboratory

Due t o t h e low neutron absorption cross s e c t i o n o f deuterium, reactors u t i l i z i n g heavy water as t h e moderator t h e o r e t i c a l l y can a t t a i n higher conversion r a t i o s than reactors using o t h e r moderators. As a p r a c t i c a l matter, however, d i f f e r e n c e s i n t h e neutron absorption i n t h e s t r u c t u r a l m a t e r i a l s and f i s s i o n products i n t h e d i f f e r e n t r e a c t o r types make t h e conversion e f f i c i e n c y more dependent on r e a c t o r design than on moderator type.

I n t h e study

reported here, a current-generation 1200-MWe CANDU design was chosen as t h e model f o r examining t h e e f f e c t s o f various f u e l c y c l e options, i n c l u d i n g t h e denatured 233U cycle, on heavy-water-moderated reactors. The CANDU design d i f f e r s from t h e LWR design p r i m a r i l y i n t h r e e areas:

i t s reference

f u e l i s n a t u r a l uranium r a t h e r than enriched uranium; i t s coolant and moderator are separated by a pressure tube; and i t s f u e l management scheme employs continuous o n - l i n e r e f u e l i n g r a t h e r than p e r i o d i c r e f u e l i n g . I n t h e development o f t h e CANDU r e a c t o r concept, neutron I

economy was stressed, t r y i n g i n e f f e c t t o take maximum advantage o f t h e D20 properties.

The

o n - l i n e r e f u e l i n g scheme was introduced t o minimize t h e excess r e a c t i v i t y requirements. U n l i k e i n most o t h e r r e a c t o r systems, i n t h e natural-uranium DO, system t h e payoff i n r e ducing p a r a s i t i c absorption and excess r e a c t i v i t y requirements i s d i r e c t and s u b s t a n t i a l i n t h e amount o f burnup achievable. These same considerations a l s o make t h e CANDU an e f f i c i e n t converter when t h e n a t u r a l uranium r e s t r i c t i o n i s removed and/or f u e l i n g schemes based on r e c y c l e m a t e r i a l s are introduced. Penalties associated w i t h t h e improved neutron economy i n t h e natural-uraniumf u e l e d CANDU i n c l u d e a l a r g e inventory o f t h e moderator ( t h e D20 being a s i g n i f i c a n t port i o n o f t h e p l a n t c a p i t a l cost), a l a r g e f u e l mass f l o w through t h e f u e l c y c l e and a l o h e r thermal e f f i c i e n c y .

I n enriched f u e l cycles, w i t h t h e r e a c t i v i t y c o n s t r a i n t reliioved, the

CANDU design can be reoptimized f o r t h e p r e v a i l i n g econonic and resource conditions.

The r e o p t i m i z a t i o n o f t h e c u r r e n t CANCU design involves t r a d e o f f s between economic considerations and t h e neutron economy (and hence t h e f u e l u t i l i z a t i o n ) . For example, t h e D20 inventory can be reduced by a s n a l l e r l a t t i c e p i t c h , b u t t h i s r e s u l t s i n a poorer f u e l u t i l i z a t i o n . Also, t h e l a t t i c e p i t c h i s constrained by t h e p r a c t i c a l l i m i t a t i o n s placed on i t by t h e r e f u f l i n g machine operations. The f u e l mass f l o w r a t e (and hence t h e fabrication/reprocessing c o s t s ) can be reduced by increasing t h e discharge burnup, b u t t h e increased burnup a l s o r e s u l t s i n a poorer fuel utilization.

I n a d d i t i o n , t h e burnup has an impact on t h e f u e l i r r a d i a t i o n perform-

ance r e l i a b i l i t y .

The f u e l f a i l u r e r a t e i s a strong f u n c t i o n o f t h e burnup h i s t o r y , and


4-31

a significant increase in burnup over the current design would require mechanical design modifications.

L

t

The thermal efficiency can be improved by increasing the coolant pressure. This would require stronger pressure tubes and thus penalize the neutron economy. The use of enriched fueling could result in a higher power peaking factor, which would require a reduced linear power rating, unless an improved fuel management scheme is developed to reduce the power peaking factor. Scoping calculations have been performed to address possible design modifications for CANDU fuel cycles other than natural ~ranium,”~ and detailed design tradeoff and optimization studies associated with the enriched fuel cycles in CANDUs are being carried out by Combustion Engineering as a part of the NASAP program. In the study reported here, in which only the relative performance of the denatured 233U cycle is addressed, the currentgeneration 1200-MWe CANDU .fuel design presented in Table 4.3-1 was assumed for all except the natural-uranium-fueled reactor. A discharge burnup of 16,000 MWD/T (which is believed to be achievable with the current design) and the on-line refueling capability were also assumed. The fuel utilization characteristics for various fuel cycle options, including the denatured 233U cycle option, were analyzed at Argonne National Laboratory5 and the results are sumnarized in Table 4.3-2. Some observations are as follows:

ibi

L

1.

Natural-Uranium Once-Through Cycle: In the reference natural uranium cycle, the 30-yr U308 requirement is about 4,700 ST/GWe, which is approximately 20% less than the requirement for the LWR once-through cycle. Even though the fissile plutonium concentration in the spent fuel is low (4.27%). the total quantity of fissile plutonium discharged annually is twice that from the LWR.

2. Slightly-Enriched-Uranium Once-through Cycle: With slightly-enriched uranium (1% 235U), a 16,000-MWD/T burnup can be achieved and the U308 consumption is reduced by 25% from the natural-uranium cycle. As shown in Fig. 4.3-1, the optimum enrichment is in the area of 1.2%, which corresponds to a burnup of about 20,000 MWD/T.

L t i

3. Pu/U, Pu Recycle: In this option, the natural uranium fuel is “topped” with 0.3% fissile plutonium. A discharge burnup of 16,000 MWD/T can be achieved and the plutonium content in the discharge is sufficient to keep the system going with only the natural-uranium makeup. The U308 requirement is reduced to about one half of that for the natural-uranium cycle. (Smaller plutonium toppings decrease the burnup and make the system a net plutonium producer; larger toppings increase the burnup and make the system a net plutonium burner.)


L

4-32

Table 4.3-1.

CANDU-PHW Design Parameters

Ratural Uranium System Fuel Element Sheath 0.d. mn Sheath i.d; mn Sheath material Pellet o.d, mn Fuel density, g/cc Fuel material

13.075 12.237 Zr-4 12.154 10.36 u02

13.081 12.244 Zr-4 12.154 9.4 Tho2

37 495.3 476.82 54.29 24.14 76.69

Bund1e Number o f elements/bundle Length, mn Active fuel length, rm Volume of end plugs, etc., cc Void i n end region. cc Coolant i n end region, cc Ring 1 No./radius, mm) Ring 2 No./radius, Ring 3(No./radius, mn Ring 4(No./radius, m

6/14.885 12/28.755 18/43.305

37 495.3 475.4 65.68 34.99 66.43 1/0.0 6/14.884 12/28.753 18/43.307

Ctiannel Number o f bundles Pressure tube material Pressure tube i.d, nun Pressure tube 0.d. mn Calandria tube material Calandria tube i.d, nnn Calandria t u b e o.d, mn P i t c h , nnn

12 Zr-lib 103.378 111.498 Zr-2 128.956 131.750 285.75

12 Zr-Nb 103.400 111.782 Zr-2 129.200 131.740 285.75

380 633 29.0

728 1229 29.7

99.75 271.3

99.75 269.3

936 290 290 68

850 293 293 57

t

!

Thorium System

Core Number o f channels

7

Net MWe Net thermal efficiency, % Operating Conditions D20 purity, % Average pin l i n e a r pqer, W/cm Average temperature, C Fuel Sheath Cool ant Moderator

1/0.0

L

G

G

c

I7

Li


Table 4.3-2.

c

r-"!r!r-'i

c"-1

kc'l

Fuel U t i l i z a t i o n C h a r a c t e r i s t i c s f o r CANDUs Under Various Fuel Cycle Optionse

~

E q u i l i b r i u m Cycle

Fuel Type

Initial Fissile Inventory (kg/GWe)

Fissile Charge (kg/GWe-yr)

Ffssile Discharge (kg/GWe-yr)

Fissile Enrichmentb

(%HM)

U,O,

Net F i s s i l e Consumption

Burnup (Mh'D/kg m)

Annual (kg/GWe-yr)

Lifetimeg (kg/GWe)

Initial Loading (ST/GWe)

Requirement

Annual (ST/GWe)

Lifetime (ST/GWe)

D i s p e r s i b l e Resource-Based Fuels N a t u r a l U. no r e c y c l e

897 235U

852 2 3 5 U

249 235Uc 340 Puf

1261 235U

561 235U

59 23 5u= 183 Puf

MEU/Th, no r e c y c l e

2121 235U

1052 235U

336 235Uc 25 Puf 476 233U

MEU/Th,

2121 235U

250 235$

99 235Ud 30 Puf 685 233U

S l i g h t l y enriched

U. no r e c y c l e

U recycle 685 2 3 3 U

603 235U -340 Puf

25605 235U , - l o 2 0 0 Puf

164

156

4688

16

502 235U -183 Puf

17530 235U -5490 P u ~

257

114

3563

1.88 (20% i n U)

16

716 235U -25 Puf -476 233U

32629 235U -750 Puf -14280 233U

538

267

8281

1.65 (13% i n U)

16

151 235U -30 Puf 0 23%

6500 235U -900 Puf 0 2331)

538

102 233u -32 P u ~

4606 233U -960 Puf

0

0

0

338 235U -29 Puf

10699 235U -870 Puf

164

73

2281

105 235U -2 Puf

5204 235U -60 Puf

548

27f

1331e

0.71 1 1.0

7.5

38d

1640e

Denatured D i s p e r s i b l e Fuel

.Denatured

1648 233U

831 233U

729 233U 32 Puf

897 235U 378 Puf

399 235U 168 Puf

61 235p NU c o n t a i n i n g 0.3% Pu 197 Puf

2159 235U

191 235Uf

86 235Uf 2 Puf 750 23311

23 3U02/Th02,

U recycle LEU, U + Pu r e c y c l e

1.46 (12% i n U)

16

Energy-Center C o n s t r a i n e d Fuel 16

Reference Fuel HEU/Th. U r e c y c l e

750 233U

1.91 (93% i n U)

16

0

2331)

EA11 cases assume 75% c a p a c i t y f a c t o r . For f r e s h f u e l . :No c r e d i t . 250 kg minus 99 kg 235UIGWe-yr i s e q u i v a l e n t t o 63 ST minus 25 ST U,O,/GWe-yr; t h u s annual U308 r e q u i r e m e n t i s 63 :Excludes t r a n s 4 t i o n requirements and o u t - o f - c o r e i n v e n t o r i e s . 191 kg minus 86 k g 235U/GWe-yr i s e q u i v a l e n t t o 48 ST minus 21 ST U308/GWe-yr; t h u s annual U308 r e q u i r e m e n t i s 48 'No c r e d i t f o r e n d - o f - l i f e core.

0

233u

- 25=38 ST/GWe. - 21=27 ST/GWe.

P W W


4-34

I.o v)

t-

z W

=U W

0.9

3 0 W

a

0.8

Om rr)

3 W

2 0.7 la -I W

U

0.6

0

0.7

0.8

0.9

1.0

1.1

I .2

1.3

1.4

1.5

1.6

INITIAL 235U ENRICHMENT, % Fig. 4.3-1.

4.

Fuel U t i l i z a t i o n C h a r a c t e r i s t i c s f o r Enriched-Uranium-Fueled CANDU.

HEU/Th, U Recycle:

With 93% 235U-enriched uranium s t a r t u p and makeup, t h e

annual U308 makeup requirements a t near-equilibrium are about 27 ST/GWe f o r the 16,00&MWD/T burnup case. This n e t consumption o f U308 i s only 14% o f the LWR once-through c y c l e and 28x o f the LWR thorium c y c l e (see Cases A and J i n Table 4.1-1). However, t h e i n i t i a l core U308 requirement i s more than double t h a t of the CANDU s l i g h t l y enriched uranium cycle. I n addition, the t r a n s i t i o n t o e q u i l i b r i u m and the out-of-core inventory requirements, depending on the recycle turn-around time, can be very s i g n i f i c a n t . 1

5.

Denatured U/Th, U Recycle ( 2 3 3 U Makeup):

The i n i t i a l core 2 3 3 U inventory require-

ment i s about 1,650 kg/GWe, w i t h an annual n e t requirement o f about 100 kg 233U/GWe.

6. MEU/Th, U Recycle (235U Makeup): as t h a t f o r the standard thorium cycle (i.e.,

The i n i t i a l core requirement i s about t h e same HEU/Th cycle); however, the e q u i l i b r i u m n e t

U308 consumption i s s l i g h t l y increased. 7.

MEU/Th, No Recycle:

This c y c l e o p t i o n i s included t o i n d i c a t e t h a t r e c y c l e o f

the self-generated 233U i s advisable f o r the MEU/Th c.ycle.. The l i f e t i m e U308 requirement f o r t h e once-through MEU/Th c y c l e i s about 8,300 ST, which i s a f a c t o r o f 2.3 higher than t h a t f o r the once-through enriched-uranium c y c l e i n CANDU reactors.


4-35

References f o r Section 4.3 F--

I

iJ

L

1.

J. S. Foster and E. Critoph, "The Status o f t h e Canadian Nuclear Power Program and Possible Future Strategies," Annals of Nuclear Energy 2, 689 (1975).

2.

S. Banerjee, S. !?. Hatcher, A. D. Lane, t!. T a m and J. I.Veeder, "Some Aspects o f t h e Thorium Fuel Cycle i n Heavy-t!ater-Moderated Pressure Tubes Reactors ,I' ~ u c l . Tech. 34, 58 (1977).

3.

E. Critoph, S. Banerjee, F.

\l. Barclay, C. Hamel, M. S. Milgram and J. 1. Veeder, "Prospects f o r S e l f - S u f f i c i e n t E q u i l i b r i u m Thorium Cycles i n CAMDU Reactors," AECL-5501 (1 976).

4.

C. E. T i l l , E. 11. Bohn, Y. I . Chang and J. 6. van Erp, "A Survey o f Considerations Involved i n I n t r o d u c i n g CANDU Reactors i n t o t h e 1I.S.

ii -

5.

,I1

ANL-76-132 (January 1977).

Y. I.Chang, C. E. T i l l , R. R. Rudolph, J. R. Deen, and M. J. King, " A l t e r n a t i v e Fuel Cycle Options: Performance C h a r a c t e r i s t i c s and Impact on Nuclear Power Growth (Sept. 1977). P o t e n t i a l ,'I "-77-70

L

, ,

c

c

,'


4-36

4.4.

GAS-COOLED THERMAL REACTORS

J. C. Cleveland Oak Ridge National Laboratory 4.4.1.

High-Temperature Gas-Cooled Reactors

The High-Temperature Gas-Cooled Reactor (HTGR) i s another candidate f o r implemen ing a l t e r n a t e f u e l c y c l e options, p a r t i c u l a r l y t h e denatured 233U cycle. U n l i k e o t h e r r e a c t o r types t h a t g e n e r a l l y have been optimized f o r e i t h e r LEU o r mixed oxide ( P u / * ~ * U ) fuel, t h e HTGR has a design based on u t i l i z a t i o n o f a thorium f u e l cycle, and although currentdesign HTGRs may n o t meet p o t e n t i a l proliferation-based f u e l c y c l e r e s t r i c t i o n s , t h e r e f e r ence design involves both 232Th and 233U, which are t h e primary m a t e r i a l s i n t h e denatured f u e l cycle. I n c o n t r a s t t o t h e fuel f o r water-cooled reactors and f a s t breeder reactors, t h e f u e l f o r HTGRs i s n o t i n t h e form of metal-clad rods b u t r a t h e r i s composed o f coated f u e l p a r t i c l e s bonded together by a g r a p h i t e m a t r i x i n t o a f u e l s t i c k . The coatings on t h e i n d i v i d u a l f u e l p a r t i c l e s provide f i s s i o n - p r o d u c t containment. The f u e l s t i c k s are loaded i n f u e l holes i n hexagonal g r a p h i t e f u e l blocks. These blocks a l s o c o n t a i n hexagonal arrays o f coolant channels through which t h e helium flows. I n t h e conventional HTGR t h e fuel p a r t i c l e s are o f two types: f i s s i l e p a r t i c l e s c o n s i s t i n g o f UC2 kernels coated w i t h l a y e r s o f pyrocarbon and s i l i c o n carbide; and f e r t i l e p a r t i c l e s c o n s i s t i n g o f Tho2 kernels coated o n l y w i t h pyrocarbon. The pyrocarbon coating on t h e f e r t i l e p a r t i c l e s can be burned off w h i l e t h e S i c c o a t i n g on t h e f i s s i l e p a r t i c l e s cannot. Therefore t h e two p a r t i c l e types can be p h y s i c a l l y separated p r i o r t o any chemical reprocessing. As i n d i c a t e d i n Chapter 5, h o t demonstrations o f t h e hcad-ecd processing operations unique t o t h i s r e a c t o r f u e l , t h e crushing and burning o f t h e f u e l elements, t h e mechanical p a r t i c l e separation, and t h e p a r t i c l e crushing and burning are needed t o ensure t h a t low-loss reprocessing can take place. An inherent f e a t u r e o f t h e HTGR which r e s u l t s i n uraniurri resource conservation i s i t s high (% 40%) thermal e f f i c i e n c y . A l l e l s e being equal, t h i s f a c t alone r e s u l t s i n a 15% reduction i n uranium resource requirements compared t o LWRs, which achieve a 34% thermal e f f i c i e n c y . This l a r g e r thermal e f f i c i e n c y a l s o leads t o reduced thermal discharges t h a t provide s i g n i f i c a n t s i t i n g advantages f o r HTGRs, e s p e c i a l l y i f many react o r s a r e t o be deployed i n c e n t r a l l o c a t i o n s such as energy centers. Other f a c t o r s inherent i n HTGR design t h a t lead t o improved

Ug8u t i l i z a t i o n

due

t o t h e improved neutrol: economy are: 1.

Absorption o f only % 1.6% o f t h e neutrons by HTGR p a r t i c l e ccatings, g r a p h i t e moderator, and helium coolant, compared t o ar: absorption o f % 5.â&#x201A;Ź% o f t h e neu-

L


4-37

t r o n s i n t h e Z i r c a l o y cladding and t h e coolant o f conventional PWRs ( ~ 4 % of all neutron absorptions i n PWRs r e s u l t from hydrogen absorption).

2.

Low 233Pa burnout due t o t h e low (7-8 W/cm3) power density.

c

t h e g r a p h i t e moderator and t h e ceramic f u e l l a r g e l y m i t i g a t e t h e consequences o f HTGR loss-

L

as i n conventional LWRs o r FBRs since t h e heat capacity o f t h e core i s maintained, theref o r e a l l o w i n g considerable time t o i n i t i a t e actions designed t o provide a u x i l i a r y core cooling.

r-

The combination o f low power d e n s i t y and l a r g e core heat capacity associated w i t h of-coolant accidents.

i]

Loss o f c o o l i n g does n o t lead t o severe conditions n e a r l y as q u i c k l y

The HTGR o f f e r s a near-term p o t e n t i a l f o r r e a l i z a t i o n o f tmproved U308 u t i l i z a t i o n . The 330-MWe F o r t St. Vrain p l a n t has been under s t a r t - u p f o r several years w i t h a c u r r e n t l i c e n s e d power l e v e l o f 70% and t h e p l a n t has operated a t t h e 70% power l e v e l f o r l i m i t e d periods. A data c o l l e c t i o n program i s p r o v i d i n g feedback on problem areas t h a t are becoming apparent d u r i n g t h i s s t a r t - u p p e r i o d and w i l l serve as t h e basis f o r improvements i n t h e

L h

comsercial p l a n t design. An advantage o f t h e HTGR steam c y c l e i s t h a t i t s commercialization could lead t o l a t e r commercialization o f advanced gas-cooled systems based on t h e HTGR technology.

These

i n c l u d e the HTGR gas t u r b i n e system which has a h i g h thermal e f f i c i e n c y o f 45 t o 50% and t h e VHTR (Very High Temperature Reactor) system f o r high-temperature process heat appl i-cation. Mass balance c a l c u l a t i o n s have been performed by General Atomic f o r several a1t e r n a t e HTGR f u e l cycles,l and some a d d i t i o n a l c a l c u l a t i o n s c a r r i e d o u t a t ORNL have v e r i f i e d c e r t a i n

GA r e s u l t s . 2

T h e i r r e s u l t s f o r t h e f o l l o w i n g f u e l cycles are presented here:

D i s p e r s i b l e Resource-Based Fuels

L

1.

LEU, no recycle. a. b.

Carbon/uranium r a t i o (C/U) = 350. C/U = 400, optimized f o r no recycle.

2.

MEU/Th (20% 235U/U mixed w i t h 232Th), C/Th = 650, no recycle.

3.

MEU/Th (20% 235U/U), C/Th = 306 f o r i n i t i a l core, C/Th = 400 f o r r e l o a d segments, 23311

recycle.

D i s p e r s i b l e Denatured Fuel

.4.

MEU/Th (15% 233U/U), C/Th = 274/300 ( i n i t i a l core/reload segments), optimized f o r uranium r e c y c l e

__

( 2 3 % t 2351)).

Energy-Center-Constrai ned Fuel

f;

t

5.

Pu/Th, C/Th = 650 (batch-loaded core).

Reference Fuels 6.

HEU(235U)/Th, C/Th = 214/238 ( i n i t i a l core/reload segments), no recycle.

7.

HEU(233U)/Th, C/Th = 150, high-gain design, uranium recycle.

8.

HEU(235U)/Th, C/Th = 180/180 ( i n i t i a l core/reload segments), uranium r e c y c l e (from r e f . 3).


4-38

A l l o f t h e above f u e l cycles are f o r a 3360-MWt, 1344-MWe HTGR w i t h a core power den-

s i t y o f 7.1 Wt/cm3. Table 4.4-1 provides a summary, obtained from t h e d e t a i l e d mass balance information i n r e f . 1, o f the conversion r a t i o , f i s s i l e requirements, f i s s i l e discharge, and Cases 1-a and 1-b i n v o l v e t h e use o f LEU f u e l w i t h U3O8 and separative work requirements. an e q u i l i b r i u m c y c l e enrichment o f 7.4 w/o and 8.0 w/o, respectively. Case 1-b would be

il

p r e f e r r e d f o r no-recycle conditions. I n Case 2 thorium i s used w i t h 20% 235U/U (MEU/Th) f o r no-recycle conditions. Note t h a t w h i l e t h e i n i t i a l U308 and f i s s i l e loading requirements are higher f o r t h e MEU/Th case than f o r the LEU cases, due t o the l a r g e r thermal absorption cross s e c t i o n o f thorium and t h e p a r t i a l unshielding o f t h e 238U resonances r e s u l t i n g from i t s reduced density, t h e cumulative This r e s u l t s from t h e high burnup

U308 requirements are s l i g h t l y l e s s f o r t h e MEU/Th case.

a t t a i n a b l e i n HTGRs and t h e res+ltant l a r g e amount o f bred

23%

which i s burned i n s i t u .

Other converter and advanced converter reactors (LWRs , SSCRs , and HWRs ) t y p i c a l l y r e q u i r e l e s s U308 f o r t h e LEU case than f o r t h e MEU/Th case w i t h no recycle.

L

c

Case 3 a l s o uses t h e MEU/Th feed b u t w i t h r e c y c l e o f 233U. The unburned 235U and plutonium discharged i n t h e denatured 235U p a r t i c l e s i s n o t recycled. The bred 233U r e covered from t h e f e r t i l e p a r t i c l e , however, i s denatured, combined w i t h thorium, and recycled. I n the c a l c u l a t i o n s f o r a l l cases i n v o l v i n g r e c y c l e o f denatured 233U, GA assumed t h a t an i s o t o p i c mix o f 15% 233U and 85% 238U provided adequate denaturing. Due t o t h e h i g h burnup and t h e f a c t t h a t the thermal-neutron spectrum i n HTGRs peaks near t h e 239Pu and 241Pu resonances, a l a r g e amount o f t h e f i s s i l e plutonium bred i n t h e denatured f u e l i s burned i n s i t u , thus r e s u l t i n g i n t h e low f i s s i l e plutonium content o f t h e f u e l a t discharge. Considerable

238U

s e l f - s h i e l d i n g i s obtained by t h e lumping o f t h e 238U i n t h e coated p a r t i c l e

kernels. Studies are c u r r e n t l y underway a t GA concerning t h e use o f l a r g e r diameter f i s s i l e p a r t i c l e s , thereby lowering t h e 238U resonance i n t e g r a l and, consequently, t h e amount o f bred plutonium d i ~ c h a r g e d . ~ Case 4 employs a denatured 233U feed and includes uranium recycle.

It represents a

f e a s i b l e successor t o Case 3 once an exogenous source o f 233U i s available. Case 5 involves Pu/Th Fuel.

Since no

238U

i s present i n t h e core, no plutonium i s

bred; o n l y 233U i s bred.

This r e a c t o r has g r e a t l y reduced requirements f o r c o n t r o l poison, r e s u l t i n g i n enhanced neutron economy. This r e s u l t s from the f a c t t h a t t h i s Pu/Th HTGR e s s e n t i a l l y achieves t h e "Phoenix" f u e l c y c l e e f f e c t , i.e., the decrease i n 239Pu content i s l a r g e l y compensated f o r by buildup o f 241Pu from 240Pu capture and by buildup o f 233U from 232Th capture, r e s u l t i n g i n a n e a r l y constant r a t i o o f f i s s i l e concentration t o 240Pu

concentration.

Therefore the f u e l r e a c t i v i t y i s r e l a t i v e l y constant over a long burnup

period, reducing the need f o r c o n t r o l poison. i.e.,

This allows the core t o be batch loaded;

the e n t i r e core i s reloaded a t approximately 5-yr i n t e r v a l s .

This r e l o a d scheme minimizes down time f o r r e f u e l i n g and eliminates problems o f power sharing between f u e l elements o f d i f f e r e n t ages. any cycle.

Furthermore, i t allows easy conversion t o a U/Th HTGR a f t e r

I t i s important t o note t h a t t h e Pu/Th case presented i n Table 4.4-1

i s not

c L f -i

Li

I '

' C


Table 4.4-1.

Fuel U t i l i z a t i o n C h a r a c t e r i s t i c s f o r HTGRs Under Various Fuel Cycle Options

f n i t i a l Core Requirements:

HM

-

E s u i l i b r i u m Cycleb Discharge o f Fissile Nonrecycl able Makeup F i s s i l e Material (kg/GWe-yr) (kg/GWe-yr)

Conversion R a t i o ( 1 s t Cy./Eq. Cy.)

Fissile Inventory (kg/GWe)

Loading (MT/GWe)

1-a, LEU. no recycle, C/U = 350

0.580/0.553

901 23511

24.6 U

608 235U

1-b, LEU, no recycle, C/U = 400

0.557/0.526

819 235U

21.6

U

2, MEU(PO% 235U)/Th.

0.630/0.541

1077 235U

3 MEU(2O% 235U)/Th,f 2 3 3 ~recycle, C/Th = 3O6/40Og

0.682/0.631

4, MEU(l5X 233U)/Th,f

U308 Requirementc (ST/GWe)

Separative Work Requiremento (lo3 k g SWU/GWe)

Initial

30-yr T o t a l f o r CF of;, 65.9%/75%

Initial

113 235U 69 Puf

217

4272/4860

142

3319/3781

576 235U

77 2 3 % 52 Puf

197

4040/4594

130

3188/3629

5.4 U 20.2 Th

551 235U

47 2% 74 23311 22 Puf

274

3967/4515

249

3640/4143

1474 235U

7.4 u 27.5 Th

397 235u

65 235U 36 Puf

371

3229/3666

340

2933/3361

0.824/0.764

1168 233U

_bisperslble Denatured Fuel 246 23311 35 Puf 7.9 U 30.7 Th

0

0

0

0

5, Pu/Th. C/Th = 650

0.617/0.617

3153 Pufh

12.2 Th

102 Puf 97 2 3 %

0

0

0

0

6. HEU(235U)/Th, no r e c y c l e . C/Th = 214/238

0.723/0.668

1358 235U

1.5 U 37.2 Th

508 235U

49 235u 183 233U 1 Puf

345

3864/4395

344

3858f 4387

7, HEU(233U)/Th, h i / g a i n . U recycle, C/Th = 150

0.915/0.859

1395 23% 139 235U

2.0 u 53.0 Th

120 2330 12 235U

0

0

0

0

Case, Fuel Type

-.

-

30-yr T o t a l f o r CF os 65.9%/75%

D i s p e r s i b l e Resource-Based Fuels

no recycle, C/Th = 650

U recycle, C/Th = 274/300

Energy-Center-Constrained Fuel 630 Puf Reference Fuels

-

:

8. HEU(235U)/Th, hi/gain, U recycle, C/Th = 180/180

f

25u

sodBk

/2280

50dsk

~

: I n i t i a l c y c l e l a s t s one calendar y e a r a t 60% capacity f a c t o r . _ E a u l l i b r i u m c y c l e c a p a c i t y f a c t o r i s 72%. >sumes 0.2 40t a i l s . -Value preceding s l a s h i s f o r an average 30-yr capacity f a c t o r o f 65.9; v a l u e f o l l o w i n g s l a s h I s f o r a constant c a p a c i t y f a c t o r o f 75%. o c r e d i t taken f o r end o f l i f e core. 23511 frm MEU p a r t i c l e o r Pu r e c y c l e d i n Case 3; a l l U r e c y c l e d i n Case 4, b u t no Pu recycled. ! I n i t i a l c o r e h e l o a d sement. :Core i s batch loaded; i n i t i a l l o a d provides f i s s i l e m a t e r i a l f o r 45 y r o f operation. jReference f u e l s a r e considered o n l y as l i m i t i n g cases. I n i t i a l c y c l e l e n g t h i s 1.6 yr. 'Numbers shown a r e f o r a c a p a c i t y f a c t o r o f 75%.

40

/2278


4-40

optimized f o r high conversion; r a t h e r i t i s a Pu burner designed f o r low fuel c y c l e costs. A Pu/Th case designed f o r high 233U production would have a C/Th r a t i o f o r t h e e q u i l i b r i u m c y c l e o f 4 3 0 r a t h e r than 650 as i n Case 5 ( r e f . 5). I n Case 6 t h e feed i s f u l l y enriched (93%) uranium and thorium and no r e c y c l e i s allowed. Such a system would provide the means f o r generating a p o t e n t i a l s t o c k p i l e o f 233U i n t h e absence o f reprocessing c a p a b i l i t y . I f 233U r e c y c l e i s n o t contemplated, t h e economical optimum once-through c y c l e would have a lower thorium loading (C/Th = 330). Case 7 involves the use o f h i g h l y enriched

23%

and uranium recycle.

The heavy fer-

t i l e loading (C/Th = 150) r e s u l t s i n t h e high conversion r a t i o (and h i g h i n i t i a l f i s s i l e loading requirement) shown i n Table 4.4-1. Case 8 involves t h e use o f f u l l y enriched (93%) uranium and thorium designed f o r r e c y c l e conditions. This i s included as t h e pre-1977 reference high-gain HEU(235U)/Th r e c y c l e case f o r comparison with the o t h e r above cases. Both GA and ORNL have performed mass balance c a l c u l a t i o n s f o r an HEU(235U)/Th f u e l c y c l e w i t h uranium recycle.2,6 These c a l c u l a t i o n s were f o r a 1160-MWe p l a n t w i t h a power d e n s i t y of 8.4 Wt/cm3 and a C/Th r a t i o f o r t h e f i r s t core and r e l o a d cycles o f 214 and 238 respectively. The GA r e s u l t s i n d i c a t e cumulative U308 and separative work requirements ( f o r a capacity f a c t o r o f 75% and an assumed t a i l s enrichment O f 0.2 w/o) o f 2783 ST U308/ GWe and 2778 kg SWU/GWe, respectively. The corresponding r e s u l t s f o r t h e ORNL c a l c u l a t i o n s are 2690 ST U308/GWe and 2684 kg SWU/GWe. As can be seen, t h e agreement i s f a i r l y good. Comparison o f these r e s u l t s w i t h the same case w i t h o u t r e c y c l e (Case 6, Table 4.4-1)

shows

a U308 savings of ~ 3 8 % i f uranium i s recycled. I t i s conventional t o compare 30-yr cumulative U308 and separative work requirements

f o r d i f f e r e n t r e a c t o r types on a per GWe basis w i t h an assumed constant capacity factor. The r e s u l t s reported i n Table 4.4-1 were generated f o r an assumed v a r i a b l e capacity f a c t o r which averaged 65.9% over t h e 30-yr l i f e .

To f a c i l i t a t e comparison w i t h U3O8 requirements

i n o t h e r sections o f Chapter 4, estimated 30-yr requirements f o r a constant capacity f a c t o r o f 75% have a l s o been included i n t h e table. These values were obtained by applying a f a c t o r o f 0.750/0.659

t o t h e c a l c u l a t e d requirements f o r t h e v a r i a b l e capacity factor.

Obviously t h i s technique i s an approximation b u t i t i s f a i r l y accurate. The 30-yr r e q u i r e ments f o r a 75% capacity f a c t o r f o r Case 8 were e x p l i c i t l y c a l c u l a t e d and n o t obtained by t h e above e s t i m a t i n g procedure. As i s i n d i c a t e d i n Table 4.4-1,

t h e MEU(2O% 235U)/Th no-recycle case i s more r e -

source e f f i c i e n t than t h e LEU no-recycle case.

This r e s u l t s from t h e h i g h exposure a t t a i n -

able i n HTGR f u e l s and t h e high i n s i t u u t i l i z a t i o n o f

L

I n water reactors, t h e oncethrough MEU(2O% 235U)/Th c y c l e requires s i g n i f i c a n t l y more U308 than t h e once-through LEU cycle.

233U.

Thus MEU(20% 235U)/Th f u e l s i n HTGRs are an a t t r a c t i v e o p t i o n f o r stowaway cycles

i n which 233U i s bred f o r l a t e r use.

hâ&#x20AC;&#x2122;


4-41

4.4.2.

Pebble-Bed High-Temperature Reactors

A second high-temperature gas-cooled thermal r e a c t o r t h a t i s a possible candidate

L

f o r t h e denatured 233U fuel c y c l e i s the Pebble-Bed Reactor (PBR). Experience w i t h PBRs began i n Augusta 1966, i n JGlich, West Germany, w i t h the c r i t i c a l i t y of t h e Arbeitgemeinshaft Versuch Reaktor (AVR),

a 46-MGlt r e a c t o r t h a t was developed t o g a i n knowledge and experience

i n t h e c o n s t r u c t i o n and operation o f a high-temperature helium-cooled r e a c t o r f u e l e d w i t h I

spherical elements comprised o f carbon-coated fuel p a r t i c l e s . This experience was intended t o serve as a basis f o r f u r t h e r development o f t h i s concept i n West Germany. Generation

1

o f e l e c t r i c i t y w i t h the AVR began i n 1967. I n a d d i t i o n t o generating e l e c t r i c power, the AVR i s a t e s t f a c i l i t y f o r i n v e s t i g a t It a l s o i s a s u p p l i e r of high-burnup high-

i n g t h e behavior o f spherical fuel elements.

temperature r e a c t o r f u e l elements f o r the West German f u e l reprocessing development work. The c o n t i n u a t i o n o f t h e PBR development i n i t i a t e d by the AVR i s represented by the THTR a t Schmehausen, a r e a c t o r designed f o r 750 MWt w i t h a n e t e l e c t r i c a l output o f 300 MW. Startup o f t h e THTR i s expected about 1980.

I Ld

I

Table 4.4-2.

conservation o f uranium resources due t o Power, Qt

3000 MWt

Power d e n s i t y Heating o f he1ium

5 MW/m3 25b985 O C

He1ium in l e t pressure P 1a n t e f f ic i ency , Qe/Qt Height o f b a l l f i l l Radius B a l l packing I n n e r f u e l i n g zone: Outer r a d i u s Number o f b a l l f l o w channels R e l a t i v e residence time Outer f u e l i n g zone:

rarus

ie i--

1

t'

The PBR concept o f f e r s favorable

PBR Core Design

Outer Number o b a l l flow channels Re1a t i ve residence time Top r e f l e c t o r :

40 atm 0.40 550 cm 589 cm 5394 b a l l s/m3 505 cm 4 9/9/9/9 589 cm

1

13

Thickness Graphite d e n s i t y Bottom r e f l e c t o r :

200 0.32

Thickness Graphite d e n s i t y Radial r e f l e c t o r :

150 1.60

Thickness Graphite d e n s i t y

100 1.60

i t s low f i s s i l e inventory requirements and t o t h e high burnup t h a t i s achievable i n

PBR elements. This has been demonstrated by the analysis o f several once-through cycles c a l c u l a t e d f o r the PBR by a physics design group7 a t KFA J u l i c h , West Germany, and summarized here.

The r e a c t o r core de-

sign used f o r the study i s described i n Table 4.4-2 Various f u e l element types were considered, d i f f e r i n g by the coated p a r t i c l e types used and by t h e heavy metal loading. The basic f u e l element design i s shown i n Table 4.4-3, the coated p a r t i c l e designs are described i n Table 4.4-4, and the compositions o f the various f u e l e l e ment types are given i n Table 4.4-5. The once-through cycles considered are described below, w i t h t h e core compositions of each given i n Table 4.4-6. Case 1, LEU,

Low-enriched uranium

i s loaded i n t o t h e coated f u e l p a r t i c l e s . The r a d i a l power p r o f i l e i s f l a t t e n e d by varying the enrichment i n t h e i n n e r and


4-42

Table 4.4-3.

PBR Fuel Element Design

o u t e r r a d i a l core zones.

The enrichment

o f t h e i n n e r zone i s 7.9 at.% and t h a t of B a l l diameter

6 cm

Thickness o f g r a p h i t e she1 1 Graphite density

0.5 cm 1.70 g/cm3

the o u t e r zone i s 11.1 at.%. Case 2, MEU/Th. (U + Th)02 fuel w i t h 201 enriched uranium i s Toaded i n t o the coated f u e l p a r t i c l e s .

The heavy metal

loading i n t h e MEU/Th f u e l element i s between t h a t o f t h e THTR and AVR elements.

As i n

Case 1, t h e r a d i a l power i s f l a t t e n e d by t h e choice o f f i s s i l e loading o f t h e elements i n t h e i n n e r and o u t e r r a d i a l core zones, 6.85 and 11.4% r e s p e c t i v e l y . The coated p a r t i c l e s would r e q u i r e some development and testing. Case 3, Seed and Breed MEU/Th. i n t o seed elements and Tho,

(U

+ Th)O, f u e l w i t h 2 o I enriched uranium i s loaded

i s loaded i n t o breed elements.

By thus separating t h e seed and

breed elements, 236U bred i n t o t h e seed elements w i l l n o t have contaminated t h e 233U produced i n t h e breed elements i n case r e c y c l e i s opted f o r l a t e r . Graphite b a l l s are added t o t h e i n n e r core zone t o a d j u s t t h e carbon/heavy metal r a t i o (C/HM)

t o t h a t o f the outer

zone. The heavy metal loading o f 6 g HM/ball i n t h e seed elements i s e s s e n t i a l l y t h e same as i n t h e AVR. The f e a s i b i l i t y of a considerably heavier l o a d i n g o f t h e breed e l e ments, 16.54 g HM/ball, i s c u r r e n t l y being tested. Case 4, HEU/Th.

(U + Th)O,

f u e l w i t h 93% enriched uranium i s loaded i n t o t h e coated

f u e l p a r t i c l e s . The coated p a r t i c l e and f u e l element designs are e s s e n t i a l l y i d e n t i c a l t o those o f THTR f u e l elements, which have been l i c e n s e d and a r e being manufactured. The o n l y m o d i f i c a t i o n i s t h e f i s s i l e loading. Again t h e f i s s i l e loading o f the elements i n t h e i n n e r and o u t e r r a d i a l core zones i s v a r i e d t o f l a t t e n the r a d i a l power d i s t r i b u t i o n , t h e i n n e r zone f i s s i l e loading being 6.23% o f t h e heavy metal and t h e o u t e r zone f i s s i l e loading being 10.9%. (U + Th)02 fuel w i t h 93% enriched uranium i s loaded Case 5 , Seed and Breed HEU/Th. i n t o seed elements and breed elements c o n t a i n Tho, only. The r a d i a l power p r o f i l e i s f l a t tened by t h e choice o f t h e mixing f r a c t i o n o f seed and breed b a l l s i n t h e i n n e r and o u t e r r a d i a l core zones, and g r a p h i t e b a l l s are added t o t h e i n n e r zone t o adapt t h e C/HM r a t i o t o t h a t o f t h e outer zone.

I n t h e seed elements t h e HEU i s mixed w i t h some Tho,

i n order

t o achieve a prompt negative Doppler c o e f f i c i e n t . Again t h e heavy metal loading o f t h e balas i s e s s e n t i a l l y t h e same as t h a t i n t h e AVR and t h e f e a s i b i l i t y o f t h e l o a d i n g o f the breed elements i s being tested. The mass f l o w data f o r t h e e q u i l i b r i u m c y c l e o f each o f t h e f i v e cases are presented i n Table 4.4-7. The high thermal cross sections o f 239Pu, ,â&#x20AC;&#x2DC;+OPu and 241Pu, t h e s o f t spectrum, and t h e low s e l f - s h i e l d i n g o f t h e f u e l element design l e a d t o a very high i n - s i t u u t i l i z a t i o n o f t h e f i s s i l e plutonium (95% f o r t h e MEU/Th cycles).

I n addition,

t h e h i g h burnup r e s u l t s i n t h e low discharge plutonium f i s s i l e f r a c t i o n s shown i n Table 4.4-7.

The b u i l d u p o f plutonium isotopes i n t h e MEU/Th c y c l e i s shown i n Fig. 4.4-1.

L

c


4-43

PBR Coated P a r t i c l e Design

Table 4.4-4.

i,

Kernel Type

I I1 I11

Materi a1

Diameter

U/Th02 U/Th02

400

9.50

85/30/80

1.0/1.6/1.85

400

UOZ

800

9.50 9.50

50180 110180

1.011.85 1.011.85

Table 4.4-5.

L

N1 t?Z

I I1 I1

52 til

I1

82

I1

L1

I11 I11 Carbon

:-

L2

i

G a

Thicknesses

Densities ( g/cm3 1

(lJm)

Composition o f PBR Fuel Elements

I

s1

i

Density ( gIcfi13 1

Type of Coated P a r t i c l e a

I d e n t i f i c a t i an

t

Carbon Coatings

Heavy Metal Loading (glball)

Moderati on Ratio ('iC/"Hff)

11.24 8.G7

â&#x201A;Ź.O

325 458 617

6.0 20.13 16.54

629 180 220

9.88

380

11.70

320

See Table 4.4-4.

-..

Table 4.4-6.

L

L L

L L

Case

1, LEU 2, MEU/Th 3, Seed and Breed MEU/Th 4, HEU/Th 5, Seed and Breed HEU/Th U

i

See Table 4.4-5.

Composition o f PBR Core Regions Used i n Mass Flow Calculations Inner Core

Outer Core

Fuel Element Type' Fissile hading (Fractional Mixing) ("fis/"HM)

Fuel F i s s i l e Loading Element T~~~ (Fractional Mixing) ( NfisINHM)

11 (1.0) 1.12 (1.0) S2 0.485 B2 0.305 G (0.210) M1 (1.0) S1 (0.40) B1 (0.39) G (0.21)

t

0.079

L2 (1.0)

0.111

0.0685 0.20

M2 (1.0) S2 (0.765) 82 (0.235)

0.114 0.20

0.0623 0.27

M1 (1.0) S1 (0.69) B1 (0.31)

0.109 0.27


.

T a b l e 4.4-7.

... .

. ..-..-.- ...

..

I

.

.

.. .

~

....

~

~.

" _ I

F u e l U t i l i z a t i o n C h a r a c t e r i s t i c s f o r E q u i l i b r i u m C y c l e s o f PBRs Under V a r i o u s Fuel O p t i o n s a w i t h No Recycle

---

Fuel Type

Case

Conversion Ratio

Fuel Elementsb

0.58

L1 + L2

Loading (kg/GWe-yr)

Discharge (kg/GWe-yr 1

Isotopic Fraction o f O i scharge Pu

Burnup (HWO/kg HM)

D i s p e r s i b l e Resource-Based Fuels 1

LEU

575 23511 6168 2 3 e U

93 23511 80 236U 5719 238U

6743 Utot.

5892 Utot. 42 26 21 24

239(P~,Np) 240Pu 24'Pu

242Pu

100

0.37 0.23 0.19 0.21

2 3 9 1 P ~ . N. o. ) ~

240Pu 241Pu

242Pu

113 Putot' 2

MEU/Th

0.58

M2

4158 Th

534 23511

-

2163 238U 2697 Utot*

91 22 39 79 1965

233(U,pa) 23411 23511 2360

239(Pu,Np) 2'0Pu 24'Pu 242Pu

36 Putot* Seed & Breed MEU/Th

0.56

52

2190 238U

30 235U 81 236U 1982 2eaU

2730 Utot*

2093 Utot'

540 235U

P

238U 2195 Utot. 9 9 5 13

3

100

3881 Th

P P

0.25 239(P~,Np) 0.26 240Pu 0.14 '"Pu 0.34 242Pu Puf/Putot'

= 0.39 201

0.24 239(Pu,Np) 0.25 240Pu 0.14 241Pu 0.38 242Pu Puf/Putot' 82

4 70 Th

3881 Th 82 22 4 1

233(U,Pa) 23411 235u

23611

108 Utot'

= 0.38

35

.


Reference F u e l s HEU/Th

0.59

M1

6302 Th

-495 235u 38 23% 533 utot*

Seed i3 Breed HEU/Th

0.58

s1

1287 Th

--

5794 Th 128 38 23 73 30 292

100

233(U,Pa) 234U

235U 236u

238U

Utot*

1.166 Putot*

0.23 23g(Pu,Np) 0.21 24OPu 0.13 Z41Pu 0.44 242PU Puf/Putoto = 0.36

1185 Th

243

496 235u ;8 23% 534

B1

4983 Th

f P VI

155 Utot. 0.227 239(Pu.Np) 0.257 240Pu 0.120 241Pu 0.500 2'2Pu 1.106 Putot' 4594 Th 91 29 5 1 126

:Calculated f o r 1000-We p l a n t o p e r a t i n g a t 75% c a p a c i t y . See Tables 4.4-3 t h r o u g h 4.4-6 f o r d e s c r i p t i o n s o f cases and f u e l elements. =Reference f u e l s are considered o n l y as l i m i t i n g cases.

233(U,Pa) 234U 23511 23%

Utot'

0.21 239(PuBNp) 0.23 24oPu 0.11 2"Pu 0.45 242Pu Puf/Putot* = 0.32 48


4-46 As can be seen, the 239Pu content peaks a t

GRlBRLL 1

30 MWD/kg,decreasing thereafter. The higher Pu isotopes tend t o peak a t higher U-239 U-240 U-242 burnups so t h a t a t discharge 242Pu doniio . o l G , nates. Compared % t o an LLJR with LEU , ftlel, the PBR v:ith HEU/Th fuel discharges only 8% ,oo nwOlac 50 HI)(II.TLUC"I.l.RuE1rEW.rT. as much f i s s i l e plutonium. Furthermore, the Fig- 4-4-10 Buildup of the Plutoniur f i s s i l e fraction o f the discharged plutonium Isotopic Composition in the MEU/Th Fuel. i s only 39% compared t o 71% f o r an LWR. '~r

Table 4.4-8 presents U308 requirements of the various once-through cycle^.^,^ The 30-yr cumulative U308 demands f o r the MEU/Th once-through cycle and the HEU/Th oncethrough cycle were determined by e x p l i c i t 30-yr calculations.8 The 30-yr cumulative U308 demands f o r the LEU, the seed-and-breed MEU/Th and the seed-and-breed HEU/Th cycles were determined from the U308 demand f o r the equilibrium cycles and estimates of the inventory of the startup core and of the requirements f o r the approach t o equilibrium.8 As can be seen from Table 4.4-8, from the viewpoint of U308 utilization f o r oncethrough cycles in the PBR, LEU fuel i s the l e a s t favorable and HEU/Th fuel i s the most favorable w i t h MEU/Th fuel having a U308 utilization between HEU/Th and LEU fuel. I t should

be noted t h a t the cases presented in Table 4.4-8 do not include recycle of the bred f i s s i l e material. Under these no-recycle constraints the MEU/Th cases have a 30-yr U3O8 demand comparable to a PWR operating with uranium and self-generated P u recycle (see Case F, Table 4.1-3). T h u s i f recycle were performed with the MEU/Th PBR cases, significantly less U308 would be required than f o r the PWR with U and Pu recycle. One option f o r the recycle i n the seed-and-breed MEU/Th PBR case would be t o cycle the f e r t i l e b a l l s back into the feed stream (without reprocessing) for an additional pass through the pebble bed i f the irradiation behavior of the f e r t i l e balls permits. Table 4.4-8. U308 Requirements f o r Once-Through PBR Cyclesa ~-

Case 1,

Case 2,

LEU

11E U / T h

Eauilihriun cycle 143 U308 demand, ST/GWE-yr

135

30-year cumulative demand,b ST/GWE

U308

45OOc

4184d

Case 3 , Seed and Ereed MEU/Th 137

4200'

Case 4 , HEU/Th

Case 5, Seed and Greed HEU/Th

126

126

4007d

40OOc

'The basis f o r these requirements i s a lOCO-H'.le plant operating a t 752 capacity factor f o r 30 years; t a i l s composition i s assumed to be 0.2 w/o. 'Assures no recycle. c

Estimated value; could d i f f e r from an e x p l i c i t 30-yr calculation by 5 3%.

dExplicit 30-yr calculation.

L L


4-47

L

References f o r Section 4.4

1.

Reactor Design Characteristics and Fuel Inventory Data, compiled by HEDL, September 1977.

2.

Thorium Aseessment Study Quarterly Progress Report for Second Quarter Fiscal 1977, ORNL/TM-5949, Oak Ridge National Laboratory (June 1977).

3.

L e t t e r , A. J. Neylan, Manager, HTGR Generic Technology Program, t o K. 0. Laughon, Jr., Chief, Thermal Gas-Cooled Reactor Branch, DOE/NRA, "Technical Data Package f o r NASAP," March 3, 1978.

4.

M. H. t i e r r i l l and R. K. Lane, "Medium Enriched Uranium/Thorium Fuel Cycle P a r a m t r i c Studies f o r the HTGR," General Atomic Report GA-A14659 (Deceniber 1977).

5.

L e t t e r , R. F. Turner, Manager, Fuel Cycles and Systems Department, General Atomic Company, t o T. C o l l i n s , DOE/NPD, May 8, 1978.

6.

Thorium Assessment Study Quarterly Progress Report f o r Third Quarter Fiscal 1977, ORNL/TM-6025, Oak Ridge National Laboratory.

I

7.

E. Teuchert, e t al., "Once-Through Cycles i n the Pebble-Bed HTR," Trans. Am. Nucl. SOC., 27, 460 (1977). (Also published as a KFA-Julich r e p o r t Jul-1470, December 1 9 7 n .

i;

8.

L e t t e r , E. Teuchert ( I n s t i t u t F i r Reaktorenturcklung, Der Kernforschungsanlage J u l i c h GmbH) t o J. C. Cleveland (ORNL), "Once-Through Cycles i n the PebbleBed HTR," May 19, 1978.

i; 1

i


4-48

LIQUID-METAL FAST BREEDER REACTORS

4.5.

T. J. Burns Oak Ridge National Laboratory A p r e l i m i n a r y a n a l y s i s o f t h e impact o f denatured f u e l on breeder r e a c t o r s was

performed by Argonne National Laboratory,

Hanford Engineering Development Laboratory,2

and Oak Ridge National Laboratory3 f o r a v a r i e t y o f f i s s i l e / f e r t i l e f u e l options.

The a n a l y s i s concentrated p r i n c i p a l l y on oxide-fueled LMFBRs due t o t h e i r advanced s t a t e o f development r e l a t i v e t o o t h e r p o t e n t i a l breeder concepts. Table 4.5-1

summarizes some o f t h e s i g n i f i c a n t design and performance parameters

f o r t h e various LMFBR designs considered.

The procedure followed by each analysis group i n assessing t h e impact o f a l t e r n a t e f u e l cycles was e s s e n t i a l l y t h e same. A reference

design ( f o r t h e P u / ~ ~ * cU y c l e ) was selected and analyzed, and then t h e performance parameters o f a l t e r n a t e f i s s i l e / f e r t i l e combinations were c a l c u l a t e d by r e p l a c i n g t h e r e f e r ence core and blanket m a t e r i a l by t h e appropriate a l t e r n a t i v e m a t e r i a l (s). As i n d i c a t e d by Case 1 i n Table 4.5-1,

a d i f f e r e n t reference design was selected

L;

by each group, emphasizing d i f f e r e n t design c h a r a c t e r i s t i c s . The t h r e e basic designs do share c e r t a i n c h a r a c t e r i s t i c s , however. Each i s a " c l a s s i c a l " LMFBR design c o n s i s t i n g o f two core zones of d i f f e r e n t f i s s i l e enrichments surrounded by blankets ( a x i a l and r a d i a l ) o f f e r t i l e m a t e r i a l . I n assessing t h e performance impact o f various f i s s i l e / f e r t i l e combinations, no attempt was made t o modify o r optimize any o f t h e designs t o account for t h e b e t t e r thermophysical p r o p e r t i e s (e.g., me1t i n g p o i n t , thermal c o n d u c t i v i t y , etc.) o f t h e a l t e r n a t e m a t e r i a l s r e l a t i v e t o t h e reference system. (Note: The question of s e l e c t i o n and subsequent o p t i m i z a t i o n o f p r o l i f e r a t i o n - r e s i s t a n t LMFBR core designs i s c u r r e n t l y being addressed as p a r t o f t h e more d e t a i l e d P r o l i f e r a t i o n - R e s i s t a n t Core Design study being c a r r i e d o u t by DOE and i t s contractor^.)^ I n a l l cases ENDF/B-IV nuclear data5 were u t i l i z e d i n t h e c a l c u l a t i o n s .

The adequacy o f these nuclear data r e l a t i v e t o d e t a i l e d evaluation o f t h e denatured f u e l c y c l e n f a s t systems i s open t o some question. Recent measurements o f t h e capture cross s e c t i o n o f 232Th,6 t h e primary f e r t i l e m a t e r i a l i n t h e denatured f u e l cycle, i n d i c a t e s i g n i f i c a n discrepancies between t h e measured and tabulated ENDF/B-IV cross sections f o r t h e energy range o f i n t e r e s t .

A d d i t i o n a l l y , t h e adequacy of t h e nuclear data f o r t h e primary de-

natured f i s s i l e species, 2 3 3 U ,

f o r the LMFBR s p e c t r a l range has a l s o been questioned.'

Due t o these p o s s i b l e nuclear data u n c e r t a i n t i e s and a l s o t o the l a c k o f design optimizat i o n o f t h e reactors themselves, i t i s prudent t o regard t h e r e s u l t s t a b u l a t e d i n Table 4.5-1 as p r e l i m i n a r y evaluations, subject t o r e v i s i o n as more data become a v a i l a b l e . The compound system f i s s i l e doubling time given i n Table 4.5-1 was c a l c u l a t e d using t h e simple approximation t h a t

C.S.D.T

= 0.693 s ( I n i t i a l Core + Eq. Cycle Charge) Eq. Cycle Charge) (RF x Eq. Cycle Discharge

-

'

I:


4-49

Table 4.5-1.

Fuel U t i l i z a t i o n Characteristics and P e r f o m n c e Parameters for LMFBRs Under Various Oxide-Fuel Options c

Case

Core

Reactor Materials Axial Radial Core Vol. Fractions, 81anket Blanket Fuel/Na/SS/Control

Capacity Factor

Thermal Efficiency

Core Specific Power. BOL (HWth per kg Fisst l e Material)

Breeding Ratio, MOEC

Apparent Canpound Fissile Doubling Ttme (yr)

Equilibrium Cycle Initial Fissile Inventory (kg/GWe)

Ftssile Charge (kg/GWe)

Net F t s s l l e Production (kg/GWe-#r) 'U.PU

Bumu (HD/kg

Y)

Calculation Dim./Gr./Cy.' Parameters,

Contributor Data

Energy-Center-Constrained Fuels 1

17.2 9.6 12.7

3424 3072 2270

1647 1453 804

0.+242 0,+363 0,+187

51

211111

.10

1.27 1.36 1.27

88

17.5 10.4 13.1

3443 3077 2291

1523 1540 804

+122.+110 +150,+197 +154,+30

51

.11

1.27 1.35 1.27

0.36 0.32

1.27 1.34

19.5 10.8

3480 3093

1674 1545

+298,-77 +299 ,+35

0.36 0.32 0.39

1.20 1.19 1.14

40.2 27.9 36.1

4016 3641 2712

1717 1806 920

42/38/20/0 41/44/15/0 43/40/15/2

0.75 0.72 0.75

0.36 0.32 0.39

42/3a/20/0 41/44/15/0 43/40/15/2

0.75 0.72 0.75

0.36 0.32 0.39

3

42/3a/20/0 41/44/15/0

0.75 0.72

4

42/38/20/0 41/44/15/0 43/40/15/2

0.75 0.72 0.75

2

0.94

Oiiii2/9/12

ANL HEN ORNL

2/11/? 2/4/2 2/9/12

ANL HEM ORML

51

2/11/? 2/4/2

ANL HEM.

+89a.-723 +79a,-662

57

+583,-493

95

2/11/? 2/4/2 2/9/12

ANL HEM ORNL

88

I

Dispersible Denatured Fuels 5

41/44/15/0

0.72

0.32

1.20

â&#x201A;Źe+

8r3+

43/40/15/2

0.75

0.39

42/3a/20/0 41/44/1510 43/40/15/2

0.75 0.72 0.75

0.36 0.32 0.39

1.25

8

43/40/15/2'

0.75

0.39

1.16

9

43/40/15/2

0.75

0.39

1.10

6 7

16.1

2937

1483

-698,923

2%

2/4/2

HEM

2Lv2 2/9/12

ORML ANL HEM ORNL

1488

46&+u8

1.13

24.2

2038

795

-354,+453

92

1.16 1.18 1.12

27.5 19.2 26.4

3135 2973 2056

1330 1498 801

-34a.+490 -~.+63a -254,+347

51 92

2/11/? 2/4/2 2/9/12

,

1.09

43.0

2208

a34

-136,+203

95

2/9/12

OWL

'

1.05

118.1

2322

875

-41 ,+7a

98

2/9/12

ORNL

---

3822 3452 2419

1673 1726 911

+31 ,O +59.0 +15,0

57

4-3

1.25

~

Reference Fuels 10

42/38/20/0 41/44/15/0 43/4U/15/2

0.75 0.72 0.75

0.36 0.32 0.39

1' ,

1.06

1.04 1.06 1.02

154.0

99

AWL 2/4/2 2/9/12

HER ORML

!

dim ens ions/Groups/Cycles.

Reference f u e l f o r LMFBR. CReference f u e l s are considered only as l i m i t i n g cases.

.

c


a

4-50 where RF i s the reprocessing recovery factor (0.98). While such an expression i s not absolutely correct, i t does provide a measure o f the r e l a t i v e growth c a p a b i l i t y o f each reactor. Since the data sumnarized i n Table 4.5-1 are based on three separate reference LMFBRs operating with a v a r i e t y of design differences and fuel management schemes, the above expression was used simply t o provide r e l a t i v e values for each system. It should also be noted t h a t sane reactor configurations l i s t e d have d i s s i m i l a r core and a x i a l blanket materials and thus would probably require modifications t o standard reprocessing procedures. The data presented i n Table 4.5-1,

although preliminary, do serve t o indicate cer-

t a i n generic characteristics regarding the impact o f the alternate LMFOR f u e l options. By considering those cases i n which s i m i l a r core materials but different blanket materials are u t i l i z e d i t i s clear t h a t the choice o f the blanket material has only a rather small e f f e c t on the reactor physics parameters. On the other hand, the impact o f changes i n the core f i s s i l e and f e r t i l e materials i s considerable, p a r t i c u l a r l y on the breeding r a t i o . U t i l i z i n g 233U as the f i s s i l e material results i n a s i g n i f i c a n t decrease i n the breeding r a t i o r e l a t i v e t o the corresponding Pu-fueled case (ranging from 0.10 t o 0.15, depending on the system). This decrease i s due p r i m a r i l y t o the lower value o f v (neutrons produced per f i s s i o n ) o f 2 % r e l a t i v e t o 239Pu and 241Pu. Somewhat compensating f o r the difference i n v i s the f a c t t h a t the capture-to-fission r a t i o o f 233U i s s i g n i f i c a n t l y less than t h a t o f the two plutonium isotopes. The differences i n breeding r a t i o s given i n Table 4.5-1 r e f l e c t the net r e s u l t o f these two effects, the decrease i n v c l e a r l y dominating. Use of

I

233U as the f i s s i l e material also results i n a s l i g h t decrease i n the f i s s i l e inventory required f o r c r i t i c a l i t y . This i s due t o two effects, the lower capture-to-fission r a t i o o f 233U r e l a t i v e t o the plutonium isotopes, and the obvious decrease i n the atomic weight o f 233U r e l a t i v e t o Pu (% 2.5%).

. The replacement o f 2 j * U by zs2Th as the core f e r t i l e material also has a s i g n i f i c a n t impact on the overall breeding r a t i o regardless o f the f i s s i l e material u t i l i z e d . As the data i n Table 4.5-1 indicate, there i s a substantial breeding r a t i o penalty associated w i t h the use o f 232Th as a core material i n an LMFBR. This penalty i s due t o the much lower fast f i s s i o n effect i n 232Th r e l a t i v e t o t h a t i n 2s*li (roughly a f a c t o r o f 4 lower). The f e r t i l e f a s t f i s s i o n e f f e c t i s r e f l e c t e d i n the breeding r a t i o i n two ways. F i r s t , although the excess neutrons generated by the f i s s i o n o f a f e r t i l e nucleus can be subsequently captured by f e r t i l e material. t h e i r production i s not a t the expense of a f i s s i l e nucleus. Moreover, the f e r t i l e f i s s i o n e f f e c t produces energy, thereby reducing the f i s s i o n r a t e required o f the f i s s i l e material t o maintain a given power level. Since both these effects act t o improve the breeding ratio, it i s not surprising t h a t use o f Th-based fuels r e s u l t i n s i g n i f i c a n t degradation i n the breeding ratio. A f u r t h e r consequence o f the reduced f a s t f i s s i o n e f f e c t o f 232Th i s a marked increase i n f i s s i l e inventory required for c r i t i c a l i t y , evident from the values given i n Table 4.5-1 f o r the required i n i t i a l loadings.

. 4

> 3 3

3 3

3

a 1 3 3 3 1 3 1 3


4-51

ments.

The c a l c u l a t i o n s f o r LMFBRs operating on denatured 233U f u e l cover a range o f enrichCases 5, 6, and 7 assume an %12%enrichment, Case 8 a 20% enrichment, and Case 9

A l l these reactors are, o f course, subject t o t h e breeding r a t i o penalty inherent i n r e p l a c i n g plutonium w i t h 233U as the f u e l material. The l e s s denatured cases (8 and 9) a l s o r e f l e c t t h e e f f e c t o f thorium i n t h e LMFBR core spectrum. (These higher

a 40% enrichment.

enrichment cases were c a l c u l a t e d i n an attempt t o parameterize t h e e f f e c t o f varying t h e

A f u r t h e r p o i n t which must be addressed regarding t h e denatured amount o f denaturing.) reactors i s t h e i r s e l f - s u f f i c i e n c y i n terms o f t h e f u e l m a t e r i a l 233U. Since t h e denatured LMFBRs t y p i c a l l y c o n t a i n both 232Th and 238U as p o t e n t i a l f i s s i l e materials, both 233U and 239Pu a r e produced v i a neutron capture. Thus i n evaluating t h e s e l f - s u f f i c i e n c y o f a f a s t breeder reactor, t h e 233U component o f t h e o v e r a l l breeding r a t i o i s o f primary importance since t h e bred plutonium cannot be recycled back i n t o t h e denatured system.

As i l l u s t r a t e d

schematically by Fig. 4.5-1, t h e 233U component o f t h e breeding r a t i o increases as t h e allowable denatured enrichment i s increased (which allows t h e amount o f thorium i n t h e f u e l m a t e r i a l t o be increased).

More importantly, t h e magnitude o f t h e 233U component o f t h e \

breeding r a t i o i s very s e n s i t i v e t o t h e allowable degree o f denaturing a t t h e lower enrichThe o v e r a l l breeding r a t i o decreases as t h e allowable ments (i.e., between 12% and 20%). enrichment i s raised, b u t a concomitant and s i g n i f i c a n t decrease i n t h e r e q u i r e d 233U makeup presents a strong i n c e n t i v e from a performance viewpoint t o s e t t h e enrichment as high as I n f a c t , based on t h e data summarized i n i s permitted by n o n p r o l i f e r a t i o n constraints. Table 4.5-1,

t h e lowest enrichment l i m i t f e a s i b l e f o r t h e conventional LMFBR type systems

anaJyzed,lies i n t h e ' l l - 1 4 % (inner-outer core) range. Such a system would u t i l i z e a l l U02 f u e l and would r e q u i r e s i g n i f i c a n t amounts o f 233U as makeup. (It should be noted t h a t t h e 233U/Th system i s n o t denatured.

It i s included i n Fig. 4.5-1

because i t represents

an upper bound on t h e 233U enrichment.) Since a l l denatured reactors r e q u i r e an i n i t i a l inventory o f 233U, as w e l l as varying amounts o f 233U as makeup m a t e r i a l , a second class o f reactors must be considered when e v a l u a t i n g t h e denatured f u e l cycle. The purpose o f these systems would be t o produce t h e 233U r e q u i r e d by t h e denatured reactors.

Possible LMFBR candidates f o r t h i s r o l e are t h e

P u / ~ ~ * rUe a c t o r w i t h thorium blankets (Cases 2 and 3 ) , a Pu/Th r e a c t o r w i t h thorium blankets (Case 41, and a 233U/Th breeder (Case lo).+ I n t h e r e d u c e d - p r o l i f e r a t i o n r i s k scenario, a l l t h r e e o f these systems, since they a r e n o t denatured, would be subject t o rigorous safeguards and operated o n l y i n nuclear 'weapon s t a t e s o r i n i n t e r n a t i o n a l l y c o n t r o l l e d energy centers. Performance parameters f o r these t h r e e types o f systems are included i n Table 4.5-1, and t h e i s o t o p i c f i s s i l e production ( o r d e s t r u c t i o n ) obtained from t h e ORNL calcuClearly, each system has i t s own unique l a t i o n s i s schematically depicted by Fig. 4.5-2. properties.

From t h e standpoint o f 233U production c a p a b i l i t y , t h e h y b r i d Pu/Th system i s

*See discussion on "transmuters" on p.4-10.


4-52

c

'L

12%

Denatured U

20 2 Denatured U

40% Denatured U

%/Th

Fig. 4.5-1. Mid-Equilibrium Cycle Breeding Ratio Isotopics for Denatured Oxide-Fueled LMFBRs. (ORNL Cases 7. 8, 9. and 10 from Table 4,5-11 clearly superior. However, it does require a large quantity of fissile plutonium as makeup since it essentially "transmutes" plutonium into 233U. The system with the thorium radial blanket generates significantly less 233U but also markedly reduces the required plutonium feed. In fact, for the case illustrated, this system actually produces a slight excess of plutonium. The 233U/Th breeder, characterized by a very small excess 233U production, does not provide a means for utilizing the plutonium bred in the denatured systems, and thus it does not appear to have a place in the symbiotic systems utilizing energy-center reactors paired with dispersed reactors. (The coupling of each type of fissile production reactor with a particular denatured system is considered in Section 7.2.) As a final point, preliminary estimates have been made of the safety characteristics of some of the alternate fuel cycle LMFBRs relative to those of the reference cycle. Initial calculations have indicated that the reactivity change due to sodium voiding of a 233U-fueled system is significantly smaller than that of the corresponding Pu-fueled system.8 Thus, the denatured reactors, since they are fueled with 233U, would have better sodium voiding characteristics relative to the reference system. However, for oxide fuels the reported results indicate that the Doppler coefficient for Tho,-based fuels is comparable to that of the corresponding 238U02-based fuels.

c 6

L

u I'

I;


4-53

77-16948

0

Fig. 4.5-2. Equilibrium Cycle Net F i s s i l e Production f o r Possible Oxide-Fueled 233U Production Reactors. (ORNL Cases 10, 2, and 4 from Table 4.5-1) References f o r Section 4.5 1.

D. R. Hoffner, R. W. Hondie, and R. P. Omberg, "Reactor Physics Parameters of

2.

Y. I. Chang, C. E. T i l l , R. R. Rudolph, J.

3.

T. J. Burns and J. R. White, "Preliminary Evaluation of Alternative LMFBR Fuel Cycle Options," ORNL-5389, (1978).

4.

"The Proliferation-Resistant Preconceptual Core Design Study," J . C. Chandler, D. R. Marr, D. C. Curry, M. B. Parker, and R. P. Omberg, Hanford Engineering Development Laboratory; and V. W. Lowery, DOE Division of Reactor Research and Technology (March, 1978).

5.

BNL-17541 (ENDF-201), 2nd Edition, "ENDF/B Summary Documentation," compiled by D. Garber (October 1975).

6.

R. L. Macklin and J. Halperin, ' 1 2 3 2 T h ( ~ , v )Cross Section from 2.6 t o 800 keV," NucZ. SCi. fig. 64, NO. 4, pp. 849-858 (1977).

7.

L. Weston, private communication, March 1971.

8.

B. R. Seghal, J. A. Naser, C. L i n , and W. B. Loewenstein, "Thorium-Based Fuels i n Fast Breeder Reactors," NEZ. Tech. 35, No. 3, p.' 635 (October 1977).

Alternate Fueled LMFBR Core Designs ,I' Hanford EngineerinQ Gevelopment Laboratory, (June 1977).

P. Deen, and M. J . Ring, "Alternative Fuel Cycle Options: Performance Characteristics of and Impact on Nuclear Power Growth Potential ,'I RSS-TM-4, Araonne National Laboratory, (July 1977).


4-54

4.6. 4.6.1.

ALTERNATE FAST REACTORS Advanced Oxide-Fueled LMFBRs

T. J. Burns Oak Ridge National Laboratory One method of improving the breeding performance of the LMFBRs discussed in the previous section is to increase the core fertile loadings. Typically, this goal is accomplished by one of two means: redesign of the pins to accommodate larger pellet diameters or the use of a heterogeneous design (i.e., intermixed core and blanket assemblies). To maintain consistency with the "classical" designs considered in the previous section, using the same fuel elements for both concepts, the latter option was i pursued to assess the impact o f possible redesign options. Table 4.6-1 summarizes some preliminary results from calculations for a heterogeneous reactor core model consisting of alternating concentric fissile and fertile annuli (primed cases) and compares them with results from calculations for corresponding homogeneous cores (unprimed cases). As the data in Table 4.6-1 indicate, the heterogeneous conffquration results

in a

.

significant increase in the overall breeding ratio relative to the corresponding homogeneous calculation. The heterogeneous reactors also require a much greater fissile loading for criticality due to the increase in the core fertile loading. However, the increase in the breeding gain more than compensates for the increased fissile requirements, resulting in an overall improvement in the fissile doubling time. On the other hand, because of the high fissile loading requirements, it appears that a heterogeneous model for the denatured cases with 12% enrichment (cases 6 or 7 of the previous section) is unfeasible; therefore, an enrichment of 1 2 0 % was considered as the minimum for the denatured heterogeneous configuration. While the denatured heterogeneous configurations result in an increase in the overall breeding ratio, it is significant that the 233U component of the breedinq ratio also improves. Figure 4.6-1 depicts the breeding ratio components for both the homogeneous and heterogeneous denatured configurations. (Aaain , the 233U/Th LMFBR is included as the upper limit.) As Fig. 4.6-1 indicates, the heterogeneous confiaurations are clearly superior from the standpoint of 233U self-sufficiency (i.e. , requirinq less makeup requirements). Moreover, if enrichments in the range of 304: 40% are allowed, it appears possible for a denatured heterogeneous reactor to produce enough 233U to satisfy its own equilibrium cycle fuel requirements. Production reactors would therefore be required only to supply the initial inventory plus the additional makeup consumed before the equilibrium cycle is reached.

-


Table 4.6-1. Comparison of Fuel Utilization Characteristics and Performance Parameters for Homogeneous and Heterogeneous LMFBRs Under Various Oxide-Fuel Options Reactor Materials Case'

Driver

1 1'

PU/U PU/U

2 2' 4 4'

PU/U PU/U

Pu/Th Pu/Th

Axial B1 anket U U U U Th Th

Internal Blanket

Radial B1 an ket

Breeding Ratio, MOEC

Fissile Doubling Time (yr) (RF.0.98)

Energy-Center-Constrained Fuels U 1.27 12.7 U 1.50 10.2 Th 1.27 13.1 Th 1.44 12.9 Th 1.14 36.1 Th 1.35 18.2

Equilibrium Cycle

Initial

Fissile Inventory (kg/GWe) 2270 3450 2291 3725 2712 41 59

Fissile Charge (kg/GWe-yr) 804 1173 804 1250 920 1365

Fissile Dlscharge (kg/GWe-yr) 2 3 3u

-

154 536 583 800

PUT

991 1517 834 1013 427 808

VI v1

Dispersible Denatured Fuels 8b 8' 9= 9

I'

10 10'

33U/(U+Th) 233u/u

3U/ (U+Th) 2 3 3u/u

3U/Th %/Th -

Th U Th U

Th Th Th Th

Th Th

Th Th

43.0 18.0 112.3 20.8

2208 3338 2322 4062

834 1624 875 1354

698 1548 835 1457

203 306 78 108

Reference Fuels e 1.02 1.20 30.1

241 9 371 8

91 1 1309

926 1454

0 0

1.09 1.29 1.05 1.29

-

P

~

'%apacitv factor is 75%; unprimed cases are for homoseneous cores,. .primed cases for heterogeneous cores; see Tabie 4.5-1 for case description. b20% *33u/u. =40% 233U/U; 'Included for illustrative purposes only; exceeds design constraints. =Reference fuels are considered only as limiting cases.


i

4-56

78-2612

n

i

8

9

i!'

9'

CASE NUMBER i

Fig. 4.6-1. Breeding R a t i o Com onents f o r LMFBRs Operating on 233U. 20% 233U/U, and Cases 9,9' f o r 40% 23 U/U; Cases 10,lO' f o r 233U/Th w i t h no 4.5-1 and 4.6-1.)

5

8,8' f o r 1Cases 38U; see Tables

The heterogeneous designs a l s o can be employed f o r t h e energy-center production r e a c t o r s r e q u i r e d by t h e denatured f u e l cycles. As i n d i c a t e d i n Table 4.6-1, t h e t h r e e possible production r e a c t o r s a l l show s i g n i f i c a n t increases i n t h e q u a n t i t y o f produced.

233U

The n e t production r a t e s are i l l u s t r a t e d schematically by Fig. 4.6-2.

More

importantly, however, use o f a heterogeneous core design w i l l a l l o w the i s o t o p i c s o f the f i s s i l e m a t e r i a l bred i n the i n t e r n a l blankets t o be adjusted f o r changing demand requirements w i t h o u t modifying the d r i v e r assemblies.

For example t h e i n t e r n a l blankets

o f the Pu/Th LMFBR could be e i t h e r Tho2 o r 238U02, depending on t h e demand requirements f o r 233U and Pu.

c! [J

c L


4-57

78-261 3

_ I

10

10'

2

2'

4

4'

CASE NUMBER

-. Fig. 41.6-2. Net F i s s i l e Production Rates f o r LMFBRs. (Cases 10,lO' f o r 233U/Th core w i t h no 238U, Cases 2,2' f o r P U / ~ ~core, ~ U and Cases 4,4' f o r Pu/Th core; see Tables 4.5-1 and 4.6-1.)


4-5%

4.6.2.

Carbide- and Metal-Fueled LMFBRS D. L. Selby P. M. Haas H. E. Knee Oak Ridge National Laboratory

Another method t h a t i s being considered f o r improving t h e breeding r a t i o s o f LMFBRs and i s c u r r e n t l y under development1 i s one t h a t uses carbide- o r metal-based f u e l s .

The

major advantages o f t h e metal- and carbide-based f u e l s are t h a t they w i l l r e q u i r e lower i n i t i a l f i s s i l e i n v e n t o r i e s than comparable oxide-based f u e l s and w i l l r e s u l t i n s h o r t e r doubling times. This i s e s p e c i a l l y t r u e f o r metal-based fuels, f o r which doubling times as low as 6 years have been calculated.2 Since f o r f a s t reactors t h e denatured f u e l c y c l e would have an i n h e r e n t l y lower breeding gain than t h e reference plutonium-uranium cycle, these advantages would be e s p e c i a l l y important; however, as discussed below, before e i t h e r carbide- o r metal-based f u e l s can be f u l l y evaluated, many a d d i t i o n a l studies are needed. Carbi de-Based Fuels Carbide-based f u e l s have been considered f o r use as advanced f u e l s in conventional Pu/U LMFBRs. Burnup l e v e l s as high as 120,000 MWD/T appear feasible, and t h e f i s s i o n gas release i s l e s s than t h a t for mixed oxide fuels.3 Carbide f u e l s a l s o have a higher thermal conduct i v i t y , which allows higher l i n e a r power r a t e s w i t h a lower c e n t e r - l i n e temperature. I n general, the breeding r a t i o f o r carbide f u e l s i s higher than the breeding r a t i o f o r oxide f u e l s b u t lower than t h a t f o r metal f u e l s .

Both helium and sodium bonds are being considered f o r carbide pins. A t present 247 carbide pins w i t h both types o f bonds are being i r r a d i a t e d i n EBR-11. Other d i f f e r e n c e s i n the p i n s i n c l u d e f u e l density, cladding type, cladding thickness, type o f shroud f o r the sodium-bonded pin, and various power and temperature conditions.

The l e a d p i n s have already

achieved a burnup l e v e l of 10 at.%, and i n t e r i m examinations have revealed no major problems. Thus t h e r e appears t o be no reason why the goal o f 12 at.% burnup cannot be achieved. I n terms o f safety, i r r a d i a t e d carbide f u e l releases g r e a t e r q u a n t i t i e s o f f i s s i o n gas Depending upon the accident scenario, t h i s could be

upon m e l t i n g than does oxide f u e l .

e i t h e r an advantage o r a disadvantage.

Another problem associated w i t h carbide f u e l s may

be the p o t e n t i a l f o r large-scale thermal i n t e r a c t i o n between the f u e l and t h e coolant [see discussion o f p o t e n t i a l FCIs ( e e l - C o o l a n t I n t e r a c t i o n s ) below]. Metal-Based Fuels Reactors w i t h metal-based fuels have been operating i n t h i s country since 1951 (Fermi-I, EBR-I, and EBR-11). R e l a t i v e t o oxide- and carbide-fueled systems, t h e metalfueled systems are characterized by higher breeding r a t i o s , lower doubling times, higher heat conductivity, and lower f i s s i l e mass. These advantages are somewhat o f f s e t , however, by several disadvantages, i n c l u d i n g fuel s w e l l i n g problems t h a t necessitate operation a t lower f u e l temperatures.


4-59

Most o f t h e i n f o r m a t i o n a v a i l a b l e on metal f u e l s i s f o r uranium-fissium (U-Fs) f u e l . (Fissium c o n s i s t s o f e x t r a c t e d f i s s i o n products, p r i n c i p a l l y zirconium, niobium, molybdenum, technetium, ruthenium, rhodium, and palladium.) Some i n f o r m a t i o n i s a v a i l a b l e f o r t h e Pu/U-Zr and U/Th a l l o y fuels b u t none e x i s t s on Pu/Th metal f u e l s . (The U/Th I n terms of i r r a d i a t i o n f u e l s do n o t r e q u i r e t h e a d d i t i o n o f anqther metal f o r s t a b i l i t y . ) experience, approximately 700 U-Fs d r i v e r f u e l elements have achieved burnups of 10 at.% w i t h o u t f a i l u r e . Less i r r a d i a t i o n i n f o r m a t i o n i s a v a i l a b l e f o r t h e Pu/U-Zr a l l o y , w i t h o n l y 16 Pu/U-Zr encapculated elements having been i r r a d i a t e d t o 4.6 a t .% b ~ r n u p . ~Fast r e a c t o r experience w i t h U/Th f u e l s i s a l s o q u i t e l i m i t e d ; however, a recent study a t Argonne National Laboratory has shown t h a t t h e i r r a d i a t i o n performance o f U/Th f u e l s should be a t l e a s t as good as t h a t o f U-Fs fuels.5

With respect t o safety, one concern with metal f u e l s i s t h e p o s s i b i l i t y o f thermal i n t e r a c t i o n s between t h e f u e l and the cladding. For most metal a l l o y s , t h e f u e l w i l l swell t o contact t h e cladding between 3 and 5 at.% burnup. This e f f e c t has been observed i n i r r a d i a t i o n experiments; however, for burnups up t o 10 at.% , no more than 4% of the cladding has been affected. Thus whether o r n o t fuel-cladding i n t e r a c t i o n s w i l l be a l i m i t i n g f a c t o r f o r f u e l burnup remains t o be determined. For t r a n s i e n t gvereower (TOP) analysis, t h e behavior o f U/Th elements hss been shown t o be s u p e r i o r t o t h e behavior o f t h e present EBR-I1 f u e l (uranium w i t h 5% fissium), t h e U/Th elements having a 136OoC f a i l u r e threshold versus 1000째C f o r t h e EBR-I1 elements.

Thus

U/Th metal p i n s would have a higher r e l i a b i l i t y d u r i n g t r a n s i e n t s than t h e f u e l pins already i n use i n f a s t reactors. On t h e o t h e r hand, fuel-coolant i n t e r a c t i o n (FCI) accidents may present a major problem, more so than f o r carbide f u e l s (see below).

P o t e n t i a l f o r Large-Scale FCIs The p o t e n t i a l f o r a large-scale FCI t h a t would be capable o f producing mechanical work s u f f i c i e n t t o breach t h e r e a c t o r vessel and thereby release r a d i o a c t i v i t y from t h e primary containment has been an important s a f e t y concern f o r LMFBRs f o r a number o f years. The assumed scenario f o r a large-scale FCI i s t h a t a l a r g e mass of molten f u e l (a major p o r t i o n of t h e core) present as the r e s u l t o f an h-ypothetical c o r e cJsruptive

accident

(HCDA) contacts and " i n t i m a t e l y mixes w i t h " about t h e same mass o f l i q u i d sodium..

The extremely r a p i d heat t r a n s f e r from t h e molten f u e l ( w i t h temperatures perhaps 3000 t o

t o t h e much c o o l e r sodium (%lOOO째K) produces r a p i d v a p o r i z a t i o n of t h e sodium. If t h e mixing and thermal conditions a r e i d e a l , t h e p o t e n t i a l e x i s t s f o r t h e vaporizat i o n t o be extremely rapid, i.e., f o r a vapor "explosion" t o occur w i t h t h e sodium vapor a c t i v e as t h e working f l u i d t o produce mechanical work.

4000'K)

A g r e a t deal of l a b o r a t o r y experimentation, modeling e f f o r t , and some " i n - p i l e " t e s t i n g has been c a r r i e d o u t i n t h i s country and elsewhere t o d e f i n e t h e mechanisms f o r and t h e necessary-and-sufficient

conditions f o r an energetic FCI o r vapor explosion f o r


4-60 I

-

1

given materials, particularly f o r oxide LMFBR fuel and sodium. Although there i s no conclusive theoretical and/or experimental evidence, the most widely accepted theory i s that f o r an energetic'vapor explosion t o occur, there must be intimate liquid-liquid contact of the fragmented molten fuel particles and the contact temperature a t the fuel-sodium surface must exceed the temperature required f o r homogeneous nucleation of the sodium. A considerable amount of evidence exists to suggest that f o r oxide fuel i n the reactor environment, the potential f o r a large-scale vapor explosion i s extremely remote. The key factor i s the relatively low thermal conductivity of the oxide fuel, which does not permit r a p i d enough heat transfer from the fuel t o cause the fuel-sodium contact temperature t o exceed the sodi um homogeneous nucleation temperature, The primary difference between carbide and/or metal fuels as opposed t o oxide fuels i s t h e i r relatively higher thermal conductivity. Under typical assumed accident conditions, i t i s possible t o calculate coolant temperatures which exceed the sodium homogeneous nucleation temperature. This does not mean, however, that a large-scale FCI will necessarily occur f o r carbide-sodium or metal-sodium systems. As noted above, these theories as mechanisms f o r vapor explosion have not been completely substantiated. However, insofar as the homogeneous nucleation criterion i s adequate, i t i s clear t h a t the potential f o r largescale vapor explosion, a t l e a s t i n clean laboratory systems, i s greater f o r carbide or metal in sodium t h a n f o r oxide in sodium. Continued theoretical and experimental study i s necessary t o gain a thorough understanding of the d e t a i l s of the mechanisms involved and t o estimate the likelihood for vapor explosion under reactor accident conditions for any breeder system. Breeding Performance of Alternate Fuel Schemes Table 4.6-2 shows t h a t in terms of f i s s i l e production, the reference Pu/U core with U blankets gives the best breeding performance regardless of fuel type (oxide, carbide, o r metal). For the carbide systems considered, a heterogeneous core design using Pu/U carbide fuel w i t h a U carbide blanket gives a breeding r a t i o of 1.550. For the metal systems considered, a nominal two-zone homogeneous. core design using U-Pu-Zr alloy fuel gives a breeding r a t i o of 1.614. The increased f i s s i l e production capability of the carbide and metal fuels i s especially advantageous f o r the denatured cycles. A breeding r a t i o as high as 1.4 has been calculated f o r a metal denatured system, and the breeding r a t i o f o r a carbide denatured system i s not expected t o be substantially smaller. However, a good part of the f i s s i l e production of any denatured system i s plutonium. Thus the denatured system i s not a good producer of 233U. However, when used with the energy park concept, where the plutonium produced by the denatured systems can be used as a fuel, the denatured carbide and metal uranium systems are viable concepts. Metal and carbide concepts may also prove t o be valuable as transmuter systems f o r producing 233U from 232Th.

t: L

I:

L'

I:


4-61

Table 4.6-2.

i b

Breeding R a t i o

L --

Blanket

Oxide Fuels

Carbide Fuels

P u / 2 W (reference)

23811

1 .44b

1. 550b

*33U/238U/p~-Zr

23811

1.614

233~/238U/pu-Zr

Th

1.537

238u

1.532 1.406

Fuela

k

L

Beginning-of-Life Breeding Ratios f o r Various LMFBR Fuel Concepts

J

33U/238U/P~/Th

Metal Fuels 1.62gC

I ,

23 3U/2 38U/Pu/Th

Th

b

Pu/Th 233U/Th

Th Th

1.3ob 1.041

1.353b 1.044

1.38lC 1.1O!jc

23%/Th

Th

0.786

0.817

0. 906c

233U/238U-Zr (denatured)

Th

--

I

1.41b

' A l l Pu i s LWR discharge Pu. 'Radial heterogeneous design. cFrom r e f . 2. O f t h e thorium metal systems considered, t h e U/Pu/Th t e r n a r y metal system was found t o

t o be t h e best 233U producer.

I r r a d i a t i o n experiments have shown t h a t t h e U/Pu/Th a l l o y can

be i r r a d i a t e d a t temperatures up t o 700?C w i t h burnups o f up t o 5.6 at.%.5

.I

Li

Beginning-of-

c y c l e breeding r a t i o s around 1.4 have been c a l c u l a t e d f o r t h i s system, and i t appears t h a t o p t i m i z a t i o n o f core and blanket geometry may increase t h e breeding r a t i o t o as h i g h as 1.5. It i s a l s o c l e a r t h a t t h e e q u i l i b r i u m c y c l e breeding r a t i o may be as much as 10% higher due t o t h e f l u x increase i n t h e blankets from t h e 233U production. This system n o t o n l y i s a pure 233U producer (no plutonium i s produced), b u t a l s o acts as a plutonium s i n k by burning plutonium produced i n l i g h t - w a t e r reactors. Summary and Conclusions Both carbide- and metal-based f u e l s have l a r g e r breeding gains and p o t e n t i a l l y lower doubling times than t h e oxide-based fuels. When t h e p r o l i f e r a t i o n issue i s considered i n t h e design aspect ( e s p e c i a l l y f o r 233U/Th concepts w i t h t h e i r i n h e r e n t l y lower breeding gains), these advantages a r e enhanced even more. I n l i g h t o f t h e emphasis on p r o l i f e r a t i o n r e s i s t a n t nuclear design, t h e carbide- and metal-fueled reactors have t h e p o t e n t i a l t o

L

c o n t r i b u t e e x t e n s i v e l y t o t h e energy requirements o f t h i s country i n t h e future. However, t h e f i r s t step i s t o e s t a b l i s h carbide and metal f u e l data bases s i m i l a r t o t h e present data base f o r oxide fuels, p a r t i c u l a r l y f o r s a f e t y analyses. Present development plans f o r carbide and metal f u e l s c a l l f o r a lead concept s e l e c t i o n f o r t h e carbide f u e l s by $1981, w i t h t h e metal f u e l s e l e c t i o n coming i n $1984.

-'i

i '

b


4-62

4.6.3

Gas-Cooled Fast Breeder Reactors

t'

T. J . Burns

Oak Ridge National Laboratory

I n a d d i t i o n t o t h e sodium-cooled f a s t r e a c t o r s discussed above, t h e impact o f t h e various a1t e r n a t e f i s s i l e / f e r t i l e f u e l combinations on t h e Gas-Cooled Fast Breeder Reactor (GCFR) has a l s o been addressed (although n o t t o t h e degree t h a t i t has f o r t h e LMFBR). GCFR design w i t h f o u r enrichment zones was selected as t h e reference case.7-8

A 1200-MWe Pu/U The various

bl

a l t e r n a t i v e f i s s i l e / f e r t i l e f u e l combinations were then s u b s t i t u t e d f o r t h e reference f u e l .

No design m o d i f i c a t i o n s o r o p t i m i z a t i o n s based on t h e a l t e r n a t e f u e l p r o p e r t i e s were perI t should a l s o be emphasized t h a t t h e r e s u l t s o f t h i s scoping e v a l u a t i o n f o r

formed.

a l t e r n a t e - f u e l e d GCFRs are n o t comparable t o t h e r e s u l t s given i n Section 4.5 f o r LMFBRs due t o markedly d i f f e r e n t design assumptions f o r t h e reference cases.

L;

The r e s u l t s o f t h e p r e l i m i n a r y c a l c u l a t i o n s f o r t h e a l t e r n a t e - f u e l e d GCFRs, sumrelative marized i n Table 4.6.3, r e f l e c t trends s i m i l a r t o those shown by LMFBRs; i.e., t o t h e reference case, a s i g n i f i c a n t breeding r a t i o penalty occurs when 233U i s used as t h e f i s s i l e m a t e r i a l and 232Th as t h e core f e r t i l e m a t e r i a l .

>loreover, t h e magnitude o f t h e penalty (ABR) i s l a r g e r f o r t h e GCFR than f o r t h e LMFBR. Owing t o t h e helium coolant, t h e c h a r a c t e r i s t i c spectrum o f t h e GCFR i s s i g n i f i c a n t l y harder than t h a t of a comparably sized LMFBR. I n l i g h t o f t h e r e l a t i v e nuclear p r o p e r t i e s o f t h e various f i s s i l e and f e r t i l e species discussed i n Section 4.5, t h i s increased p e n a l t y due t o t h e harder spectrum i s n o t s u r p r i s i n g .

The number o f neutrons produced per f i s s i o n ( v ) o f

t h e f i s s i l e Pu isotopes i n t h e GCFR i s s i g n i f i c a n t l y higher than t h e number produced I n t h e s o f t e r spectrum o f an LMFBR. The value o f v f o r 233U, on t h e o t h e r hand, i s r e l a t i v e l y i n s e n s i t i v e t o s p e c t r a l changes.

Hence, t h e l a r g e r penalty associated w i t h

233U-based f u e l s i n t h e GCFR i s due t o t h e b e t t e r performance o f t h e Pu reference system r a t h e r than t o any marked changes i n 233U performance. A s i m i l a r argument can be made f o r the replacement o f core f e r t i l e m a t e r i a l .

Owing t o t h e harder spectrum, t h e f e r t i l e

f a s t - f i s s i o n e f f e c t i s more pronounced i n t h e GCFR than i n an LMFBR. Thus, t h e r e d u c t i o n i n t h e f e r t i l e f i s s i o n cross s e c t i o n r e s u l t i n g from replacement o f 238U b y 232Th r e s u l t s i n a l a r g e r decrease i n t h e breeding r a t i o .

I t should a l s o be noted t h a t as i n t h e LMFBR

case, 233U-fueled GCFRs r e q u i r e smaller f i s s i l e i n v e n t o r i e s than do t h e corresponding

I;

Pu-fueled cases. The b e t t e r breeding performance o f Pu i n t h e harder spectrum o f t h e GCFR, on t h e o t h e r hand, i n d i c a t e s t h a t t h e GCFR would be a v i a b l e candidate f o r t h e r o l e o f energy center "transmuter," e i t h e r as a Pu/Th system o r as a Pu/U + Tho, r a d i a l b l a n k e t system. It must be emphasized, however, t h a t these conclusions are t e n t a t i v e as they a r e based

c


4-63

on o n l y t h e p r e l i m i n a r y data presented i n Table 4.6-3.

The p o s s i b i l i t y o f employing

heterogeneous designs and/or carbide- o r metal-based f u e l s has n o t been addressed.

It

should a l s o be noted t h a t evaluation o f which type o f r e a c t o r i s best s u i t e d f o r a given r o l e i n the denatured f u e l c y c l e must a l s o r e f l e c t nonneutronic considerations such as c a p i t a l cost, p o s s i b l e i n t r o d u c t i o n date, etc. Table 4.6-3.

Fuel U t i l i z a t i o n Characteristics and Performance Parameters f o r GCFRs Under Various Fuel Optionsa (2% losses assumed i n reprocessing)

Reactor Materials Core

Axial Blanket

Radial Blanket

Initial Fissile Inventory (kg/GWe)

Breeding Ratio, MOEC

Fissile Doubling Time ( y r ) (RF=0.98)

Equi 1ibrium Cycle Fissile F i s s i l e Discharge Charge (kg/GWe-yr) (kg/GWe-yr) 713' U Pu'

Energy-Center-Constrained Fuels PU/U

U U

Pu/Th

Th

PU/U

U Th

2641 2693

1.301 1.276

14.3 15.4

965 987

224

1163 941

Th

3170

1.150

48.3

1158

626

619

Dispersible Denatured Fuels Z33UIUb

U

Th

2538

1.088

50.5

1001

671

400

233u/uc

Th Th Th

Th Th Th

2587 2720 2956

1.074 1.060 1.004

66.8 98.4

1019 1031 1131

822 a71 1054

256 208 81

1192

1169

233U/U + Thd

233U/U + The

Reference Fuels 233U/Th

Th

Th

3108

0.970

bCapacit f a c t o r i s 75%. 117.9% 2r3U/U d17.7% 233u/u: 20% 233u/u. 9 0 % 233U/U. Reference f u e l s are considered only as l i m i t i n g cases.

References f o r Section 4.6

1.

J. M. Simnons, J. A. Leary, J. H. K i t t e l and C. M. Cox, "The U.S. Advanced LMFBR Fuels Development Program," Advanced W B R Fuels, pp. 2-14, ERDA 4455 (1977).

2.

Y. I . Chang, R. R. Rudolph and C. E. T i l l , " A l t e r n a t e Fuel Cycle Options (Performance C h a r a c t e r i s t i c s and Impact on Power Growth P o t e n t i a l ) ,I' June 1977.

3.

A. Strasser and C. Wheelock, 'Uranium-Plutonium Carbide Fuels f o r Fast Reactors ,'I Fast Reactor Technology National Topical Meeting, D e t r o i t , Michigan ( A p r i l 26-28, 1965).

4.

W. F. Murphy, W. N. Beck, F. L. Brown, 6. J. Koprowshi, and L. A. Meimark, "Posti r r a d i a t i o n Examination o f U-Pu-Zr Fuel Elements I r r a d i a t e d i n EBR-I1 t o 4.5 Atomlc Percent Burnup," ANL-7602, Argonne National Laboratory (November 1969).

5:

B. R. Seidel, R. E. Einziger, and C. M. Walter, "Th-U M e t a l l i c Fuel: LMFBR P o t e n t i a l Based Upon EBR-I1 Driver-Fuel Performance," !m2?28. h. NUCZ. SOC. 27, 282 (1977).

6.

B. Blumenthal, J. E. Sanecki, D. E. Busch, and D. R. O'Boyle, "Thorium-UraniumPlutonium A l l o y s as P o t e n t i a l Fast Power-Reactor Fuels, P a r t 11. Properties and I r r a d i a t i o n Behavior o f Thorium-Uranium-Plutonium A1 l o y s ,'I ANL-7259, Argonne National Laboratory (October 1969).

7.

L e t t e r t o D. E. B a r t i n e from R. J. Cerbone, A p r i l 22, 1976, 760422032 GCFR.

8.

L e t t e r t o D. E. B a r t i n e from MWe GCFR Data.

R. J. Cerbone, A p r i l 22, 1976, 760422032, Subject: 1200-


c


--

CHAPTER 5

b

IMPLEMENTATION OF DENATURED FUEL CYCLES

I

i '

b Chapter Out1 i n e

L

5.0.

I n t r o d u c t i o n , T. J. Burns, ORNL

5.1.

Reactor Research and Development Requirements, 5.1.1. 5.1.2. 5.1.3. 5.1.4. 5.1.5. 5.1.6.

5.2.

N . L . S h a p i r o , CE

Light-Water Reactors High-Temperature Gas-Cooled Reactors Heavy-Water Reactors S p e c t r a l - S h i f t - C o n t r o l l e d Reactors R,D&D Schedules Summary and Conclusions

Fuel Recycle Research and Development Requirements, 5.2.1. 5.2.2. 5.2.3.

r.

S p i e w a k , ORNL

Technology Status Summary Research, Development, and Demonstration Cost Ranges and Schedules Conclusions


c

e E

c


5-3

W 5.0.

INTRODUCTION

T. J. Burns Oak Ridge National Laboratory

--

I

L 1

&

Currently, a major p o r t i o n o f t h e nuclear generating capacity i n the U.S. o f LWRs operating on t h e LEU once-through cycle.

consists

Implementation o f t h e denatured 233U f u e l

c y c l e w i l l r e q u i r e t h a t t h e nuclear f u e l c y c l e be closed; thus research and development e f f o r t s d i r e c t e d a t nuclear f u e l c y c l e a c t i v i t i e s , t h a t is, reprocessing, f a b r i c a t i o n o f f u e l assemblies containing r e c y c l e m a t e r i a l , etc.,

w i l l be necessary, as w e l l as research

and development o f s p e c i f i c r e a c t o r systems designed t o u t i l i z e these a l t e r n a t e f u e l s .

To

date, most f u e l c y c l e R&D has been d i r e c t e d a t c l o s i n g t h e Pu/U f u e l c y c l e under t h e

c

assumption t h a t plutonium would e v e n t u a l l y be recycled i n the e x i s t i n g LWRs.

With t h e

exception o f t h e HTGR ( f o r which a 330-MWe prototype r e a c t o r i s undergoing t e s t i n g a t F o r t S t . Vrain), and t h e L i g h t Water Breeder Reactor (LWBR) a t Shippingport, Pa.,

U.S.

reactors

have n o t been designed t o operate on thorium-based fuels, and thus the R&D f o r thoriumbased f u e l cycles has n o t received as much a t t e n t i o n as t h e R&D f o r t h e Pu/U cycle. r e s u l t , any s t r a t e g y f o r implementation o

I, b

As a

t h e denatured f u e l c y c l e on a t i m e l y basis must

be concerned w i t h f u e l c y c l e research and development.

I t must a l s o be concerned w i t h

r e a c t o r - s p e c i f i c research and development since t h e implementation o f t h e denatured 233U c y c l e i n any r e a c t o r w i l l necessitate des gn changes i n t h e reactor. The f o l l o w i n g two sections o f t h i s chapter contain estimates o f t h e research and development costs and p o s s i b l e schedules f o r the r e a c t o r - r e l a t e d research and development and t h e f u e l -cycle-related research and development r e q u i r e d f o r implementation o f t h e denatured f u e l c y c l e i n t h e various types o f reactors t h a t have been considered i n e a r l i e r chapters o f t h i s r e p o r t . connected:

It should be noted t h a t these two sections are i n t r i n s i c a l l y

t h e implementation o f a r e a c t o r operating on r e c y c l e f u e l necessitates t h e

p r i o r implementation o f t h e reprocessing and f a b r i c a t i o n f a c i l i t i e s attendant t o t h a t f u e l , and conversely, t h e decision t o c o n s t r u c t a reprocessing f a c i l i t y f o r a s p e c i f i c r e c y c l e f u e l type i s d i c t a t e d by t h e existence ( o r p r o j e c t e d existence) o f a r e a c t o r discharging the fuel.

c


5-4

5.1.

REACTOR RESEARCH AND DEVELOPMENT REQUIREMENTS N. L. Shapiro Combustion Engineering Power Systems

The discussions i n the preceding chapters, and a l s o the discussion t h a t f o l l o w s i n Chapter 6, a l l assume t h a t LWRs and advanced converters based on t h e HTGR, HWR, and SSCR concepts w i l l be a v a i l a b l e f o r commercial operation on denatured uranium-thorium (DUTH) f u e l s on a r e l a t i v e l y near-term time scale. I f t h i s commercialization schedule i s t o be achieved, s u b s t a n t i a l r e a c t o r - r e l a t e d research and development w i l l be required.

The purpose o f t h i s

s e c t i o n i s t o delineate t o the degree p o s s i b l e a t t h i s p r e l i m i n a r y stage o f development the magnitude and scope o f the r e a c t o r R,D&D requirements necessary f o r implementation o f the reactors on DUTH f u e l s and, f u r t h e r , t o determine whether there are s i g n i f i c a n t R,D&D c o s t d i f f e r e n c e s between t h e r e a c t o r systems.

The requirements l i s t e d are those b e l i e v e d t o be

necessary t o resolve the technical issues t h a t c u r r e n t l y preclude the deployment o f the various r e a c t o r concepts on DUTH f u e l s , and no attempt i s made t o prejudge o r t o i n d i c a t e a p r e f e r r e d system. It i s t o be emphasized t h a t t h e proper development o f r e a c t o r R,D&D costs and schedules

would r e q u i r e a comprehensive i d e n t i f i c a t i o n o f design and l i c e n s i n g problems, t h e development o f d e t a i l e d programs t o address these problems, and the subsequent development o f costs and schedules based upon these programs. Unfortunately, the assessment o f a1 t e r n a t e converter concepts has n o t as y e t progressed t o t h e p o i n t t h a t problem areas can be f u l l y i d e n t i f i e d , and so d e t a i l e d development o f R,D&D programs i s g e n e r a l l y i m p r a c t i c a l a t t h i s stage.

Con-

sequently, we have had t o r e l y on somewhat s u b j e c t i v e evaluations o f the technological s t a t u s o f each concept, and upon r a t h e r approximate and somewhat i n t u i t i v e estimates o f the costs r e q u i r e d t o resolve the s t i l l undefined problem areas.

A more d e t a i l e d development o f t h e

requirements f o r many o f the candidate systems w i l l be performed as p a r t o f t h e characterizat i o n and assessment programs c u r r e n t l y under way i n . the N o n p r o l i f e r a t i o n A1 t e r n a t i v e Systems Assessment Program (NASAP) . I n general, r e a c t o r R,D&D requirements can be d i v i d e d i n t o two major categories: (1) the R,D&D p e r t a i n i n g t o the development o f t h e r e a c t o r concept on i t s reference fuel

cycle; and (2) the R,D&D necessary f o r the deployment o f t h e r e a c t o r operating on an a l t e r n !

a t e f u e l c y c l e such as a DUTH f u e l cycle.

I n the discussion presented here i t i s assumed

t h a t , w i t h t h e exception o f the HTGR (whose reference f u e l c y c l e already includes thorium), the reference cycles o f the advanced converters would i n i t i a l l y be the uranium c y c l e (i.e., 235U/238U)

and t h a t no r e a c t o r would employ DUTH f u e l u n t i l a f t e r i t s s a t i s f a c t o r y per-

formance had been assured i n a l a r g e - p l a n t demonstration.

Although i t i s p o s s i b l e t o

consider t h e development o f advanced converters using DUTH f u e l as t h e i r reference f u e l cycle, such simultaneous development could be a p o t e n t i a l impediment t o c o m e r c i a l i z a t i o n since surveys o f the u t i l i t y and manufacturing sectorsâ&#x20AC;&#x2122; i n d i c a t e a near u n i v e r s a l reluctance t o e h a r k on e i t h e r a new r e a c t o r technology o r a new f u e l c y c l e technology, l a r g e l y because


5-5

o f t h e u n c e r t a i n t i e s w i t h respect t o r e a c t o r o r fuel c y c l e performance, economics, l i c e n s a b i l i t y , and t h e s t a b i l i t y o f government p o l i c i e s .

Thus attempts t o introduce a new re-

a c t o r technology c o n d i t i o n a l upon the successful development o f an u n t r i e d f u e l c y c l e technology would o n l y compound these concerns and complicate t h e already d i f f i c u l t problem o f comnercialization.

The development o f advanced converter concepts intended i n i t i a l l y f o r

uranium f u e l i n g would a1 low research and development, design, and t h e eventual demonstrai

t i o n o f t h e concept t o proceed simultaneously w i t h the separate development o f t h e DUTH

Lf

cycle. The R,D&D r e l a t e d t o the r e a c t o r concept i t s e l f t y p i c a l l y can be d i v i d e d i n t o three components:

-

(1)

Proof o f p r i n c i p l e (operating t e s t r e a c t o r o f small s i z e ) .

ij

(2)

Design, construction, and operation o f prototype p l a n t (intermediate s i z e ) .

(3)

Design, construction, and operation of comnercial-size demonstration p l a n t (about 1000 We).

i

I,

Each stage t y p i c a l l y involves some degree of b a s i c research, component design and t e s t i n g , and l i c e n s i n g development. I n c e r t a i n instances, various stages o f the development can be bypassed. This i s p a r t i c u l a r l y t r u e o f technologies representing o n l y a modest departure

L

from the present r e a c t o r technology, i n which case prototype r e a c t o r c o n s t r u c t i o n may be bypassed completely and demonstrations performed on comnercial-size u n i t s . I f a decision i s made t o do t h i s , the t i m e r e q u i r e d t o introduce comnercial-size u n i t s can be shortened, b u t f i n a n c i a l r i s k s are increased because o f t h e l a r g e r c a p i t a l commitment r e q u i r e d f o r f u l l scale u n i t s .

On the o t h e r hand, t o t a l R&D costs are somewhat reduced, since some f r a c t i o n

of t h e R&D r e q u i r e d f o r prototype design u s u a l l y proves n o t t o be a p p l i c a b l e t o l a r g e - p l a n t

--

design.

f '

t

It i s a l s o p o s s i b l e i n c e r t a i n instances t o perform component R&D and design f o r t h e

prototypes i n such a fashion t h a t i d e n t i c a l components can be used d i r e c t l y i n the demonThus, by employing components of the same design and s i z e i n both systems

s t r a t i o n units.

the R&D necessary t o scale up components c o u l d be avoided. Each o f t h e t h r e e advanced converter reactors discussed i n t h i s s e c t i o n has already proceeded through t h e proof-of-principle

O f these, t h e HTGR i s t h e most h i g h l y develop-

stage.

ed w i t h i n t h e United States, w i t h a 330-MWe prototype c u r r e n t l y operating ( t h e F o r t S t . Vrain p l a n t ) . HWRs have received much l e s s development w i t h i n t h e United States, b u t reactors o f t h i s type have been commercialized i n t h e Canadian CANDU reactor. h w e v e r , due t o d i f f e r e n c e s i n design between the CANDU and the HWR p o s t u l a t e d f o r U.S. pected use o f s l i g h t l y enriched f u e l i n a U.S.

L

s i t i n g i i o r example, the ex-

HWR) and a l s o t o differences i n l i c e n s i n g

c r i t e r i a , i t would s t i l l be desirable t o c o n s t r u c t a U.S. t o t h e commercial-size demonstration p l a n t phase.

prototype p l a n t before proceeding

The SSCR represents o n l y a modest

departure from t h e design o f PWRs already operating, b u t even so, the c o n s t r u c t i o n and operation o f a prototype p l a n t would a l s o be t h e l o g i c a l next stage i n the e v o l u t i o n o f t h i s concept.


5-6

As has been pointed o u t above, r e l a t i v e l y r a p i d i n t r o d u c t i o n schedules f o r t h e various reactors have been postulated i n t h e nuclear power scenarios described i n Chapter 6.

This i s because one o f t h e o b j e c t i v e s o f t h i s r e p o r t i s t o e s t a b l i s h t h e degree t o

which advanced converters and t h e denatured uranium-thorium (DUTH) c y c l e can c o n t r i b u t e t o improved uranium resource u t i 1 i z a t i o n so as t o defer the need f o r plutonium-fueled breeder r e a c t o r s and t o e l i m i n a t e from f u r t h e r consideration those concepts which cannot c o n t r i b u t e s i g n i f i c a n t l y t o t h i s goal even i f r a p i d l y introduced. The SSCR i s assumed t o be i n t r o duced i n 1991 and HWRs and HTGRs i n 1995.

I n view o f the time requirements f o r p l a n t

c o n s t r u c t i o n and l i c e n s i n g , i t i s c l e a r t h a t t h e prototype p l a n t stage w i l l have t o be bypassed i f these i n t r o d u c t i o n dates are t o be achieved.

Consequently, f o r t h e discussion

below i t has been assumed t h a t t h e program f o r each r e a c t o r w i l l be d i r e c t e d toward t h e c o n s t r u c t i o n o f t h e demonstration p l a n t . t u r n d i v i d e d i n t o two p a r t s :

provide t h e basic i n f o r m a t i o n necessary f o r t h e design and l i c e n s i n g o f a commercial-site demonstration f a c i l i t y ; and another c o n s i s t i n g o f t h e f i n a l design, construction, and operation o f the f a c i l i t y . For t h i s demonstration program, continued government funding has been assumed because o f the s u b s t a n t i a l R&D and f i r s t - o f - a - k i n d engineering costs t h a t w i l l be i n c u r r e d and because o f t h e increased r i s k s associated w i t h bypassing t h e prototype stage.

I n considering f u e l - c y c l e - r e l a t e d r e a c t o r R,D&D,

(2)

Reactor components development.

(3)

Reactor/fuel c y c l e demonstration.

L L L

R,D&D re-

q u i r e d t o s h i f t t o an a l t e r n a t e c y c l e ( s p e c i f i c a l l y a DUTH c y c l e ) need be addressed.*

Data-base development.

1'

i t i s assumed t h a t t h e demonstration

o f t h e r e a c t o r concept on i t s reference c y c l e has been accomplished and o n l y t h a t

(1)

L L

This r e a c t o r / f u e l c y c l e demonstration i s i n

one c o n s i s t i n g o f the generic r e a c t o r R&D r e q u i r e d t o

b a s i c ,types o f f u e l - c y c l e - r e l a t e d r e a c t o r

I;

The

R,D&D are:

The purpose o f t h e data base development R&D i s t o provide physics v e r i f i c a t i o n and f u e l performance i n f o r m a t i o n necessary f o r t h e design and l i c e n s i n g o f r e a c t o r s operating on t h e subject f u e l cycle; t h e i n t e n t here i s t o provide i n f o r m a t i o n s i m i l a r t o t h a t which has been developed f o r t h e use o f mixed-oxide f u e l s i n LWRs.

Physics v e r i f i c a t i o n experiments

have t y p i c a l l y consisted o f c r i t i c a l experiments t o provide a basis t o demonstrate t h e a b i l i t y o f a n a l y t i c a l models t o p r e d i c t such important s a f e t y - r e l a t e d parameters as r e a c t i v i t y l e v e l , c o e f f i c i e n t s o f r e a c t i v i t y , and poison worths.

Safety-related f u e l performance R&D might

L L

c o n s i s t o f such aspects as f u e l r o d i r r a d i a t i o n s t o e s t a b l i s h i n - r e a c t o r performance and discharge i s o t o p i c s ; special r e a c t o r experiments t o e s t a b l i s h such parameters as i n - r e a c t o r swelling, d e n s i f i c a t i o n , c e n t e r - l i n e temperature and f i s s i o n gas release; and t e s t s of t h e *Note t h a t the R,D&D requirements included are those r e l a t e d t o the design, l i c e n s i n g and operation of the reactor onty. The requirements f o r developing t h e f u e l c y c l e i t s e l f are considered separately (see Section 5.2). The prime example o f such f u e l - c y c l e - r e l a t e d r e a c t o r R,D&D i s t h a t already performed f o r plutonium recycle. Here, f a i r l y extensive R,D&D was performed both by the government and by the p r i v a t e s e c t o r t o develop r e a c t o r design changes and/or r e a c t o r - r e l a t e d constraints, l i c e n s i n g information, and i n - r e a c t o r gemonstrations t o support the eventual u t i l i z a t i o n o f mixed-oxide f u e l s .

I'

i;


5-7

-

U

L [I.

performance o f the f u e l during a n t i c i p a t e d operational t r a n s i e n t s . Since such s a f e t y - r e l a t e d fuel performance information would be developed as p a r t o f the f u e l r e c y c l e program d i s cussed i n Section 5.2,

t h e R&D costs f o r t h i s aspect are mentioned here o n l y f o r completeness.

Reactor components development has been included since, i n p r i n c i p l e , t h e use o f a l t e r n a t e f u e l s might change t h e bases f o r r e a c t o r design s u f f i c i e n t l y t h a t a d d i t i o n a l components development could be required.

ibd

The e x t e n t o f t h e r e a c t o r design m o d i f i c a t i o n s r e -

q u i r e d t o accomnodate a change from a r e a c t o r ' s reference f u e l t o denatured f u e l would, o f course, vary w i t h t h e r e a c t o r type. The t h i r d aspect o f f u e l - c y c l e - r e l a t e d R&D i s t h e r e a c t o r / f u e l c y c l e d e m n s t r a t i o n . This demonstration includes t h e core physics design and s a f e t y analysis, which i d e n t i f i e s any changes i n design basis events o r i n r e a c t o r design necessitated by the denatured uranium-thorium f u e l cycles, t h e preparation of an analysis r e p o r t (SAR) , and t h e subsequent i n - r e a c t o r demonstration o f s u b s t a n t i a l q u a n t i t i e s o f denatured f u e l s . I n summary, a number o f assumptions have been made t o a r r i v e a t a p o i n t o f r e f e r ence f o r evaluating t h e research and development r e q u i r e d f o r reactors t o be commercialized

b

L

on a DUTH f u e l c y c l e w i t h i n the postulated schedule.

I n p a r t i c u l a r , i t has been assumed

t h a t t h e prototype p l a n t stage e i t h e r has been completed o r can be bypassed f o r HTGRs, HWRs, and SSCRs, and thus t h e remaining R,D&D r e l a t e d t o the r e a c t o r concept i t s e l f i s t h a t r e q u i r e d t o operate a connnercial-size demonstration p l a n t .

The demonstration p l a n t s

are based on each r e a c t o r ' s reference f u e l r a t h e r than on a DUTH f u e l ; t o convert t h e r e a c t o r s t o a DUTH f u e l w i l l r e q u i r e a d d i t i o n a l R,D&D t h a t w i l l be fuel-cycle-related. For t h e LWRs, which have long passed t h e demonstration stage on t h e i r reference f u e l , a l l t h e r e a c t o r R,D&D r e q u i r e d t o operate t h e reactors on a DUTH f u e l i s fuel-cycle-related. The demonstration program i n t h i s case would be t h e demonstration o f DUTH f u e l i n a

1

u

current-generation LWR. (Note: This discussion does n o t consider r e a c t o r R,D&D t o s u b s t a n t i a l l y improve t h e resource u t i l i z a t i o n o f LWRs, which, as i s pointed o u t i n Section 4.1 and Chapters 6 and 7, i s c u r r e n t l y being studied as'one approach f o r increasi n g t h e power production from a f i x e d resource base.) /

T h i s e v a l u a t i o n has a l s o r e q u i r e d t h a t assumptions be made regarding t h e degree o f f i n a n c i a l support t h a t could be expected from t h e government.

These assumptions, and t h e

c r i t e r i a on which they are based, are presented i n the discussions below on each r e a c t o r type. While t h e assumptions regarding government p a r t i c i p a t i o n are unavoidably a r b i t r a r y and may be s u b j e c t t o debate, i t i s t o be p o i n t e d o u t t h a t b a s i c a l l y t h e same assumptions

I; L

have been made f o r a l l r e a c t o r types. Thus the reader may scale the costs presented t o correspond t o o t h e r sets o f assumptions. F i n a l l y , i t i s t o be noted t h a t w h i l e t h e nuclear power systems included i n t h i s study o f t h e denatured 233U fuel c y c l e i n c l u d e f a s t breeder reactors, no estimates are i n c l u d e d i n t h i s s e c t i o n f o r FBRs.

..&

Estimated research and development c o s t schedules for


5-8

t h e LMFBR on i t s reference c y c l e are c u r r e n t l y being revised, and a study o f L . e denatured f a s t breeder f u e l cycle, which includes f a s t transmuters and denatured breeders, i s included as p a r t o f t h e INFCE program ( I n t e r n a t i o n a l Nuclear Fuel Cycle Evaluation).

The r e s u l t s

from the INFCE study should be a v a i l a b l e i n t h e near f u t u r e . 5.1.1.

Light-Water Reactors

Preliminary evaluations o f design and s a f e t y - r e l a t e d considerations f o r LWRs operati n g on t h e conventional thorium c y c l e i n d i c a t e thorium-based f u e l s can be employed i n LWRs w i t h l i t t l e o r no m o d i f i c a t i o n . Consequently, the R&D costs given here have been estimated under the assumption t h a t denatured f u e l w i l l be employed i n LWRs o f e s s e n t i a l l y present design.

This assumption i s n o t meant t o exclude minor changes t o r e a c t o r design ( f o r

example, changes i n t h e number of c o n t r o l drives, shim loadings, o r f u e l management, etc.) b u t r a t h e r r e f l e c t s our c u r r e n t b e l i e f t h a t design changes necessitated by DUTH f u e l s w i l l be s u f f i c i e n t l y s t r a i g h t f o r w a r d so as t o be accommodated w i t h i n t h e engineering design t y p i c a l l y performed f o r new plants. As has been described i n the discussion above, the f i r s t phase o f such fuel-cycle-

r e l a t e d research consists o f the development o f a data base from which s a f e t y - r e l a t e d parameters and f u e l performance can be p r e d i c t e d i n subsequent core physics design and s a f e t y analysis programs. F i r s t , e x i s t i n g thorium m a t e r i a l s and f u e l performance i n f o r mation should be thoroughly reviewed, and a p r e l i m i n a r y e v a l u a t i o n o f s a f e t y and l i c e n s i n g issues should be made i n order t o i d e n t i f y missing i n f o r m a t i o n and guide the subsequent development program.

Although t h i s i n i t i a l phase i s r e q u i r e d t o f u l l y d e f i n e t h e r e q u i r e d

data base R&D, i t i s p o s s i b l e t o a n t i c i p a t e i n advance the need t o e s t a b l i s h i n f o r m a t i o n i n the areas o f physics v e r i f i c a t i o n and s a f e t y - r e l a t e d f u e l performance. As shown i n Table 5.1-1,

ment i s estimated t o c o s t

the physics v e r i f i c a t i o n program under data base develop-

%$lo m i l l i o n .

This program should be designed both t o provide

the i n f o r m a t i o n r e q u i r e d t o p r e d i c t important s a f e t y - r e l a t e d physics parameters and t o demonstrate t h e accuracy o f such p r e d i c t i o n s as p a r t o f the s a f e t y analysis.

Improved

values must be obtained f o r cross sections o f thorium and o f isotopes i n t h e thorium d e p l e t i o n chains, such as 233U and protactinium, a l l o f which have been l a r g e l y neglected i n t h e past. Resonance i n t e g r a l measurements should a l s o be performed f o r denatured f u e l s both a t room temperature and a t elevated temperatures, such experiments being very i m p o r t a n t f o r accurately c a l c u l a t i n g s a f e t y - r e l a t e d physics c h a r a c t e r i s t i c s and a l s o f o r e s t a b l i s h i n g the quanti t i e s o f plutonium produced during i r r a d i a t i o n . F i n a l l y , an LWR physics v e r i f i c a t i o n program should i n c l u d e a s e r i e s o f c r i t i c a l experiments, p r e f e r a b l y both a t room temperature and a t elevated moderator temperatures, f o r each of t h e f u e l types under consideration (i .e. , f o r thorium-based f u e l s u t i l i z i n g denatured 23sU, denatured 233U, o r plutonium).

These experiments would serve as a basis f o r demonstrating t h e adequacy

o f t h e cross-section data sets and o f t h e a b i l i t y o f a n a l y t i c a l models t o p r e d i c t such s a f e t y - r e l a t e d parameters as r e a c t i v i t y , power d i s t r i b u t i o n s , moderator temperature r e a c t i v i t y c o e f f i c i e n t s , boron worth, and c o n t r o l r o d worth.


5-9

Table 5.1-1. Government Research and Development Required t o Convert Light-Water Reactors t o Denatured Uranium-Thorium Fuel Cycles (20% 235U/238U-Th o r 20% 233U/238U-Th) Assumptions:

A l l basic r e a c t o r R&D r e q u i r e d f o r commercialization o f LWRs operating on t h e i r reference f u e l c y c l e (LEU) has been completed. Use o f denatured f u e l can be demonstrated i n a current-generation LWR. Because u t i l i t y sponsoring demonstration w i l l be t a k i n g some r i s k o f decreased r e a c t o r a v i l a b i l i t y , a 25% government subsidy i s assumed f o r a 3-year demonstration program.

Note:

LWRs can be operated on t h e denatured 235U/238U-Th fuel c y c l e before any o t h e r r e a c t o r system; however, they cannot be economically competitive w i t h LWRs operating on t h e LEU once-through c y c l e because higher U308 requirements a r e associated w i t h thorium f u e l . Any commercial LWRs operating on a denatured c y c l e before the y e a r 2000 must be subsidized. cost ($MI

Research and Development

A.

Data base development Al.

Physics v e r i f i c a t i o n program Improve cross sections f o r Th, 233U, Pa, etc. Measure resonance i n t e g r a l s f o r denatured uraniumthorium fuels a t room temperature and a t elevated temperatures.

10

I

Perform and analyze c r i t i c a l experiments f o r each f u e l . A2.

Fuel-performance program

(30

- 150)=

Perform i n - r e a c t o r p r o p e r t i e s experiments Perform power ramp experiments Perform f u e l - r o d i r r a d i a t i o n experiments Perform transient t e s t s B.

Reactor components development (develop hand1 i n g equipment/procedures f o r r a d i o a c t i v e 232Lcont a i n i n g f r e s h f u e l elements).

C.

Demonstration design and l i c e n s i n g

C1.

Develop core design changes as r e q u i r e d f o r denatured fuels

C2.

Perform safety analysis o f m o d i f i e d core

C3.

Prepare s a f e t y analysis r e p o r t (SAR); through l i c e n s i n g

5

- 25

20

- 100

5ob

- 200-

carry

D. Demonstration o f LWR operating on denatured f u e l ( p r obab 1y 5U/ 8U-T h ) aWould be included i n f u e l r e c y c l e R&D costs (see Section 5.2). % J o t e n t i a l government subsidy; i.e.,

t o t a l c o s t o f demonstration i s $200M.


5-10

The fuel performance program under LWR data-base development would consist of the establ ishment of safety-related fuel performance information such a s transient fuel damage limits, thermal performance both f o r normal operation and w i t h respect t o LOCA* margins on stored heat, dimensional s t a b i l i t y (densification and swelling), gas absorption and release behavior, and fuel cladding interaction. The i n i t i a l phase of this program should consist of in-reactor properties experiments, power ramp t e s t s , transient fuel damage t e s t s , and fuel rod irradiations. The in-reactor properties experiments would be similar t o the program currently underway i n Norway's Halden HWR and would be designed t o provide information on such parameters as center-line temperature, swelling and densification, and fissiongas release d u r i n g operation. The power ramp experiments would consist of preirradiation of the fuel rod segments in existing LWRs and the subsequent power ramping of these segments i n special t e s t reactors t o establish anticipated fuel performance during power changes typically encountered i n the operation of LWRs. Examples of such programs are the international inter-ramp and over-ramp programs currently being undertaken a t Studsvi k . The transient fuel damage experiments would be designed t o provide information on the performance of the denatured fuels under the more rapid transients possible d u r i n g operation and i n postulated accidents. Lastly, the fuel rod irradiation experiments would provide information on the irradiation performance of prototypical thorium-based fuel rods, and, w i t h subsequent post-irradiation isotopic analyses, would also provide information on burnup and plutonium production. (As noted previously, the fuel performance program costs are included, though not specifically delineated, under the fuel cycle R,D&D discussed in Section 5.2.) In addition t o the data base development, some as yet unidentified reactor components development could be expected. To cover t h i s aspect of the program, an estimated cost of $5 - $25 million i s included in Table 5.1-1. The remaining fuel-cycle-related R&D f o r LWRs would be devoted t o developing core design changes and safety analysis information i n preparation f o r a reactor/fuel cycle demonstration. In t h i s phase of the program, safety-related behavior of alternate fuel would be determined using the specific design attributes of the demonstration reactor. The effects of alternate fuel cycles on plant safety and licensing would require examination o f safety c r i t e r i a and the dynamic analyses of design basis events. Appropriate safety c r i t e r i a , such as acceptable fuel design limits and limits on maximum energy deposition i n the fuel, would have t o be determined. Changes i n core physics parameters that result from alternate fuel loadings and the implication of these changes on reactor design and safety would also have t o be identified and accommodated w i t h i n the design. For example, changes in fuel and moderator temperature reactivity coefficients, boron worth, control-rod worth, prompt-neutron lifetime and delayed-neutron fraction must be addressed since they can have a large impact on the performance and safety of the system. The effects of alternate fuel cycles on the dynamic system responses should be determined f o r a l l transients required by Regulatory Guide 1.70, Revision 2. I t would also be necessary t o determine the implications of denatured fuel cycles on plant operation and load change performance t o determine whether the response of plant control and protection systems i s *LOCA = Loss-of-Coolant Accident.


5-1 1

W

t,

a l t e r e d . A s a f e t y a n a l y s i s r e p o r t f o r denatured thorium f u e l s would be prepared as p a r t o f t h i s development task and pursued w i t h l i c e n s i n g a u t h o r i t i e s through approval. The r e a c t o r development c o s t associated w i t h comnercializing t h e LWR on the DUTH f u e l c y c l e i s thought t o be about $200 m i l l i o n .

This r e l a t i v e l y low c o s t r e s u l t s from t h e commercial s t a t u s o f t h e LWR and from the r e l a t i v e l y small r i s k associated w i t h deploying a

L L

new f u e l type, since ifthe demonstration program i s unsuccessful, the r e a c t o r can always be returned t o uranium fueling.

The estimated c o s t f o r t h e l i g h t - w a t e r r e a c t o r i s based

on an assumed 25% government subsidy f o r a three-year i n - r e a c t o r demonstration. The 25% subsidy i s intended p r i m a r i l y t o ensure t h e sponsoring u t i l i t y against the p o t e n t i a l f o r decreased r e a c t o r avai l a b i 1it y which might r e s u l t from u n s a t i s f a c t o r y performance o f the DUTH f u e l . (The c o s t of the f u e l i t s e l f i s included i n t h e f u e l r e c y c l e development costs discussed i n Section 5.2.) 5.1.2.

b'

Hi gh-Temperature Gas-Cool ed Reactors

A1 though a number of a1t e r n a t e high-temperature gas-cooled r e a c t o r technologies have been o r are being developed by various countries, t h i s discussion considers the r e a c t o r con-

U. s. experience w i t h high-temperature gascooled r e a c t o r s dates from March 3, 1966, when the 40-MWe Peach Bottom Atomic Power S t a t i o n cept developed by t h e General Atomic Company.

-

t

became operable.

More recently, t h e 330-MWe F o r t S t . Vrain HTGR p l a n t has been completed

and i s c u r r e n t l y undergoing i n 1 t i a l rise-to-power t e s t i n g . the

Consequently, HTGR s t a t u s i n

U. S. i s considered t o be a t the prototype stage and the basic r e a c t o r development

s t i l l r e q u i r e d i s t h a t associated w i t h t h e demonstration o f a l a r g e p l a n t design.

Al-

though the success of the F o r t S t . Vrain prototype cannot be f u l l y assessed u n t i l a f t e r several years of operation, i n t h i s discussion s a t i s f a c t o r y performance o f t h e F o r t S t .

L la 4

i

h

Vrain p l a n t has been assumed. Cost estimates f o r the R&D requirements f o r the development o f a l a r g e commercial These estimates i n c l u d e o n l y HTGR on i t s reference HEU/Th c y c l e are shown i n Table 5.1-2. t h a t R&D r e q u i r e d r e l a t i v e t o t h e F o r t S t . Vrain p l a n t . As these t a b l e s i n d i c a t e , t h e m a j o r i t y o f t h e R&D expenditures would be d i r e c t e d toward component R&D and component design, s p e c i f i c a l l y f o r t h e development o f t h e PCRV (prestressed concrete r e a c t o r vessel), steam generator, instrumentation and c o n t r o l , m a t e r i a l s and methods, and t h e main helium c i r c u l a t o r s and s e r v i c e systems.

I n addition, an estimated $30 m i l l i o n t o $60 m i l l i o n

would be r e q u i r e d f o r l i c e n s i n g and preparing a s a f e t y analysis r e p o r t f o r t h e i n i t i a l power r e a c t o r demonstration program. The c o s t o f a power r e a c t o r demonstration p l a n t f o r the HTGR on i t s reference c y c l e would be s i g n i f i c a n t l y higher than t h e Cost given e a r l i e r f o r an LWR on a DUTH cycle,

L

r e f l e c t i n g the increased c o s t and r i s k associated w i t h deploying new concepts. I n developing the p o t e n t i a l r e a c t o r demonstration costs f o r the HTGR, we have assumed t h a t s u b s t a n t i a l government subsidy (50%) would be required f o r t h e f i r s t u n i t . Since i t w i l l be necessary t o commit a t l e a s t the second through f i f t h o f a k i n d p r i o r t o t h e a

successful operation o f t h i s i n i t i a l demonstration u n i t i f the postulated deployment


. 5.1.2.

Government Research and Development Required t o Demonstrate HTGRS, HWRs, and SSCRs on Their Reference Cycles Assumptions 1.

A l l reactors except LWRs s t i l l require basic reactor research and development f o r operation on t h e i r reference fuel cycles.

2.

Logical progression o f basic reactor RLD (excluding fuel performance and recycle RLD) i s : A. E. C.

3.

Proof o f principle with small t e s t reactor. Design, construction, and operation o f prototype reactor andlor component testing f a c i l i t y . Design. construction, and operation o f demonstration plant.

Substantial government subsidies are required f o r rapid comnercialization o f reactors since unfavorable near-term economics and/or high-risk factors make early comnitmnt on concepts by private sector unattractive.

................................................................. H i gh-Temperature Gas-Cooled Reactors (Reference Fuel Cycle: HEU/Th) a

Research and DeVelODment

Research and Development

x -_

A.

Proof o f principle accomplished i n Peach Bottom Reactor

--

A.

Proof o f principle accomplished by Canada

E.

Prototype reactor operation i n progress (Ft. S t . Vraln plant)

--

E.

Prototypes o f natural-uranium fueled reactors already operated a t <lo00 We by Canada

C.

Large plant design and licensing

C.

Large plant design and licensing

C1.

D.

Component R I D PCRV; steafflgenerators; control and instrumentation ; materials; main helium c i r culators and service systems

Spectral -Shi ft-Control led Reactors b (Reference Fuel Cycle: LEU)

Heavy-Water Reactors'b*c (Reference Fuel Cycle: SEU)

80-90

C2.

Component design

50- 100

C3.

Licensing and SAR development 30-60

Large plant demonstration

D.

A.

Proof o f principle accomplished i n BR3 reactor i n Belgium

--

8.

Prototype operation not believed t o be necessary

--

C.

Large plant design and licensing

120

C1.

Technology transfer and manufacturing license fee

C2.

Component RAD Core modifications; development and modification f o r U.S. s i t i n g Licensing and SAR development 30-100

C3.

Research and Development

C1.

Component R I D Develop D20 upgrader technology; perform theml-hydraulic tests; valve, seal, and pump development t o minimize leakage; develop refueling techniques

30-f@

C2.

Licensing and SAR development

20-50

60-150

Large plant demonstration

0.

50% subsidy o f f i r s t u n i t

400

50% subsidy o f f i r s t u n i t

400

25% subsidy o f next four units

700

25% subsidy o f next four units

700

Large plant demonstration ( i n modified PUR) 100%subsidy o f extra equipment (plus other costs) f o r f i r s t u n i t 100%subsidy o f extra equipmnt f o r next four units

'Estimates based on those from Arthur D. L i t t l e . Inc. study. "Gas Cooled Reactor Assessment." August, 1976. plus subsequent experience a t the Ft. St. Vrain plant. bDemonstration plant may require reactivation o f U.S. heavy-water f a c i l i t i e s ; comnercialization o f these reactors w i l l necessitate development o- f D.0- - oroduction industrv. 'Assumed t o be CANDU-PHWR-based design deployed under Canadian license; RID costs would be significantly higher for U.S.-originated design. Under t h i s assumption, a U.S. prototype i s not thought necessary. although i t may s t i l l be desirable. The use o f SEUIhigher burnups can be demonstrated 4n .. fnnadlnn - o l a-n t s . while other desian mdifications such as higher operating pressures can be demonstrated i n the lead plant o f the large plant demonstration program a f t e r completih o f component RID.

__.

140 100


5-1 3

schedule i s t o be maintained, our costs presume f u r t h e r governmental support w i l l be necessary (a 25% subsidy i s assumed) f o r the second through f i f t h u n i t s .

As noted i n Table 5.1-2, a 50% subsidy o f the f i r s t u n i t i s expected t o be about $400 m i l l i o n , and a 25% subsidy o f t h e next f o u r u n i t s i s expected t o t o t a l $700 m i l l i o n . Since t h e assumptions

underlying government subsidies o f the r e a c t o r demonstration program shown i n Table 5.1-2 have been defined, these costs can be adjusted t o r e f l e c t e i t h e r d i f f e r e n t l e v e l s o f government support o r a change i n t h e o v e r a l l c o s t o f t h e demonstration program. \

As has been s t a t e d above, i t has been assumed t h a t t h e advanced converters such as t h e HTGR would a l l be successfully demonstrated on t h e i r reference cycles before they are converted t o DUTH cycles. However, since t h e reference c y c l e f o r the HTGR i s already a thorium-based cycle, i t i s l i k e l y t h a t a denatured c y c l e could be designated as t h e reference c y c l e f o r t h i s r e a c t o r and thus t h a t the lead p l a n t demonstration program would be f o r a DUTH-fueled HTGR.

I f t h i s were done, t h e a d d i t i o n a l costs r e q u i r e d t o convert

t h e HTGR t o a denatured f u e l might be smaller than those associated w i t h converting LWRs from t h e i r uranium-based f u e l c y c l e t o a thorium-based cycle. 5.1.3.

Heavy-Water Reactors

A1 though a number o f a1t e r n a t e heavy-water r e a c t o r concepts have been developed by various nations, o n l y t h e CANDU pressurized heavy-water r e a c t o r has been deployed i n s i g n i f i c a n t numbers.

Therefore, as noted previously, t h e CANDU r e a c t o r i s taken as the

reference r e a c t o r f o r deployment i n t h e United States.

The R&D c o s t can vary considerably,

depending on whether developed Canadian technology i s u t i l i z e d or whether t h e U.S. e l e c t s t o independently develop a heavy-water-reactor concept. It i s assumed here t h a t t h e U.S. HWR w i l l be based on t h e CANDU-PHWR and deployed under Canadian l i c e n s e and w i t h Canadian

cooperation.

Thus, our costs address o n l y those aspects r e q u i r e d t o extend t h e present

CANDU design t o t h a t o f a l a r g e p l a n t (1,000-MWe) f o r U.S. s i t i n g . An order o f magnitude higher R&D comnitment would be r e q u i r e d i f i t were necessary t o reproduce t h e development and demonstrations which t h e Canadians have performed t o date. Research and development requirements f o r the HWR are included i n Table 5.1-2.

In-

herent i n these requirements i s the assumption t h a t although t h e U.S. design would be based on the CANDU-PHWR, s i g n i f i c a n t changes would have t o be made i n order t o r e a l i z e a commercial o f f e r i n g i n t h e U.S.

These m o d i f i c a t i o n s c o n s i s t o f the development o f a l a r g e

p l a n t design (l,OOO-MWe), t h e use o f s l i g h t l y enriched f u e l both t o improve resource u t i l i z a t i o n and t o reduce power costs, m o d i f i c a t i o n s o f t h e HWR design t o reduce c a p i t a l

The r a t h e r l a r g e range o f p o t e n t i a l R&D costs shown i n Table 5.1-2,

particularly

f o r l i c e n s i n g and SAR development, i s i n d i c a t i v e o f the u n c e r t a i n t y introduced by l i c e n s i n g , i.e.,

t o the degree t o which t h e HWR w i l l be forced t o conform t o l i c e n s i n g

c r i t e r i a developed f o r the LWR.


5-1 4

The f i r s t aspect o f l a r g e p l a n t design and l i c e n s i n g R&D, i d e n t i f i e d as component R&D, i s r e l a t e d p r i m a r i l y t o the extension o f the CANDU t o 1,000 MWe, t h e use o f s l i g h t l y enriched f u e l , and possible increases i n system pressure so as t o reduce e f f e c t i v e c a p i t a l

cost. I n general, increasing t h e power output o f the HWR t o 1,000 MWe should be more readil y accomplished than w i t h other concepts such as the LWR, since i t can be accomplished simply by adding a d d i t i o n a l f u e l channels and an a d d i t i o n a l coolant loop. The use o f s l i g h t l y enriched f u e l and higher operating pressures should r e s u l t i n no fundamental changes t o CANDU design, b u t nevertheless w i l l necessitate some development i n order t o

1;

accommodate the higher interchannel peaking expected w i t h s l i g h t l y enriched f u e l s and the e f f e c t o f higher system pressures on pressure-tube design and performance. f o r U.S.

Modifications

s i t i n g a r e somewhat d i f f i c u l t t o q u a n t i f y since a thorough l i c e n s i n g review o f

the HWR has y e t t o be completed.

Although there i s no doubt o f t h e fundamental s a f e t y o f

the CANDU, m o d i f i c a t i o n s f o r U.S. s i t i n g and l i c e n s i n g are nevertheless a n t i c i p a t e d f o r such reasons a t d i f f e r i n g seismic c r i t e r i a (due t o the d i f f e r i n g geology between the U.S. and Canada) and because o f d i f f e r i n g l i c e n s i n g t r a d i t i o n s .

A d d i t i o n a l experimental informa-

t i o n on the performance o f s l i g h t l y enriched uranium f u e l should a l s o be developed by i r r a d i a t i n g such f u e l i n e x i s t i n g HWRs (such as i n Canada's NPD p l a n t near Chalk River) t o the discharge burnups a n t i c i p a t e d f o r t h e reference design (about 21,000 MWe/TeM). Methods o f analyzing the response o f the HWR t o a n t i c i p a t e d operational occurrences and o t h e r p o s t u l a t e d accidents w i l l have t o be developed and approved by the Nuclear Regulatory Commission, and a s a f e t y analysis r e p o r t i n conformance w i t h NRC c r i t e r i a w i l l have t o be devel oped and defended.

I:

li

c c

As i s t h e case f o r t h e HTGR, t h e c o s t f o r a power demonstration p l a n t f o r t h e HWR would be s i g n i f i c a n t l y higher than the c o s t f o r a DUTH-fueled LWR. The l a r g e p l a n t demons t r a t i o n costs shown i n Table 5.1-2 have been estimated under the same s e t o f assumptions used f o r e s t i m a t i n g the HJGR p l a n t . The c o s t o f a program t o convert an HWR from i t s reference uranium c y c l e t o denatured

a

f u e l would be approximately equal t o t h a t p r e v i o u s l y described f o r the LWR. 5.1.4.

S p e c t r a l - S h i f t - C o n t r o l l e d Reactors

As was noted i n Chapter 4, the SSCR consists b a s i c a l l y o f a PWR whose r e a c t i v i t y c o n t r o l system u t i l i z e s heavy water i n s t e a d o f s o l u b l e boron t o compensate f o r r e a c t i v i t y changes during t h e operating cycle.

c

Since t h e SSCR proof-of- p r i n c i p l e has already been

demonstrated by the operation o f t h e BR3 r e a c t o r i n Belgium, and since various components r e q u i r e d f o r heavy-water handling and reconcentration are w e l l e s t a b l i s h e d by heavy-water r e a c t o r operating experience, the SSCR i s considered t o be a t a stage where e i t h e r a prototype o r a l a r g e power p l a n t demonstration i s required.

,

For most a l t e r n a t i v e r e a c t o r concepts a t t h i s stage o f development, a p r o t o t y p e program would be necessary because o f the c a p i t a l c o s t and high r i s k associated w i t h

I '

II


5-1 5

bypassing t h e prototype stage and c o n s t r u c t i n g a l a r g e power r e a c t o r demonstration.

Such

a prototype program may a l s o be d e s i r a b l e f o r the SSCR, p a r t i c u l a r l y i f t h e prototype program i n v o l v e d t h e m o d i f i c a t i o n o f an e x i s t i n g PWR f o r s p e c t r a l - s h i f t c o n t r o l r a t h e r than the c o n s t r u c t i o n of a w h o l l y new p l a n t f o r t h i s purpose.

However, the estimates o f t h e

r e a c t o r R&D requirements given f o r the SSCR i n Table 5.1-2 are based on t h e assumption t h a t t h i s p r o t o t y p e stage i s bypassed.

This can be j u s t i f i e d on t h e basis t h a t the SSCR i s

r a t h e r unique among the various a l t e r n a t i v e s because o f i t s close r e l a t i o n s h i p t o present

PWR technology. I n p a r t i c u l a r , no r e a c t o r development would be r e q u i r e d and t h e r e a c t o r could be designed so t h a t the p l a n t would be operated i n e i t h e r t h e conventional poison c o n t r o l mode o r i n t h e s p e c t r a l - s h i f t c o n t r o l mode. As a r e s u l t , a g r e a t m a j o r i t y o f the c a p i t a l investment i n t h e p l a n t and the power output of the p l a n t i t s e l f i s n o t a t r i s k . Likewise, t h e p o t e n t i a l f o r serious l i c e n s i n g delays i s l a r g e l y mitigated, since the reac-

i LI'

L c i-

L

t o r c o u l d i n i t i a l l y be operated as a poison-controlled PWR and e a s i l y reconfigured f o r the s p e c t r a l - s h i f t c o n t r o l once the l i c e n s i n g approvals were obtained. Consequently, t h e c a p i t a l a t r i s k i s l i m i t e d t o the a d d i t i o n a l expenditures r e q u i r e d t o r e a l i z e s p e c t r a l -

-

s h i f t c o n t r o l , roughly $30 $60 m i l l i o n f o r component R&D, p l u s r e n t a l charges on the heavy water inventory. The a d d i t i o n a l expenditures f o r design and l i c e n s i n g , $20 $50

-

m i l l i o n , would have a l s o been necessary f o r the prototype. The component R&D would c o n s i s t o f a thermal-hydraulic development task; valves and seal development; development o f design and t e s t i n g .

D20 upgrader technology; and r e f u e l i n g methods development,

The thermal-hydraulic t e s t s would be designed t o produce a departure

from nucleate b o i l i n g c o r r e l a t i o n f o r the SSCR moderator s i m i l a r t o t h a t which has been developed f o r t h e PWR l i g h t - w a t e r moderator.

The c o r r e l a t i o n s are expected t o be very

s i m i l a r , b u t t e s t s t o demonstrate t h i s assumption f o r t h e various mixtures o f heavy and l i g h t water w i l l be required. Valves and seal development w i l l be necessary i n order t o minimize leakage o f t h e heavy-water mixture; reduction o f coolant leakage i s important both from an economic standpoint (because o f t h e c o s t o f D20) and because of t h e p o t e n t i a l r a d i o l o g i c a l hazard from t r i t i u m which i s produced i n the coolant. Methods o f reducing coolant leakage from valves and seals have been e x t e n s i v e l y explored as p a r t o f t h e design e f f o r t on heavy-

I

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i

b

L

water r e a c t o r s and u t i l i z a t i o n o f heavy-water r e a c t o r experience i s assumed. The R&D program would address t h e a p p l i c a t i o n o f t h e technologies developed f o r the heavy-water r e a c t o r t o t h e l a r g e r s i z e components and higher pressures encountered i n t h e SSCR. The D20 upgrader employed i n t h e SSCR i s i d e n t i c a l i n concept t o t h e upgraders used on heavy-water r e a c t o r s and i n t h e l a s t stage ( f i n i s h i n g stage) o f D20 production f a c i l i t i e s . The s i z i n g o f various components i n the upgrader would, however, be somewhat d i f f e r e n t f o r

SSCR a p p l i c a t i o n because o f the range o f D20 concentration feeds ( r e s u l t i n g from t h e changing D20 concentration d u r i n g a r e a c t o r operating cycle), and because o f t h e l a r g e volume o f low D20 concentration coolant which must be upgraded toward the end o f each operating cycle.

The upgrader R&D program would consider t h e s i z i n g o f the upgrader,


5-1 6

'I

I, and should a l s o address methods of minimizing t h e D20 i n v e n t o r y i n the upgrader so as t o minimize D20 i n v e n t o r y charges. Lastly, component R&D should address methods f o r r e f u e l i n g and f o r coolant exchange during r e f u e l i n g .

Refueling should be performed w i t h pure l i g h t water present i n t h e reac-

t o r (so as t o avoid the r a d i o l o g i c a l hazard o f t r i t i u m ) ; t h e l i g h t water must subsequently be replaced w i t h the light-water/heavy-water m i x t u r e p r i o r t o i n i t i a t i n g the n e x t operating cycle. I n order t o accomplish t h i s r e f u e l i n g / c o o l a n t exchange w i t h o u t n e c e s s i t a t i n g l a r g e volumes o f heavy water f o r t h i s purpose, a m o d i f i e d bleed-and-feed procedure i s being exp l o r e d i n which t h e d i f f e r e n c e s i n d e n s i t y between t h e warm water i n t h e core and the cool makeup water i s e x p l o i t e d i n order t o minimize coolant mixing and the amount o f excess D20 i n v e n t o r i e s required.

Scale t e s t s o f t h i s r e f u e l i n g procedure ( o r any o t h e r r e f u e l i n g /

coolant exchange procedure selected) w i l l be required. The R&D r e l a t e d t o s a f e t y and l i c e n s i n g should c o n s i s t f i r s t o f data development f o r t h e SSCR operating on the uranium f u e l cycle. This data base has been p a r t i a l l y developed i n t h e i n i t i a l SSCR development work performed by the USAEC i n t h e 1960s. However, a d d i t i o n a l work, p r i m a r i l y i n t h e area o f physics v e r i f i c a t i o n o f s a f e t y - r e l a t e d parameters (i .e., c r i t i c a l experiments which e s t a b l i s h r e a c t i v i t y p r e d i c t i o n s , power d i s t r i b u t i o n s , D20 worths, and cont r o l r o d worths) are r e q u i r e d f o r uranium f u e l .

The second aspect o f t h e s a f e t y and l i c e n s -

i n g R&D should c o n s i s t o f a p r e l i m i n a r y system design, the performance o f a s a f e t y analysis f o r t h e SSCR, and the development o f a s a f e t y a n a l y s i s r e p o r t f o r s p e c t r a l - s h i f t - c o n t r o l operation.

A t t h i s stage, component design and development would be l i m i t e d t o those areas

i n which some design changes would be r e q u i r e d i n Order t o sure t h a t t h e consequences of p o s t u l a t e d accidents and a n t i c i p a t e d operational occurrences w i t h t h e SSCR would be comparable t o those f o r the conventional PWR. The main areas thought t o r e q u i r e a t t e n t i o n are the i m p l i c a t i o n s o f c o e f f i c i e n t s o f r e a c t i v i t y on accidents t h a t r e s u l t i n a cool-down o f the primary coolant, t h e D20 d i l u t i o n accident, and t r i t i u m production.

The i m p l i c a t i o n s o f the s p e c t r a l - s h i f t mode o f c o n t r o l

on p l a n t operation and load change performance should a l s o be addressed as p a r t o f the p r e l i m i n a r y design evaluation. With respect t o t h e l a r g e p l a n t demonstration of the SSCR, the f i n a n c i a l r i s k t o u t i l i t i e s would be l i m i t e d t o the e x t r a c a p i t a l equipment r e q u i r e d t o r e a l i z e s p e c t r a l - s h i f t control.

Because the proposed schedule f o r commercialization i s more r a p i d f o r t h e SSCR

than f o r any o f t h e other advanced converters, i t has been assumed here t h a t t h e government would e s s e n t i a l l y purchase t h e e x t r a equipment r e q u i r e d f o r the f i r s t f i v e u n i t s ( a t $25 m i l l i o n per u n i t ) .

I n t h e case o f the f i r s t u n i t , a d d i t i o n a l funding t o m i t i g a t e t h e lower

capacity factors a n t i c i p a t e d f o r an experimental u n i t have been added. the f i r s t u n i t includes t h e c a r r y i n g charges on t h e D20 inventory.

Also t h e c o s t f o r

D20 c a r r y i n g charges

are n o t included f o r t h e second through f i f t h u n i t s since i t should be p o s s i b l e t o demonstrate the s p e c t r a l - s h i f t c o n t r o l on the f i r s t u n i t before t h e D20 f o r t h e remaining u n i t s needs t o be purchased, so t h a t a decision t o employ s p e c t r a l - s h i f t c o n t r o l i n subsequent u n i t s would be one which i s p u r e l y commercial i n nature.

C'

IJ


5-1 7

LJ

It i s u n l i k e l y t h a t an SSCR would be converted t o t h e denatured f u e l c y c l e unless a

s i m i l a r change had p r e v i o u s l y occurred i n the LWR. I n t h i s case, o n l y a demonstration o f the performance o f denatured f u e l i n the s p e c t r a l - s h i f t mode of c o n t r o l would be needed. These incremental costs a r e estimated t o be $10

L

5.1.5.

- $60 m i l l i o n .

R,D&D Schedules

Schedules f o r completing the R,D&D e f f o r t delineated above are summarized i n Fig. Although i t can be argued that, given strong governmental support both i n funding 5.1-1. and i n h e l p i n g usher the various concepts through the l i c e n s i n g process, these schedules

L Y

could be accelerated, the schedules shown are thought t o be on the o p t i m i s t i c s i d e o f what can reasonably be expected t o be achieved. I n p a r t i c u l a r , a nine-year p e r i o d has been assumed f o r the design, l i c e n s i n g and c o n s t r u c t i o n o f a new r e a c t o r type; t h i s would appear somewhat o p t i m i s t i c since i t i s c u r r e n t l y t a k i n g longer t o b r i n g conventional LWRs on l i n e . I t should a l s o be noted t h a t i n general t h e time scale required t o develop a l t e r n a t e f u e l

I' L

c y c l e technologies ( c f . Section 5.2) i s estimated t o be a t l e a s t as long, and sometimes longer, than t h a t r e q u i r e d t o develop r e a c t o r - r e l a t e d aspects.

I n general, t h i s i s because

t e s t f a c i l i t i e s ( f o r example, t o perform demonstration i r r a d i a t i o n ) are a v a i l a b l e e i t h e r i n the U.S.

o r i n Canada, so t h a t R&D work p r i o r t o the design, l i c e n s i n g , and c o n s t r u c t i o n

o f a l a r g e demonstration p l a n t could be r a p i d l y i n i t i a t e d . 5.1.6.

i

Sumnary and Conclusions

I t has been the purpose o f t h i s s e c t i o n t o d e l i n e a t e t h e magnitude and scope o f reac-

I,

t o r R,D&D expenditures associated w i t h the use o f DUTH f u e l i n converter reactors and t o determine i f t h e r e are s i g n i f i c a n t R,D&D c o s t d i f f e r e n c e s between r e a c t o r systems.

Recommendations f o r the f u r t h e r development o f s p e c i f i c denatured r e a c t o r s are provided i n Section 7.5 where t h e R&D requirements

discussed here are weighed a g a i n s t the p o t e n t i a l

b e n e f i t s o f various nuclear power systems u t i l i z i n g denatured fuels, Chapter 6.

as presented i n

I n developing t h e nuclear power scenarios examined i n Chapter 6, i t was recognized t h a t t h e b e n e f i t s o f operating LWRs and a l t e r n a t e r e a c t o r types on DUTH f u e l s are dependent upon the speed and e x t e n t t o which the systems can be deployed.

Since t h e primary goal o f

t h i s i n t e r i m r e p o r t i s t o e s t a b l i s h whether t h e r e i s an i n c e n t i v e f o r DUTH-fueled systems,

L

a r a t h e r r a p i d deployment schedule was assumed so t h a t t h e maximum b e n e f i t s t h a t could be a n t i c i p a t e d from each r e a c t o r / f u e l c y c l e system could be determined.

Systems f o r which

t h e r e i s i n s u f f i c i e n t i n c e n t i v e f o r f u r t h e r development could thus be i d e n t i f i e d and e l i m i n a t e d from f u r t h e r consideration.

Trade-offs between t h e prospects f o r connnercialization, R&D

costs, and deployment schedules and economic/resource i n c e n t i v e s could then be evaluated

i

t

i n g r e a t e r d e t a i l f o r the remaining options.


a

5-18

t

LWRs on Denatured Cycle'

I

I

CALENDARYEAR

1978 1980

DATA BASE DEMu)pMENT DEMO DESIGN AM) LICENSING MMONSTRATI ON

II

1985

1990

1995

2000

2005

I

-

ESTIWTED ($M) ulsTs

40 25 -'05

I

- 160b - 125

I

L L

200

I;

-

H

-. ESTIWTED CALENCARYEAR 1978 1980

1985

1990

1995

2000

2005

Mrn GAS-

(HEU/Th CYCLE)

m (m

ci

u L

PROTOTYPE CONSTRUCTION AND OPERATION DEMO DESIGN AND LICENSING MM) CONSTRUCTION DEMO OPERATION

RFmm (SEU CYCLE)

PROTOTYPE CONSTRUCTION AND OPERATION MM) MSIGN AND LICENSING DEM) coNsTRucT1oN DEM) OPERATION

c

S P F m-SHIFT-allmllm (LEU CYCLE)

DEMO CONSTRUCTION DEMO OPERATION dFirst demonstration unit only. xcludes cost of D20 plant facilities. Incremental costs above PWR costs.

2

Fig. 5.1-1.

R&D Schedules and Costs f o r Government-Supported Demonstration o f Various Reactor Systems

w L L:


5-1 9

The most r a p i d deployment schedule considered t o be feasible was one i n which time was allowed t o resolve t e c h n i c a l problems b u t one t h a t was l a r g e l y unimpeded by c o m e r c i a l i z a t i o n considerations.

The R,D&D schedules t h a t have been presented i n t h i s s e c t i o n a r e

c o n s i s t e n t with t h i s approach.

However, i t i s recognized t h a t the h i g h - r i s k f a c t o r s and

p o t e n t i a l l y unfavorable near-term economics of such a schedule would make i t u n a t t r a c t i v e t o t h e p r i v a t e sector, e s p e c i a l l y f o r those systems r e q u i r i n g l a r g e - p l a n t demonstration. Demonstration program costs are viewed as h i g h l y u n c e r t a i n and dependent upon t h e s p e c i f i c economic i n c e n t i v e s f o r each reactor/cycle concept and on such f a c t o r s as t h e l i c e n s i n g c l i m a t e and general h e a l t h o f t h e i n d u s t r y p r e v a i l i n g a t t h e time o f deployment. Thus t h e costs associated w i t h t h e R,D&D schedules are assumed t o be l a r g e l y government financed.

A comparison o f the t o t a l estimated costs t o the government f o r the various r e a c t o r systems discussed above i s presented i n Table 5.1-3. As noted, t h e R,D&D costs are lowest

Table 5.1-3. Estimated Total Government Support Required f o r Demonstration o f LWRs on DUTH Fuels and Advanced Converters on Various Fuels Total Costs

($MI

Systern 85

- 215a

HTGR; HEU/Th Fuel

560

- 750b

HWR; SEU Fuel

610

- 77ObSc

SSCR; LEU Fuel

190

- 25ObSc

LblR; DUTH Fuels

Comments I n current-generation LWR; no demons t r a t i o n p l a n t required.

Advanced Converters; Reference Fuels

I f DUTH f u e l selected as reference f u e l , a d d i t i o n a l incremental c o s t probably l e s s than c o s t o f converti n g LWRs t o DUTH fuels. A d d i t i o n a l incremental c o s t t o conv e r t t o DUTH f u e l s approximately equal t o t h a t f o r LWR conversion. Could be converted t o DUTH f u e l f o r S l O M $60M i f LWRs already converted.

-

QIncludes 25% subsidy f o r demonstration o f LWR on DUTH f u e l ; excludes f u e l performance program (see Table 5.1-2). bCovers f i r s t demonstration u n i t only; 25% subsidy o f f o u r a d d i t i o n a l u n i t s a n t i c i p a t e d (see Table 5.1-2). C Excludes costs o f heavy-water p l a n t f a c i l i t i e s .

f o r t h e LWR on denatured f u e l because o f the already widespread deployment o f t h i s r e a c t o r concept.

I t i s assumed t h a t a l l basic R&D r e q u i r e d f o r commercialization o f LWRs operat-

i n g on t h e i r reference f u e l c y c l e (LEU) has been completed, and t h a t t h e use o f denatured f u e l can be demonstrated i n current-generation LWRs. as such, w i l l n o t be required.

Thus, an LWR demonstration p l a n t ,

The commitment o f an LWR t o DUTH f u e l s w i l l e n t a i l some

r i s k s , however, and a 25% government subsidy i s assumed t o be necessary f o r a three-year demonstration program.


5-20

~

The R,D&D costs are highest f o r t h e HTGR and HWR, which are y e t t o be demonstrated on t h e i r reference cycles f o r the l a r g e u n i t s i z e (1000-MWe) postulated i n t h i s r e p o r t . The c o s t o f these demonstration u n i t s c o n s t i t u t e s t h e l a r g e s t f r a c t i o n o f t h e t o t a l e s t i mated R,D&D costs, although s u b s t a n t i a l costs w i l l a l s o be incurr'ed f o r l a r g e p l a n t design and l i c e n s i n g , which includes component R&D, component design, and l i c e n s i n g and SAR development. The R,D&D requirements f o r the HTGR and HWR are judged t o be s i m i l a r under the assumption t h a t experience equivalent t o t h a t o f t h e F o r t S t . Vrain HTGR prototype can be obtained from Canadian technology. The SSCR i s viewed as having R,D&D costs intermediate between those o f the LWR and those o f the HTGR because o f t h e heavy r e l i a n c e of the SSCR on LWR technology.

As has been discussed i n t h e t e x t , once these reactors

have been demonstrated on t h e i r reference cycles, a d d i t i o n a l R,D&D w i l l be r e q u i r e d t o convert them t o DUTH f u e l s .

Section 5.1 References 1.

j

"The Economics and U t i l i z a t i o n of Thorium i n Nuclear Power Reactors," Associates, Inc., January 16, 1968 ( d r a f t ) .

Resource Planning


5-21

5.2.

FUEL RECYCLE RESEARCH AND DEVELOPMENT REQUIREMENTS

I . Spiewak Oak Ridge National Laboratory The purpose of t h i s section i s t o summarize the technical problems t h a t must be addressed by a f u e l recycle research and development program before r e a c t o r systems producing

b

and using denatured uranium-thorium (DUTH) f u e l s can be deployed commercially.

Preliminary

estimates of the schedule and costs f o r such a program are a l s o included t o provide some perspective on t h e comnitments t h a t w i l l be required w i t h the i n t r o d u c t i o n o f reactors

1

operating on denatured fuels.

Wide ranges i n the estimates r e f l e c t the c u r r e n t uncertain-

However, d e t a i l e d studies o f the research and development requirements f o r t h e r e c y c l e of DUTH fuels are now being conducted by the DOE Nuclear Power D i v i s i o n ' s

t i e s i n the program.

Advanced Fuel Cycle Evaluation Program (AFCEP),

and when the r e s u l t s from these studies be-

come available, the u n c e r t a i n t i e s i n costs and schedules should be reduced. 5.2.1. ..

L

Technology Status Summary

The technological areas i n a f u e l recycle program cover f u e l fabrication/refabrication ( f u e l m a t e r i a l preparation, rod f a b r i c a t i o n , element assembly); f u e l q u a l i f i c a t i o n ( i r r a d i a t i o n performance t e s t i n g and evaluation); f u e l reprocessing (headend treatment, solvent e x t r a c t i o n , product conversion, off-gas treatment) ; and waste treatment (concentration, c a l c i nation, v i tri f ica tion, and radioact ive-gas treatment). Fuel Fabri cation/Refabri c a t i on and Quali fic a t i on

c

I n general, t h e basic technology f o r the f a b r i c a t i o n o f uranium oxide p e l l e t f u e l s is established, w i t h the f a b r i c a t i o n o f both LWR and HWR uranium f u e l s being conducted on a commercial scale. I n contrast, Pu/U oxide p e l l e t f u e l s have been f a b r i c a t e d o n l y on a small p i l o t - p l a n t scale, and a s i g n i f i c a n t amount of research and development i s s t i l l required. Areas r e q u i r i n g f u r t h e r study include demonstration o f : (1)

a p e l l e t i z i n g process t o ensure uniform product c h a r a c t e r i s t i c s and performance;

(2)

methods f o r v e r i f y i n g and c o n t r o l l i n g the c h a r a c t e r i s t i c s o f the Pu/U fuels;

( 3 ) processes f o r the recovery o f contaminated scrap;

L i

(4)

a r e l i a b l e nondestructive assay system f o r powders, f u e l rods, and wastes;

(5)

t h e a b i l i t y t o operate a large-scale p l a n t remotely, b u t w i t h hands-on maintenance ( i n the case where Pu/U oxides containing high q u a l i t y plutonium are being fabricated); and

(6)

s a t i s f a c t o r y i r r a d i a t i o n performance o f Pu/U f u e l s produced i n commercial-scale processes and equipment.


5-22

I n t h e case o f metal-clad oxide f u e l s t h a t are thorium based, the areas r e q u i r i n g f u r t h e r study are e s s e n t i a l l y t h e same as those l i s t e d above f o r t h e Pu/U oxide fuels; however, i n c o n t r a s t t o Pu/U-oxide fuels, where s i g n i f i c a n t e f f o r t has already been devoted toward r e s o l v i n g t h i s l i s t

0;

areas, r e l a t i v e l y l i t t l e R&D has been performed t o date f o r

thorium-based f u e l s and consequently a l a r g e r amount o f research and development would be required. The intense r a d i o a c t i v i t y o f the decay daughters o f 232U (which i s produced i n t h e thorium along w i t h t h e 233U) requires t h a t the r e f a b r i c a t i o n processes a l l be remotely operated and maintained. This requirement w i l l necessitate a d d i t i o n a l development o f t h e r e f a b r i c a t i o n processes and may r e q u i r e t h e development o f new f a b r i c a t i o n methods. The q u a l i f i c a t i o n o f U/Th and Pu/Th oxide f u e l s w i l l a l s o r e q u i r e a d d i t i o n a l R&D efforts. HTGR f u e l s are coated uranium oxide o r carbide microspheres embedded i n a graphite

f u e l element.

The process and equipment concepts f o r r e f a b r i c a t i n g HTGR fuel remotely have been i d e n t i f i e d ; however, a d d i t i o n a l R&D p r i o r t o c o n s t r u c t i o n o f a h o t demonstrat i o n f a c i l i t y i s needed. This should cover:

(1)

t h e scaleup o f r e f a b r i c a t i o n equipment,

(2)

t h e r e c y c l e o f scrap material,

( 3 ) t h e c o n t r o l o f e f f l u e n t s , and (4)

t h e assay o f f u e l - c o n t a i n i n g materials.

A d d i t i o n a l R&D w i l l a l s o be required f o r q u a l i f i c a t i o n o f t h e r e c y c l e fuel. While the reference HTGR f u e l c y c l e already includes thorium, f u r t h e r development work

w i l l be r e q u i r e d t o f a b r i c a t e DUTH f u e l s f o r HTGRs because o f t h e requirement of a higher uranium content o f the f i s s l e p a r t c l e and the increased production o f plutonium d u r i n g irradiation. Fuel Reprocessing The basic technology f o r reprocessing o f uranium and uranium/plutonium oxide p e l l e t f u e l s w i t h low burnup e x i s t s i n the Purex process.

This technology i s based on many years

o f government reprocessing experience w i t h m i l i t a r y - r e l a t e d fuels;

however, a commercial

reprocessing p l a n t f o r mixed oxide power r e a c t o r f u e l s t h a t conforms t o c u r r e n t U.S. and s t a t e requirements has n o t y e t been operated.

federal

A d d i t i o n a l l y , w h i l e engineering o r

p i l o t - s c a l e work has been s u c c e s s f u l l y c a r r i e d o u t on â&#x20AC;&#x2DC; a l l important processes and components o f t h e reprocessing p l a n t , o p e r a b i l i t y , r e l i a b i l i t y , and costs o f an i n t e g r a t e d p l a n t have n o t been demonstrated i n a l l cases a t f u e l exposures expected i n commercial reactors. S p e c i f i c areas t h a t s t i l l r e q u i r e development work i n c l u d e the f o l l o w i n g : (1 )

operation and maintenance o f the mechanical headend equipment;

( 2 ) â&#x20AC;&#x2DC;methods f o r handling h i g h l y r a d i o a c t i v e residues t h a t remain a f t e r the d i s s o l u t i o n o f high-burnup fuel;

( 3 ) t h e technology f o r reducing r a d i o a c t i v e off-gas releases (e.g., t r i t i u m ) t o conform t o a n t i c i p a t e d regulations;

Kr-85,

i o d i n e and


5-23

remotely operated and d i r e c t l y maintained conversion processes f o r plutonium from

(4)

8,

1'

power r e a c t o r fuels; and

\

(5) high-level waste s o l i d i f i c a t i o n and v i t r i f i c a t i o n t o prepare f o r terminal storage. The technology f o r reprocessing thorium-based oxide p e l l e t f u e l s i s l e s s advanced than t h a t f o r uranium-based fuels.

The Thorex process has been used t o process i r r a d i a t e d t h o r i -

um oxide f u e l s o f low burnup i n government p l a n t s and i n l i m i t e d q u a n t i t i e s i n a small-scale i n d u s t r i a l plant.

Thorium oxide f u e l s have not been processed i n a large-scale p l a n t s p e c i f -

i c a l l y designed f o r thorium processing, nor has h i g h l y i r r a d i a t e d thorium oxide f u e l

I,

been processed by the Thorex process i n engineering-scale equipment. The p r i n c i p a l differences between the reprocessing development required t o reprocess

b L i

-

L

ii

metal-clad thorium-based oxide f u e l s and graphite-based HTGR f u e l occur i n the headend treatment. P a r t i t i o n i n g o f f u e l materials from both classes o f r e a c t o r f u e l can then be accomplished by a Thorex-type solvent e x t r a c t i o n process. I n the case o f metal-clad oxide fuels, a d d i t i o n a l headend process R&D i s required t o determine how zirconium cladding can be removed and the ThOZ fuel dissolved.

Significant

waste handling problems may be encountered i f f l u o r i d e i s required t o .dissolve Thoz. I n t h e case of the headend process development f o r graphi te-based HTGR fuels, development work i s needed w i t h i r r a d i a t e d m a t e r i a l s i n the crushing, burning and p a r t i c l e separation operations, and i n t h e treatment o f " T - c o n t a i n i n g off-gases associated w i t h the headend o f the reprocessing p l a n t .

I

S p e c i f i c areas o f solvent e x t r a c t i o n process development work required t o reprocess a1 1 t h o r i um-containing r e a c t o r fuel include: (1) (2)

f u e l d i s s o l u t i o n , feed adjustment, and c l a r i f i c a t i o n ; technology development f o r containing 220Rn and other r a d i o a c t i v e gases t o conform t o regulations;

( 3 ) recovery o f f u l l y i r r a d i a t e d thorium i n large-scale f a c i l i t i e s ; (4) p a r t i t i o n i n g o f f u e l s o l u t i o n s containing U, Pu, and Th;

(5) recovery and handling o f h i g h l y r a d i o a c t i v e product streams;

u

( 6 ) . process and equipment design i n t e g r a t i o n ; and

(7)

h i g h - l e v e l waste concentration and v i t r i f i c a t i o n .

'

Waste Treatment

L

Waste treatment R&D requirements common t o a l l f u e l cycles i n v o l v e development o f the technology needed f o r i m m o b i l i t l n g high-level and intermediate-level s o l l d and gaseous wastes.

c,,

L

E

Processes f o r concentration, c a l c i n a t i o n , and v i t r i f i c a t i o n o f these are needed.

The waste treatment requirements f o r the various f u e l cycles are s i m i l a r , b u t they would be more complex f o r t h e thorium-based cycles i f f l u o r i d e s were present i n t h e wastes.


5-24

5.2.2.

Research,

Development, and Demonstration Cost Ranges and Schedules

While f u e l r e c y c l e R&D needs can be i d e n t i f i e d f o r a v a r i e t y o f a l t e r n a t e f u e l cycles and systems, t h e launching o f a major developmental e f f o r t t o i n t e g r a t e these a c t i v i t i e s i n t o a s p e c i f i c i n t e g r a t e d f u e l c y c l e must await a U.S. decision on t h e f u e l c y c l e and r e a c t o r development s t r a t e g y t h a t would best support our n o n p r o l i f e r a t i o n object i v e s and our energy needs. Whether i t would be more expeditious t o develop i n d i v i d u a l cycles independently i n separate f a c i l i t i e s o r t o plan f o r an i n t e g r a t e d r e c y c l e development f a c i l i t y w i l l depend on t h e nature and t i m i n g o f t h a t decision. I f a number o f r e l a t e d cycles were developed i n the same f a c i l i t i e s , the t o t a l costs would be o n l y moderately higher than t h e costs associated w i t h any one cycle. Since the denatured 23311 c y c l e i m p l i e s a system o f symbiotic reactors (233U producers and 233U consumers), such an approach i s l i k e l y t o be a t t r a c t i v e i f a decision were made t o develop t h e denatured

233U

L;

u

c

li

cycle.

The existence of major u n c e r t a i n t i e s i n the fuel r e c y c l e development and demonstration programs make c o s t p r o j e c t i o n s h i g h l y uncertain. There are, f i r s t , d i f f i c u l t i e s i n h e r e n t i n p r o j e c t i n g the costs of process and equipment development programs which address t h e resolut i o n of t e c h n i c a l problems associated w i t h p a r t i c u l a r reactors and f u e l cycles. I n addition, there are u n c e r t a i n t i e s conunon t o p r o j e c t i n g costs and schedules f o r a l l f u e l r e c y c l e development programs; s p e c i f i c a l l y , u n c e r t a i n t i e s i n the future s i z e o f the commercial nuclear i n d u s t r y cause problems i n program d e f i n i t i o n . I t i s necessary t o i d e n t i f y the r e a c t o r growth scenario associated w i t h t h e f u e l c y c l e system so t h a t f u e l loads can be p r o j e c t e d and t y p i c a l p l a n t sizes estimated. This i s c r i t i c a l from t h e standpoint o f e s t a b l i s h i n g t h e scale o f t h e technology t o be developed and the p r i n c i p a l steps t o be covered i n t h e development.

For example, i f t h e end use o f a f u e l c y c l e i s i n a secure energy center, smaller p l a n t s are i n v o l v e d and t h e development could conceivably be terminated w i t h a p l a n t t h a t would be considered a prototype i n a l a r g e (1500 MT/yr) commercial reprocessing f a c i l i t y development sequence.

S i m i l a r l y , growth r a t e s f o r p a r t i c u l a r r e a c t o r types may be

much smaller than others, o r t h e f u e l loads may be smaller because o f higher f u e l burnup. Thus, smaller f u e l c y c l e p l a n t s would be required. The problem i s f u r t h e r complicated by the f a c t t h a t the f u e l r e c y c l e i n d u s t r y has f o r a number o f years been confronted w i t h uncertain and e s c a l a t i n g r e g u l a t o r y requirements. Permissible r a d i a t i o n exposure l e v e l s f o r operating personnel, acceptable safeguards

L

systems, and environmental and s a f e t y requirements, a l l o f which a f f e c t costs, have n o t been specified. Nevertheless, based upon experience w i t h previous f u e l r e c y c l e development programs, t y p i c a l f u e l r e c y c l e R,D&D costs f o r the f u e l cycles o f i n t e r e s t can be presented i n broad ranges.

I n the past, reprocessing costs had been developed f o r t h e U/Pu systems with p a r t i t i o n e d and decontaminated product streams. These have been used here t o provide base-line costs. Any i n s t i t u t i o n a l consideration, such as a secure f u e l service

center, t h a t would permit conventional Purex and Thorex reprocessing t o take p l a c e would g i v e more credence t o the base-line technology development costs used here.

L


t

5-25

i

Estimated c o s t ranges and times f o r the development and commercialization of a new reprocessing technology and a new r e f a b r i c a t i o n technology a r e presented i n Tables 5.1 and 5.2 r e s p e c t i v e l y . From these tables, i t can be seen t h a t t h e t o t a l c o s t t o t h e federal government t o develop a new reprocessing technology would range between $0.8 b i l l i o n and $2.0 b i l l i o n . The corresponding cost f o r a new r e f a b r i c a t i o n technology would be

I;

I

Table 5.2-1. Estimated Cost Range f o r Development and Commercialization o f a T y p i c i l New iieprocessing Technology

1

Base technology R&D Hot p i l o t p l a n t t e s t i n g

2.1

0.5 0.6 0.2 0.8

Subtotal Large-scale c o l d prototype t e s t i n g b Total Large-scale demonstration p l antc

(1,O

p o r t f o r i n i t i a l demonstration facilities.

treatment techno1 ogy development

l a.Estimated t a p e d time requirements from i n i t i a l devetopment through demonstnation ranges from 1 2 y m r s : f o r established technology t o 20 years f o r new technology. bGovernaent might i n c u r costs o f t h i s magnitude as D a r t o f demonstration program. c Comnercial f a c i l i t y e x t e n t o f government p a r t i c i p a t i o n d i f f i c u l t t o d e f i n e a t t h i s time.

Table 5.2-2. Estimated Cost Range f o r Development and Demonstration o f a Typigal New Refabri c a t i on Techno1ogy

Base technology

0.1

Cold component testin: I r r a d i a t i o n performance t e s t i n g Total Large-scal e demonstration

b

- 0.3

- 0.4 0.1 - 0.4 0.4 - 1.1 (0.7 - 1.4) 0.â&#x20AC;&#x2122;2

â&#x20AC;&#x153;Estimated lapsed time requirements from i n i t i a l development through demonstration ranges from about 8 10 years f o r technology near t h a t bestablished t o about 15 years f o r new technology. e x t e n t of government Comnercial f a c i 1ity p a r t i c i p a t i o n d i f f i c u l t t o d e f i n e a t t h i s time.

-

-

To these costs should

be added the costs o f t h e waste

- 3.0)

-

c o n s t r u c t i o n and operation o f p i l o t plants, development o f largescale prototype equipment, and sup-

- 0.5 - 1.0 - 1.5 - 0.5 - 2.0

Unescalated b i l l i o n s o f Dollars

1

the costs t r a d i t i o n a l l y borne by t h e government i n c l u d e b a s i c R&D,

Unescalated B i l l i o n $ of Dollars I

L ir

between $0.4 b i l l i o n and $1.1 b i l l i o n . For f u e l r e c y c l e development,

needed t o close t h e f u e l cycle. The c a p i t a l costs estimated f o r a comnercial demonstration f a c i l i t y a r e l i s t e d separately i n Tables 5.1 and 5.2 because t h e e x t e n t t h a t t h e government might support these f a c i l i t i e s i s unknown.

Since they w i l l be

commercial f a c i l i t i e s , costs i n c u r r e d e i t h e r by t h e government o r by a p r i v a t e owner could be recovered i n fees.

The t o t a l

c a p i t a l costs might range between $1.0 b i l l i o n and $3.0 b i l l i o n f o r a l a r g e reprocessing demonstration f a c i l i t y and between $0.7 b i l l i o n and $1.4 b i l l i o n f o r a r e f a b r i c a t i o n demonstration f a c i 1it y

.

Tables 5.1 and 5.2 show t h a t the major costs associated w i t h

,

comnercialization o f f u e l cycles l i e a t t h e f a r end o f t h e R&D progression, namely, i n the steps i n v o l v i n g p i l o t plants, large-scale prototype equipment development, and demonstration plants, i f required. The r a t e and sequencing of R&D expenditures can be i n f e r r e d from Tables 5.2-1 and 5.2-2. process and equipment concepts may r e q u i r e 2-6 years.

Base technology R&D t o i d e n t i f y The engineering phase o f the development


5-26

program, i n c l u d i n g h o t t e s t i n g , may r e q u i r e 5-12 years. Reference f a c i l i t y design and cons t r u c t i o n might r e q u i r e 8-12 years. There can be considerable overlapping o f phases so t h a t f o r a given f u e l c y c l e the t o t a l lapsed time from i n i t i a l development t o commercialization o f f u e l r e c y c l e ranges from about 12-20 years. The t o t a l time would depend upon t h e i n i t i a l technology status, the degree t o which the R&D program steps are telescoped t o save time, and The tHqrium cycles would be a t

the stage t o which the development program must be c a r r i e d . the f a r end o f the development time range.

Table 5.2-3 presents t h e R&D cost ranges i n terms o f r e a c t o r types and f u e l r e c y c l e systems.

For a l l f u e l cycles, the u n c e r t a i n t y i n the R&D costs should be emphasized.

Thus,

i n water reactors, the estimated range o f R&D costs i s $1.3-2.3 b i l l i o n f o r U/Pu r e c y c l e development, and $1 -8-3.3 b i l l i o n f o r DUTH r e c y c l e development. For HTGRs, t h e correspondi n g ranges are $1.4-2.6

b i l l i o n and $1.8-3.3

b i l l i o n f o r U/Pu and DUTH r e c y c l e development,

respectively; f o r FBRs, the corresponding ranges are $1.6-3.0

b i l l i o n and $2.0-3.6

billion,

respectively.

Although there i s a s i g n i f i c a n t c o s t u n c e r t a i n t y f o r each r e a c t o r type and f u e l cycle, f o r a given r e a c t o r type the t r e n d i n costs as a f u n c t i o n o f f u e l c y c l e i s significant.

Generally, the reference U/Pu c y c l e would be l e a s t expensive and t h e DUTH

c y c l e the most expensive, w i t h t h e Pu/Th and HEU/Th cycles intermediate.

Table 5.2-3.

Estimated Range o f Fuel Recycle R&D Costs* B i l l i o n s of Dollars

Reactor Type

PuITh

U/PU

DUTH

HEU/Th

Water Reactors HTGRs

1.3-2.3 1.4-2.6

1.6-3.0 1.6-3.0

1.8-3.3 1.8-3.3

1.6-2.9 1.6-2.9

FBRs

1.6-3.0

1.8-3.2

2.0-3.6

1.7-3.1

*Includes costs f o r developing reprocessing and r e f a b r i c a t i o n technologies and a p o r t i o n o f t h e waste treatment technology development costs. Conclusions

5.2.3.

A decision t o develop r e a c t o r systems operating on denatured f u e l cycles requires a

government commitment t o spend $0.5 b i l l i o n t o $2 b i l l i o n more on a f u e l r e c y c l e development program than would be r e q u i r e d t o develop reactors operating on t h e reference ( p a r t i t i o n e d , uncontaminated products) U/Pu cycles.

The d i f f e r e n t i a l i s even l a r g e r when

reactors operating on DUTH cycles are compared w i t h reactors operating on once-through cycles.

No comparison has been made w i t h the costs o f developing d i v e r s i o n - r e s i s t a h t U/Pu

cycles (using co-processing, spiking, etc.). Expenditures t o develop r e c y c l e systems f o r DUTH f u e l s would span a p e r i o d o f 20 years from i n i t i a l development t o commercialization.

The p r i n c i p a l expenditures would

occur i n t h e second h a l f of t h i s period, when l a r g e f a c i l i t i e s w i t h h i g h operating costs are needed.

t


CHAPTER 6 EVALUATION OF NUCLEAR POWER SYSTEMS UTILIZING DENATURED FUEL M. R. Shay, D. R. Haffner, W. E. Black, T. M. Helm, W. G. J o l l y , R. W. Hardie, and R. P. Omberg

Hanford Engineering Development Laboratory

Chapter Out1 i n e 6.0.

Introduction

6.1.

Basic Assumptions and Analysis Technique 6.1 .l. 6.1.2. 6.1.3. 6.1.4.

6.2.

Discussion o f Results f o r Selected Nuclear P o l i c y Options 6.2.1. 6.2.2. 6.2.3. 6.2.4. 6.2.5. 6.2.6. 6.2.7.

6.3.

The U30, Supply Reactor Options Nuclear P o l i c y Options The A n a l y t i c a l Method

The Throwaway/Stowaway Option Converter System w i t h Plutonium Recycle Converter System w i t h Plutonium Throwaway Converter System w i t h Plutonium Production Minimized; P u - ~ o - ~ "Transmutation" ~ ~ U Converter System with Plutonium Production Not Minimized; P u - ~ o -3U~ ~"Transmutation" Converter-Breeder System w i t h L i g h t Plutonium 'Transmutation" Converter-Breeder System w i t h Heavy P1u t o n i um "Transmutation"

Conclusions


c

c


6-3

6.0.

INTRODUCTION

I n t h i s chapter c i v i l i a n nuclear power systems t h a t u t i l i z e denatured 233U f u e l t o various degrees are analyzed t o determine whether they could meet p r o j e c t e d nuclear power The r e a c t o r s employed i n the systems

demands w i t h t h e ore resources assumed t o be a v a i l a b l e .

are those discussed i n e a r l i e r chapters o f t h i s r e p o r t as being the reactors most l i k e l y t o be developed s u f f i c i e n t l y f o r commercial deployment withi-n t h e planning horizon, which i s assumed t o extend t o the year 2050. The reactors included are L i g h t Water Reactors (LWRs), S p e c t r a l - S h i f t - C o n t r o l l e d Reactors (SSCRs) , Heavy Water Reactors (HWRs) , HighlTemperature Gas-Cooled Reactors (HTGRs), and Fast Breeder Reactors (FBRs). I n each case, t h e nuclear power system i s i n i t i a t e d G i t h c u r r e n t l y used LWRs operating on the low-enriched 235U f u e l cycle, and o t h e r converter reactors and/or f u e l cycles are added as they become a v a i l a b l e . On t h e basis o f i n f o r m a t i o n provided by the r e a c t o r designers, i t i s assumed t h a t 235U-fueled LWRs alone w i l l be u t i l i z e d through the 1980s and t h a t LWRs operating on denatured 233U and 239Pu w i l l become a v a i l a b l e i n t h e e a r l y 1990s. It i s a l s o assumed t h a t SSCRs operating on t h e various f u e l cycles w i l l become a v a i l a b l e i n t h e e a r l y 1990s. Thus nuclear power systems c o n s i s t i n g o f LWRs alone o r o f LWRs and SSCRs i n combination, w i t h several f u e l c y c l e options being a v a i l a b l e , could be introduced i n t h e e a r l y 1990s. LWR-HWR and LWR-HTGR systems could be expected i n t h e mid 199Os, and FBRs could be added t o any o f the systems a f t e r the year 2000. The nuclear power systems u t i l i z i n g denatured 233U f u e l were +vided i n t o two major categories: those c o n s i s t i n g o f thermal converter reactors o n l y ar.d those c o n s i s t i n g o f both thermal converters and f a s t breeders. Three "nuclear p o l i c y options" were examined under each category, the i n d i v i d u a l options d f f f e r i n g p r i m a r i l y i n t h e e x t e n t t o which plutonium i s produced and used t o breed a d d i t i o n a l f i s s i l e m a t e r i a l . For comparison, a throwawaylstowaway o p t i o n employing LEU converters was a l s o analyzed, and two options u t i l i z i n g the c l a s s i c a l plutonium-uranium c y c l e were studied, one using converters o n l y and the o t h e r using both converters and breeders. A l l o f the options studied were based on t h e concept of secure energy centers and dispersed r e a c t o r s discussed i n previous chapters. Thus, a l l enrichment, reprocessing, and

f u e l f a b r i c a t i o n / r e f a b r i c a t i o n a c t i v i t i e s , as w e l l as f u e l and/or waste storage, were assumed t o be confined t o t h e energy centers. I n addition, a l l r e a c t o r s operating on plutonium o r h i g h l y enriched uranium were assigned t o t h e centers, w h i l e reactors operating on low-enriched o r denatured uranium were permitted t o be outside the centers. Determining t h e p r e c i s e nature and s t r u c t u r e of t h e energy center was n o t w i t h i n the scope o f t h i s study. Presumably i t c o u l d be a r e l a t i v e l y small l o c a l i z e d area o r a l a r g e geographical region covering an e n t i r e nuclear state, o r even a c o l l e c t i o n of nuclear states. I f more than one country were involved, t h e s e n s i t i v e f a c i l i t i e s could be n a t i o n a l l y owned b u t operated under i n t e r n a t i o n a l safeguards. But whatever t h e character o f t h e center an important consideration f o r any nuclear p o l i c y opt i o n i s i t s "energy support r a t i o , " which i s defined as t h e r a t i o o f t h e nuclear capacity i n s t a l l e d outside t h e center t o t h e capacity i n s t a l l e d i n s i d e t h e center. Only as t h e supp o r t r a t i o increases above u n i t y i s t h e c a p a b i l i t y o f t h e system t o d e l i v e r power t o d i s -


6-4

- a fact

which i s p a r t i c u l a r l y important i f nuclear s t a t e s are planning t o provide nuclear f u e l assurances t o nonnuclear states. persed areas ensured

The philosophy used i n t h i s study i s i l l u s t r a t e d i n Fig. 6.0-1.

U308 supply and a s p e c i f i e d s e t o f r e a c t o r

ClVEN SRCIFIED

Given a s p e c i f i e d

UpsSUPPLY

development options, t h e p o t e n t i a l r o l e o f

NUCLEAR GROWM POKNTIAL

nuclear power, t h e resources r e q u i r e d t o

a a a

SPECIFIED REAClOR DMLOPMENT OPTIONS

achieve t h i s r o l e , and t h e composition and movement o f f i s s i l e m a t e r i a l were calculated.

aEk

The deployment o f t h e indiv'idual r e a c t o r s and

22 Ej

t h e i r associated f u e l c y c l e f a c i l i t i e s were i n a l l cases c o n s i s t e n t w i t h t h e nuclear

TIME

RESOURCE REQUIRWNTS AND FISSILE M E R I A L LOCATION

-&fbT3!4 I

pol i c y o p t i o n under consideration. The i n t r o duction date f o r each i n d i v i d u a l r e a c t o r concept and f u e l c y c l e f a c i l i t y was assumed t o be

MDL 7ID2-W I

t h e e a r l i e s t t e c h n o l o g i c a l l y f e a s i b l e date. This allows an e v a l u a t i o n o f t h e maximum i m pact o f t h e system on any p a r t i c u l a r nuclear

Fig. 6.0-1. The Philosophy o f the Nuclear Systems Assessment Study.

option. The e f f e c t o f delaying t h e deployment o f a r e a c t o r / c y c l e because i t produces undesirable consequences was determined simply by e l i m i n a t i n g i t from t h e option. It was assumed t h a t a nuclear power system was adequate i f i t s i n s t a l l e d nuclear capacity

was 350 GWe i n t h e y e a r 2000 and a n e t increase o f 15 GWe/yr was r e a l i z e d each y e a r t h e r e a f t e r , w i t h t h e increase sustained by t h e U308 supply. used i n t h e study.

Two d i f f e r e n t o p t i m i z i n g p a t t e r n s were

A few runs were made assuming economic competition between nuclear

f u e l and coal, t h e p l a n t s being selected t o minimize t h e l e v e l i z e d c o s t o f power over time.

These runs, described i n Appendix D, i n d i c a t e d t h a t f o r t h e assumptions used i n

t h i s a n a l y s i s nuclear power d i d n o t compete w e l l a t U3O8 p r i c e s above $160/lb; therefore, i n t h e remaining runs an attempt was made t o s a t i s f y t h e demand f o r nuclear power w i t h U308 a v a i l a b l e f o r l e s s than $160/lb U308. I t i s these runs t h a t are described i n t h i s chapter. The s p e c i f i c assumptions regarding t h e

U3O8

supply are presented i n Section 6.1 below,

which a l s o includes d e s c r i p t i o n s o f t h e operating c h a r a c t e r i s t i c s o f t h e i n d i v i d u a l r e a c t o r s u t i l i z e d , t h e various nuclear p o l i c y options chosen f o r analyses, and t h e a n a l y t i c a l method applied.

Section 6.2 then compares t h e r e s u l t s obtained f o r a selected s e t o f nuclear p o l i c y

options, and Section 6.3 summarizes t h e conclusions reached on t h e basis o f those comparisons. The economic data base used f o r these studies i s given i n Appendix B y and d e t a i l e d r e s u l t s f o r a l l t h e nuclear p o l i c y options are presented i n Appendix C.


6-5

6.1

.. BASIC ASSUMPTIONS AND ANALYSIS TECHNIQUE 6.1.1.

The U308 Supply

The most recent estimates o f t h e supply o f U308 a v a i l a b l e i n the United States as re-

id

1 8;

L

p o r t e d by DOE'S D i v i s i o n o f Uranium Resources and Enrichment (URE) are summarized i n Table 6.1-1 (from r e f . 1).

On the basis o f a maximum forward c o s t o f $50/lb, t h e known reserves

plus probable p o t e n t i a l resources t o t a l 2,325 x 103 ST.

lo3

140 x

ST i s a v a i l a b l e from byproducts (phosphates and copper), so t h a t the amount o f

U308 probably a v a i l a b l e t o t a l s 2.465 x l o 3 ST ( o r approximately 2.5 m i l l i o n ) . I f the "possible" and "speculative" resources are also considered, the URE estimates a r e increased t o approximately 4.5 m i l l i o n ST. Neither o f these estimates i n c l u d e U308 which may be a v a i l a b l e from o t h e r U.S. sources, such as the Tennessee shales, o r from o t h e r nations.*

The actual

U3O8

supply curves used i n the analysis were based on the long-run marginal

costs of e x t r a c t i n g U308 r a t h e r than the forward costs.

L

t B

The long-run marginal costs con-

t a i n the c a p i t a l costs o f f a c i l i t i e s c u r r e n t l y i n operation plus a normal p r o f i t f o r t h e industry; thus they a r e probably more appropriate f o r use i n a nuclear s t r a t e g y analysis. The actual long-run marginal costs used i n t h i s analysis are shown i n Table 6-7 o f Appendix

B and are p l o t t e d i n Fig. 7.4-1 i n Chapter 7. o f the

U3O8

These sources show t h a t i f the r e c o v e r a b i l i t y

supply i s such t h a t l a r g e q u a n t i t i e s can be e x t r a c t e d o n l y a t high costs, then

t h e supply a v a i l a b l e a t a c o s t o f l e s s than $160/lb i s probably no more than 3 m i l l i o n ST. If, however, t h e r e c o v e r a b i l i t y i s such t h a t t h e e x t r a c t i o n costs f a l l i n what i s considered t o be an intermediate-cost range, then as much as 6 m i l l i o n ST a c o s t o f l e s s than $160/lb.

The r a t e a t which t h e

U3O8

could be a v a i l a b l e a t

I n t h e remainder o f t h i s study, these two assumptions are

r e f e r r e d t o as "high-cost" and "intermediate-cost"

hi

URE estimates t h a t an a d d i t i o n a l

U3O8

U3O8

supply assumptions.

resource i s e x t r a c t e d i s a t l e a s t as important as t h e s i z e

o f the resource base. URE has estimated t h a t i t would be d i f f i c u l t f o r the U.S. t o mine and m i l l more than 60,000 ST of U3O8 per year i n the 1 9 9 0 ' s ( r e f . 3). (Note: This estimate was based on developing reserves and p o t e n t i a l resources a t forward costs o f l e s s than

$30/lb. These costs do n o t i n c l u d e c a p i t a l costs o f f a c i l i t i e s o r i n d u s t r y p r o f i t s . ) Although the combined maximum c a p a b i l i t y o f a c o a l i t i o n o f states may exceed t h i s , i t i s n o t p o s s i b l e t o s p e c i f y a d e f i n i t e upper l i m i t u n t i l more i s known about t h e l o c a t i o n s o f t h e sources o f

U3O8

and t h e d i f f i c u l t i e s encountered i n recovering i t .

Recognizing t h i s ,

and a l s o recognizing t h a t the annual capacity i s s t i l l an important variable, t h e nuclear p o l i c y options analyzed i n t h i s study were considered t o be more feasible i f t h e i r annual t mining and m i l l i n g r a t e was l e s s than 60,000 ST o f U3O8 per year. * E d i t o r ' s Note:

li

I n 1977 the U.S.

produced 15,000 ST o f

U3O8

concentrate (ref. 2).

'Editor's Note: I n 1977 the U.S. gaseous d i f f u s i o n p l a n t s produced 15.1 m i l l i o n kg SWU per y e a r ( r e f . 4). A f t e r completion o f t h e cascade improvement program ( C I P ) and cascade upd a t i n g program (CUP) i n t h e 1 9 8 0 ' ~t h~e U.S. capacity w i l l be 27.4 m i l l i o n kg SWU p e r year (refs. 5 and 6). A gas c e n t r i f u g e add-on o f 8.8 m i l l i o n SWU has been proposed f o r the government-owned enrichment f a c i l i t y a t Portsmouth, Ohio. Considerable enrichment capacity a1 so e x i s t s abroad; therefore, enrichment capacity i s i n h e r e n t l y a l e s s r i g i d c o n s t r a i n t than uranium requirements o r production c a p a b i l i t i e s .


6-6

6.1.2.

Reactor Options

The r e a c t o r designs included i n t h i s study have n o t been optimized t o cover every conceivable nuclear p o l i c y option.

Such a task i s c l e a r l y impossible u n t i l t h e options have

been reduced t o a more manageable number.

However, the designs selected have been developed

by using d e t a i l e d design procedures and they a r e more than adequate f o r a r e a c t o r s t r a t e g y study such as i s described here. Table 6.1-1.

Known

Probable

Possible

Speculative

t

c

Estimates o f U s 0 8 Supply A v a i l a b l e i n U.S.A.a Resources ( l o 3 ST)

Forward cost ($71 b)

c

To t a l

I]

~

15

360

485

165

1,570

30

690

1,065

1,120

41 5

3,290

50b

875

1,450

1,470

5 70

4 ,365

560

aFrom r e f . 1.

b A t $50/lb, t h e known reserves o f 875 x lo3 ST p l u s t h e probable reserves o f 1,450 x 103 ST p l u s 140 x 103 ST from byproducts (phosphates and copper) t o t a l 2,465 x 103 ST ( o r % 2.5 m i l l i o n ST). Ift h e p o s s i b l e and speculative resources are included, t h e t o t a l i s increased t o 4,505 x l o 3 ST ( o r % 4.5 m i l l i o n ST).

L

I:

LWRs, represented by Pressurized Water

Four general types of reactors are included:

Reactors (PWRs) ; HWRs, represented by Canadian Deuterium Uranium Reactors (CANDUS) ; High Temperature Gas Cooled Reactors (HTGRs); and Fast Breeder Reactors (FBRs).

The data f o r t h e

PWRs were provided by Combustion Engineering (CE) and Hanford Engineering Development Labo r a t o r y (HEDL); t h e data f o r t h e CANDUs by Argonne National Laboratory (ANL); t h e data f o r the HTGRs by General Atomic (GA);

and t h e data f o r t h e FBRs by HEDL.

I n addition t o the

standard LWRs (PWRs) , spectral-shift-controlled PWRs (SSCRs) a r e a l s o included i n t h e study, t h e data f o r t h e SSCRs being provided by CE. Descriptions o f the i n d i v i d u a l reactors used i n the study a r e given i n Tables 6.1-2 and 6.1-3

( r e f . 7), and t h e economic data base f o r

each i s given i n Appendix B. The LWR designs i n c l u d e reactors f u e l e d w i t h low-enriched and denatured 23511, denatured and plutonium, t h e d i l u e n t f o r t h e denatured designs c o n s i s t i n g o f e i t h e r 2381) or thorium, o r both. I n a d d i t i o n , a low-enriched LWR design optimized f o r throwaway has been

233U,

studied, and a l s o t h r e e SSCRs f u e l e d w i t h low-enriched 23511, denatured 2 3 3 U , and Pu/Th. The HWRs a r e represented by three 235U-fueled reactors ( n a t u r a l , s l i g h t l y enriched, ~ U and a Pu/Th r e a c t o r . The HTGR and denatured), a denatured 233U reactor, a P U / ~ ~reactor, designs c o n s i s t o f low-enriched,

denatured, and h i g h l y enriched z 3 5 U reactors; denatured*

and h i g h l y enriched 233U reactors; and a Pu/Th reactor. The FBR designs c o n s i s t o f two P U / * ~ ~ core U designs (one w i t h a 238U b l a n k e t and one w i t h a thorium blanket) and one Pu/Th core design ( w i t h a thorium blanket). 233U/238U

core design w i t h a thorium b l a n \ e t has been studied.

I n addition, a

The 233U enrichment i s l e s s

than 12%, and thus t h i s FBR i s a denatured design. *In c o n t r a s t t o t h e o t h e r r e a c t o r types, t h e denatured 233U HTGR design i s assumed t o c o n t a i n 15% 2 3 3 U i n 238U i n s t e a d o f 12%.

L L E L I; I, L


6-7

I n t r o d u c t i o n dates f o r each r e a c t o r type are included i n Table 6.1-2.

A s l i g h t modifica-

t i o n t o an e x i s t i n g PWR f u e l design, such as a t h i c k e r f u e l p i n cladding t o extend t h e d i s charge exposure, was introduced i n 1981. A more extensive modification, such as a denatured 235U PWR f u e l pin, was delayed u n t i l 1987. The remaining PWR designs, i n c l u d i n g t h e SSCRs,

L

1 1

were introduced i n 1991.

The HWRs and HTGRs were a l l introduced i n 1995, w h i l e t h e FBRs

were n o t introduced u n t i l 2001. The lifetime-averaged 233U, 235U, and f i s s i l e plutonium flows given i n Table 6.1-3 show t h a t f o r t h e throwaway cycle, low-enriched HTGRs o f f e r s i g n i f i c a n t ( a l w x t 20%) uranium o r e savings compared t o low-enriched PWRs. S l i g h t l y enriched HWRs reduce uranium ore requirements by an a d d i t i o n a l 20% over HTGRs and more than 35% over LWRs.

Although low-enriched

LWRs and HTGRs have roughly t h e same enrichment requirements, t h e s l i g h t l y enriched HWRs r e q u i r e 5 t o 6 times l e s s enrichment. The low-enriched SSCR o f f e r s about a 22% savings i n e n r i chment

.

Core discharge exposures f o r FBRs are approximately t w i c e t h e exposures f o r LWRs,

I

w h i l e exposures f o r HWRs are about h a l f those f o r LWRs.

i

uranium HWR, which has a discharge exposure o f one-fourth t h a t f o r t h e LWR. HTGR d i s charge exposures are extremely l a r g e - n e a r l y 200 MWd/kg f o r t h e Pu/Th fuel design!

An exception i s t h e n a t u r a l -

b r

I

b

The two FBRs w i t h Pu-U cores have breeding r a t i o s o f 1.34 t o 1.36. Replacing t h e uranium i n t h e core w i t h thorium reduces t h e breeding r a t i o by 0.15, w h i l e r e p l a c i n g t h e plutonium w i t h 233U reduces t h e breeding r a t i o by 0.16.

F i n a l l y , comparing 235U-fueled

thermal reactors w i t h 233U-fueled reactors shows t h a t t h e 233U-fueled reactors have conversion r a t i o s about 0.10 t o 0.15 higher.

L

shown

The most s t r i k i n g observation t h a t can be made from t h e t o t a l f i s s i l e f u e l requirements n Table 6.1-3 i s t h e s i g n i f i c a n t l y lower f i s s i l e requirements f o r t h e denatured 233U-

f u e l e d SSCRs and HWRs and f o r t h e h i g h l y enriched 233U/Th-fueled HTGR. F i n a l l y , a few comments should be made about t h e r e l a t i v e u n c e r t a i n t i e s o f t h e performance c h a r a c t e r i s t i c s f o r t h e r e a c t o r designs i n t h i s study. 235U-fueled LWR (PblR) has

low performance u n c e r t a i n t i e s .

C l e a r l y , t h e low-enriched

Numerous PWRs t h a t have been designed

using these methods are c u r r e n t l y i n operation. The h i g h l y enriched 235U-fueled HTGR a l s o would be expected t o be q u i t e accurate since F o r t St. Vrain s t a r t e d up i n 1977. For t h e same

I

reason, t h e successful operation o f HWRs i n Canada gives a high l e v e l o f confidence i n t h e n a t u r a l uranium f u e l e d CANDUs. The Pu-U-fueled FBRs have had a g r e a t deal o f c r i t i c a l experiment backup, and a few FBRs have been b u i l t i n t h e U.S.

and abroad, g i v i n g assurance i n t h e c a l c u l a t e d performance

parameters o f these reactors. Most o f t h e remaining reactors, however, have r a t h e r l a r g e u n c e r t a i n t i e s associated w i t h t h e i r performance c h a r a c t e r i s t i c s . This i s because these r e a c t o r s have n o t been b u i l t , and most have n o t even had c r i t i c a l experiments t o v e r i f y t h e designs. The u n c e r t a i n t y f o r t h e a l t e r n a t e - f u e l e d r e a c t o r designs i s even g r e a t e r since t h e e f f o r t i n developing nuclear data f o r 233U and thorium has been modest compared t o t h a t expended i n developing data f o r 235U,

238U,

and plutonium.


Table 6.1-2.

C h a r a c t e r i s t i c s o f Various Reactors

..-.-. Lifetime Introduction Date

Power Level (We)

LWR-U5(LE)/U-S LUR-U5(LE)/U-EE

1969 1981

LWR-U5(DE)/U/Th LWR-U3(OE)/U/Th

Reactor/Cycl ea

3 8

(tons i3i8/GWe)b

Requirements Enrichment ( I O 6 k g SWU/GWe)C

Charge

Discharge

Net

Charge

1150

5236

1157

4078

3.11

1150

4904

0

4904

3.11

1987

1150

8841

3803

5038

1991

1150

0

0

0

8;03 0

LWR-PU/U

1991

1150

950

0

950

0

LWR-PU/Th

1991

1150

0

0

0

0

SSCR-U5(LE)/U SSCR-U3( OE)/U/Th

1991 1991

1300 1300

4396

908

3489

0

0

0

0

0

SSCR-PU/Th

1991

1300

0

0

0

0

0

.

2.42

Discharge 0.17

0 3.20

Net 2.94

25.8

30

0.60

3.11

18.2

43

0.54

4.83

24.1 24.1

33

0.66

32

0.80

25.7

30

0.70

22.6

33

25.3 23.0

30

0

0

23.0

33

0

114.9

7.5 16 16

0 0 0 0.05

0 0 0 2.37

HWR-U5( NAT)/U

1995

1000

4156

0

4156

HWR-US( SEU)/U

1995

1000

3187

3187

0.59

HWR-US( DE)/U/Th

1995

1000

7337

0 2402

4935

6.66

HWR-U3( DE )/U/Th

1995

1000

HWR-PU/U

1995

1000

2030

1995

1000

0 0 0

0 0

HWR-Pu/Th

0 2030 0

0

0

HTGR-U5 (LE )/U-T

1995

1344

401 7

0

4017

3.23

HTGR-US (LE)/U

1995

1344

401 7

HTGR-U5( DE)/U/Th

1995

1344

3875

431 465

3586 3410

3.23 3.52

0.12 0.30

HTGR-US(HE) /U/Th

1995

1344

3903

558

3345

3.90

0.55

3.35

HTGR-U3( OE)/U/Th

1995

1344

0

0

0

0

0

HTGR-U3/Th

1995

1344

0

0

1995

1344

0

0 0

0

HTGR-PU/Th

0 0 0

0

0

0 0

FBR-PU-U/U

2001

1200

0

0

0

FBR-Pu-U/Th FER-Pu-Th/Th

2001

1200

2001

1200

0 0

2001

1200

0

0 0 0

0 0

FBR-U3-U/Th

0 0 0 0

0 0 0 0

0 0 0 0

0

0

0 0 1.94

-0

'LE = low enriched; DE = denatured; NAT = natural; SEU = s l i g h t l y enriched; HE = highly enriched; U5 = bextended discharge exposure; T = optimized f o r throwaway. l l i t h 1%fabrication and 1% reprocesslng losses; enrichment t a i l s assay 0.2%. "Core/Radi a1 B1 anket/Axi a1 B1 anket.

0.59

53.9

4.73

53.9

235U;

U3 =

33

53.9 53.9

16

53.9

16

3.23

8.2

80

0.50

3.11 3.22

7.2

91 104

0.50 0.54

8.9

74

0.67

10.4

63

0.65

14.0

47

0.86

3.4

196

0.62

12.7/5.1/7.1

62

1.36

12.7f4.6f6.4 11.6f4.6f6.4

62

1.34

68

1.19

12.7/4.6/6.4

63

1.18

0 0 0 0

E q u i l i b r i u m Conditions heavy Metal Core Breedi nq Discharge or Fabrfcation Requirements Exposure Conversion (MWD/kg) Ratio (MT/GWe-yr)

0 0 0-

*33U;

6.3

s

16

standard LWR; EE = LWR w i t h


Table 6.1 -3.

2 3 3 U (kg/GWe-yr) Reactor/Cycle LWR-US(LE)/U-S LWR-U5(LE)/U-EE LWR-U5( OE)/U/Th LWR-U3(OE)/U/Th LWR-PU/U LWR-PU/Th SSCR-US( LE)/U SSCR-U3( OE)/U/Th SSCR-PU/Th HWR-U5( NAT)/U HWR-U5( SEU)/U HWR-U5( OE)/U/Th HWR-U3( DE )/U/Th

Charge

0 0 0 807.0 0

0 0

619.9 0

0 0 0 765.8

Pu (kg/GWe-yr)

2 3 5 U (kg/GWe-yr)

Discharge

Net

0 0 256.2 530.4

276.6

0 239.0

Average F i s s i l e Mass Flows* f o r Various Reactors

Charge

Discharge

Net

Charge

0

736.9

213.4

523.5

0 -256.2

683.3 1169.7

0 507.9

683.3

13.5

16.8

-3.3

0 0 0 0

173.1 0

91.2

82.0

2.3

-2.3

626.6

169.3

457.3

31.2

-4.4

4.3

-4.3

0

-239.0

0

0

426.2

193.7

281.2

-281.2

26.8 0

Discharge 146.8

T o t a l (kg/GWe-yr) Net -146.8

Charge

Discharge

736.9

360.2

376.7

0 77.8

0 -77.8

1169.7

841.9

327.8

88.2

-88.2

820.5

635.4

185.1

472.2 620.2

228.5 673.9

873.7 1294.1

563.4 861.5

310.5 432.6

0

185.0

-185.0

626.6

354.3

272.3

0 1202.3

72.9

-72.9

646.7

530.3

116.4

556.4

645.9

1202.3

841.9

360.4

290.4

-290.4

757.4

518.2

239.2

159.8

-159.8

521.8

232.0

289.9

22.5

-22.5

970.8

763.5

207.3

26.9

-26.9

799.4

728.6

70.8

661.8

700.6 1294.1

683.3

0

Net

683.3

0

0

757.4

227.8

529.6

0

0

521.8

72.2

449.7

418.2

-41 8.2

322.8

664.7

101.1

970.8 33.6

37.0

648.0 -3.4

0

369.9

67.2

302.7

156.6

177.7

-21.1

526.5

244.9

281.6

0

2.8

-2.8

895.5

234.4

661.2

895.5

629.1

266.4

0 -43.1

540.1

43.1

540.1

0 112.2

427.9

27.3

-27.3

689.0

161 .O

528.0

512.3 424.2

261.2

251.1

157.3

266.9

575.3

458.9

116.4

510.3

637.0

223.7

413.3

HWR-PU/U

0

HWR-PU/Th

0

HTGR-U5 (LE)/U-T HTGR-U5 (LE)/U

0 0

HTGR-U5( DE )/U/Th

0

68.9

HTGR-U5 (HE)/Th

0

0 391.9

-391.9

0 0

0

540.1 540.1

69.1

540.1 471 .O

-68.9

689.0

64.8

624.2

-186.9 302.5

512.3 13.2

73.3 21 .o

439.0 -7.7

73.8

69.9

3.9

2.9

-2.9

0 0

HTGR-U3( OE)/U/Th

411.0

186.9 108.4

HTGR-U3/Th

501.5

389.0

112.5 -94.1

0

0 0 0 0

0 0 0 0 0 0

0

1.o 27.9

0

-1 .o -27.9

0

540.1

HTGR-PU/Th

0

94.1

FBR-Pu-U/U

0 0 0

69.7

48.1

21.6

1253

1526

-273.3

1322.7

1574.1

-251.7

237.5

-237.5

31.8

14.0

1261

1283

-21.9

1292.8

1538.3

-245.4

743.2

-743.2

853.7

630.7

1484

1596.9

-112.9

844.5

368.0

0 33.3

17.8 0 19.4

13.9

499.8

-499.8

1245.8

1363.7

-117.9

FER-PU-U/Th FER-Pu-Th/Th FBR-U3-U/Th

1212.5

0

0

* L i f e t i m e average w i t h 1% f a b r i c a t i o n and 1%reprocessing losses.

0

637.0

1484

0

126.7


6-10

6.1.3.

Nuclear P o l i c y Options,

L

Under t h e assumption t h a t t h e r e a c t o r / f u e l cycles l i s t e d i n Tables 6.1-2 and 6.1-3 could be deployed, a s e t o f nuclear p o l i c y options were developed f o r studying t h e r e l a t i v e c a p a b i l i t i e s o f t h e various r e a c t o r s t o produce c i v i l i a n nuclear power d u r i n g t h e p e r i o d

As was pointed o u t above, i t was assumed t h a t f o r a system t o be adequate, i t should have an i n s t a l l e d nuclear capacity o f 350 GWe by t h e y e a r 2000 and a from 1980 t o 2050.

a;

order t o determine t h e e f f e c t o f a lower growth rate, a few cases were a l s o r u n f o r an

L

i n s t a l l e d capacity o f 200 GWe i n the year 2000 and 10 GWe/yr t h e r e a f t e r . )

1

n e t increase o f 15 GWe t h e r e a f t e r , w i t h each p l a n t having a 30-yr l i f e t i m e .

(Note:

In

I t was a l s o

-,

L4

assumed t h a t r e a c t o r s f u e l e d w i t h natural, low-enriched, s l i g h t l y enriched, o r denatured uranium could be dispersed outside t h e secure energy centers and those f u e l e d w i t h h i g h l y

I

A l l enrichment, reprocessing, and f a b r i c a t i n g f a c i l i t i e s would a l s o be confined w i t h i n t h e centers.

enriched uranium o r w i t h plutonium would be confined w i t h i n t h e centers.

The nuclear p o l i c y options f e l l under four'major categories:

(1) t h e throwaway/

stowaway option; (2) c l a s s i c a l plutonium-uranium options; ( 3 ) denatured uranium options employing thermal converters only; and' (4) denatured uranium options employing both converters and breeders.

The various options under these categories are described i n Table 6.1-4, and Schematic repret h e s p e c i f i c r e a c t o r s u t i l i z e d i n each o p t i o n are i n d i c a t e d i n Table 6.1-5. sentations o f t h e options are presented i n Figs. 6.1-1 through 6.1-4.

Runs were made f o r

both intermediate-cost and high-cost y308 supply assumptions. These nuclear options cannot be viewed as p r e d i c t i o n s o f t h e f u t u r e i n s o f a r as nuclear power i s concerned; however, they can provide a l o g i c framework by which t h e f u t u r e implicat i o n o f c u r r e n t nuclear p o l i c y decisions can be understood.

Suppose, f o r example, a group

o f natSbns agree t o supply nuclear f u e l t o another group o f nations p r o v i d i n g t h e l a t t e r agree t o forego reprocessing.

A careful a n a l y s i s o f t h e nuclear system options o u t l i n e d

I;

L 6' I;

fd:

above can i l l u s t r a t e t h e l o g i c a l consequences o f such a d e c i s i o n upon t h e c i v i l i a n nuclear power systems i n both groups of nations.

Only those nations p r o v i d i n g t h e fuel would main-

t a i n secure energy centers, since t h e nations r e c e i v i n g t h e f u e l would be operating dispersed reactors only.

(Note:

The analysis presented here considers o n l y t h e U.S.

ore supply.

A

s i m i l a r analysis f o r a group of nations would begin w i t h d i f f e r e n t assumptions regarding t h e

L I]

ore supply and nuclear energy demand.) For t h e purposes o f t h i s analysis, a l l t h e nuclear system options were assumed t o be mutually exclusive.

That i s , i t was assumed t h a t any o p t i o n selected would be pursued t o

L

I n a c t u a l i t y , a n a t i o n would have t h e a b i l i t y t o change p o l i c i e s i f consequences o f t h e p o l i c y i n e f f e c t were determined t o be undesirable. However, t h e a b i l i t y i t s u l t i m a t e end.

t o s u c c e s s f u l l y change a p o l i c y a t a f u t u r e date would be q u i t e l i m i t e d i f t h e necessity o f changing has n o t been i d e n t i f i e d and incorporated i n t o t h e c u r r e n t program.

The purpose

o f t h e study contained i n t h i s r e p o r t was t o i d e n t i f y t h e basic nuclear system options, and t o determine t h e consequences o f pursuing them t o t h e i r u l t i m a t e end. (Note: A study of t h e consequences o f changing p o l i c i e s a t a f u t u r e date programs

- will

be analyzed i n a l a t e r study.)

- and thereby t h e

implication o f current

i:


t

he 6-1 1

6.1.4.

Iw

The A n a l y t i c a l Method

The p r i n c i p a l components o f the a n a l y t i c a l method used i n t h i s study are i l l u s t r a t e d i n Fig. 6.1-5 and a r e based on t h e f o l l o w i n g assumptions:

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I

1

L b

L

(1) Given a s p e c i f i e d demand f o r nuclear energy as a f u n c t i o n of time, nuclear u n i t s are constructed t o meet t h i s demand consistent w i t h t h e nuclear p o l i c y o p t i o n under cons i d e r a tion. (2) As nuclear u n i t s r e q u i r i n g U308 are constructed, t h e supply of U3O8 i s continuously depleted. The d e p l e t i o n r a t e i s based on both the f i r s t core load and t h e annual reloads r e q u i r e d throughout the l i f e o f t h e nuclear u n i t . The long-run marginal c o s t o f U308 i s

assumed t o be an i n c r e a s i n g f u n c t i o n o f t h e cumulative amount mined.

This i s i n d i c a t i v e o f

a continuous t r a n s i t i o n from higher grade t o lower grade resources.

( 3 ) I f t h e nuclear p o l i c y o p t i o n under consideration assumes reprocessing, t h e f u e l i s stored a f t e r discharge u n t i l reprocessing i s a v a i l a b l e . A f t e r reprocessing, t h e f i s s i l e plutonium and 233U a r e a v a i l a b l e f o r r e f a b r i c a t i o n and reloading.

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L I

b

( 4 ) A nuclear u n i t which requires 239Pu o r 233U cannot be constructed unless t h e supply o f f i s s i l e m a t e r i a l i s s u f f i c i e n t t o provide t h e f i r s t core load plus t h e reloads on an annual basis throughout t h e u n i t ' s l i f e . (5) The number o f nuclear u n i t s s p e c i f i e d f o r operation through t h e 1980's i s exogenously c o n s i s t e n t w i t h t h e c u r r e n t c o n s t r u c t i o n plans o f u t i 1 i t i e s . (6) A nuclear p l a n t design which d i f f e r s from established technology can be i n t r o duced o n l y a t a l i m i t e d maximum r a t e . A t y p i c a l maximum i n t r o d u c t i o n r a t e i s one p l a n t d u r i n g t h e f i r s t biennium, two p l a n t s d u r i n g t h e second biennium, f o u r during t h e t h i r d , eight during the fourth, etc.

li

(7) I f t h e manufacturing c a p a b i l i t y t o produce a p a r t i c u l a r r e a c t o r type i s w e l l established, t h e r a t e a t which t h i s r e a c t o r type w i l l l o s e i t s share o f t h e new c o n s t r u c t i o n market i s l i m i t e d t o a s p e c i f i e d f r a c t i o n per year.

A t y p i c a l maximum c o n s t r u c t i o n market

l o s s r a t e i s lO%/yr. This r e f l e c t s t h e f a c t t h a t some u t i l i t i e s w i l l continue t o purchase p l a n t s o f an e s t a b l i s h e d and r e l i a b l e technology, even though a new technology may o f f e r an improvement.

i t '

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The a c q u i s i t i o n o f f i s s i l e m a t e r i a l w i l l be t h e p r i n c i p a l goal o f any n a t i o n embarked upon a nuclear weapons program.

Therefore, any a n a l y s i s o f a d i v e r s i o n - r e s i s t a n t c i v i l i a n

nuclear power s t r a t e g y must i n c l u d e a d e t a i l e d a n a l y s i s o f t h e nuclear f u e l cycle.

The

steps i n t h e nuclear f u e l c y c l e which were e x p l i c i t l y modeled i n t h i s a n a l y s i s a r e shown i n Fig. 6.1-6.

They include:

the mining o f U308; the conversion o f U 3 0 8 t o

UF,; t h e

enrichment o f t h e uranium by e i t h e r the gaseous d i f f u s i o n technique o r t h e c e n t r i f u g e


Table 6.1-4.

Nuclear Policy Options'

Throwaway/Stowaway Option (see Fig. 6.1-1) b Option 1: LEU (235U/238U) converters operating on the throwaway/stowaway cycle are permitted outside the energy centers and no react o r s are operated inside the centers. Spent fuel is returned t o the secure energy centers f o r ultimate disposal.

Plutonium-Uranium Options (see Fig. 6.1.2) Option 2: LEU (235U/238U) converters are operated outside the secure energy centers and Pu/U converters and 295U(HE)Th, 233U/Th, and Pu/Th HTGR's are permitted inside the centers. Uranium is recycled in a l l reactors, and plutonium i s recycled i n energy-center reactors. Option 3: LEU (235U/238Uj converters a r e operated outside the secure energy centers and Pu/U converters, Pu-U/U breeders, and 235U(HE)/Th, 233U/Th, and Pu/Th HTGRs are permitted inside the centers. Uranium i s recycled i n a l l the reactors, and plutonium i s recycled i n the energy-center reactors. Denatured Uranium Options w i t h Converters Only (see Fig. 6.1-3) converters and denatured 235U and 233U converters are operated outside the energy centers and no reactors are Option 4: LEU (235U/23eU) operated inside the centers. The f i s s i l e uranium i s recycled into the converters, but the plutonium i s stored inside the centers e i t h e r f o r . ultimate disposal or f o r future use a t an unspecified date. converters and denatured 235U and 233U converters are operated outside the energy centers and Pu/Th conoption 5U: LEU (235U/238U) verters are permitted inside the centers. The fissile uranium i s recycled into the outslde reactors and the plutonium into the inside reactors. The goal i n t h i s case i s t o minimize the mount of pZutonium produced and t o "transmute" a l l t h a t i s produced into 233U i n the energycenter reactors. Option 5T: LEU (235U/238U) converters and denatured 233U converters are operated outside the energy centers and Pu/Th converters are permitted inside the centers. The f i s s i l e uranium is recycled into the outside reactors and the plutonium into the inside reactors. The goal i n this case is not t o minimize the mount of plutonium produced but "transmute" a l l that is produced t o 233U i n the energy-center reactors. Denatured Uranium Options w i t h Converters and Breeders (see Fig, 6.1-4) Option 6: LEU (233U/238U)converters and denatured 235U and 233U converters are operated outside the energy centers and Pu/Th converters and Pu-U/Th breeders (Pu-U cores, Th blankets) are permitted inside the centers. The f i s s i l e uranium is recycled into the outside reactors and the inside breeders and plutonium is recycled into the inside converters and breeders. With the reactors used, only a light '%-to-233U" transmutation r a t e is realized. Option 7: LEU (235U/238U) converters, denatured 235U and 233U converters. and denatured 23311 breeders are operated outside the energy centers and Pu/Th converters and Pu-U/Th breeders (Pu-U cores, Th blankets) are permitted inside the centers. The f i s s i l e uranium i s recycled i n t o the outside reactors and the inside breeders and plutonium is recycled i n the inside converters and breeders. With the reactors trunsmtution r a t e is realized. T h i s case represents the f i r s t time a denatured breeder is introduced i n used, only a l i g h t % - t 0 - ~ 3 3 U " the system. converters, denatured 2351) and 233U converters, and denatured 233U breeders are operated outside the energy Option 8: LEU (235U/238U) Centers and Pu/Th converters and Pu-Th/Th breeders (Pu-Th cores, Th blankets) are permitted inside the centers. The f i s s i l e uranium i s U~' recycled into the outside reactors and the plutonium into the inside reactors. With the reactors used, a heavy " F U - ~ O - ~ ~ ~transmutation r a t e is realized. Again a denatured breeder is u t i l i z e d i n the system. 'In a l l options except Option 1 , spent fuel is returned t o the secure energy centers for reprocessing. returned t o the center f o r ultimate disposal. bHWRs t h a t are fueled w i t h natural or s l i g h t l y enriched uranium are included l n this category.

For Option 1 , the Spent fuel i s


Table 6.1-5.

FBR-Pu-U/U FBR-PU-U/Th FBR-Pu-Th/Th FBR-UW/Th

-

Reactors A v a i l a b l e i n Secure (S) Centers o r Dispersed (0) Areas f o r Various Nuclear P o l i c y Options

- - - -

_ _ _ _

- - - -

- - - -

-

- - - -

- - - - - - a

_

-

-

- - - - - - - _ _ - _ - _ _ - _ _ _ - - - - - - - - - - - s s s s

- - - -

- - - -

- - - -

- - - -

_ _ _ _

- - - -

_ _ _ _

s s s s

s s s s

- - - -

D D D D

- - - - - - - -

_ _ _ _

--

- _ _ _ - - - s s s s D D D D

*LE low enriched; DE denatured; NAT = natural; SEU = s l i g h t l y enriched; HE h i g h l y enriched; US = 235U; U3 = 233U; S = standard LWR; EE LWR w i t h extended discharge exposure; T = optimized f o r throwaway. L. 5 . H. and G i n d i c a t e type o f converter employed i n option, where L = LWR. S SSCR. H HWR, and G = HTGR.


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7

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THROWAWAY

SWU

U5ILEIN-

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L L

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L L I'

HEDL 7801-78.7

Option 1: I n t h i s o p t i o n , LEU (235U/238U) converters* o p e r a t i n g on t h e throwaway/ stowaway c y c l e a r e p e r m i t t e d o u t s i d e t h e energy centers and no r e a c t o r s a r e o p e r a t e d i n s i d e t h e centers. Spent f u e l i s r e t u r n e d t o t h e secure energy centers f o r u l t i m a t e disposal

.

Fig. 6.1-1.

Option 1:

The Throwaway/Stowaway Option.

technique; t h e f a b r i c a t i o n o f 235U, 233U, and 239Pu f u e l s ; t h e d e s t r u c t i o n and transmutation of f i s s i l e and f e r t i l e isotopes o c c u r r i n g d u r i n g power production i n t h e r e a c t o r ; t h e storage of spent f u e l , and, i f permitted, t h e reprocessing o f spent f u e l ; t h e s i z e and composition of f i s s i l e s t o c k p i l e s as a f u n c t i o n o f time; and t h e amount o f spent f u e l o r h i g h - l e v e l waste which must be stored as a f u n c t i o n o f time.

Thus, t h e amount, composition, and move-

ment o f a l l f i s s i l e m a t e r i a l i n t h e c i v i l i a n nuclear power system were a c c u r a t e l y c a l c u l a t e d f o r each case under t h e nuclear p o l i c y options shown i n Tables 6.1-4 and 6.1-5The c o s t o f each nuclear o p t i o n and t h e t o t a l power c0s.t o f each nuclear u n i t i n t h e o p t i o n were a l s o calculated; however, t h e t o t a l power c o s t o f a nuclear u n i t d i d n o t determine whether i t would be constructed. Generally i t was constructed i f (1) i t was a v a i l a b l e i n t h e p o l i c y under consideration, and (2) i t had a lower U308 consumption r a t e than t h e o t h e r nuclear u n i t s a v a i l a b l e under t h e same p o l i c y option.

This approach

was adopted because i t i s p o s s i b l e t o c a l c u l a t e t h e U308,f i s s i l e plutonium, and 233U requirements o f a nuclear u n i t w i t h reasonable accuracy, w h i l e i t i s very d i f f i c u l t t o

"WRs

t h a t a r e f u e l e d w i t h n a t u r a l o r s l i g h t l y enriched uranium a r e included i n t h i s category.

c L L b L

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6-15

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L L

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UEDL 7801-78.6

I n t h i s option, LEU (235U/238U) converters are operated outside the Option 2: secure energy centers and Pu/U converters and 235U(HE)/Th, 233U/Th, and Pu/Th HTGRs are Spent f u e l i s returned t o the centers f o r reprocessing. permitted i n s i d e the centers. Uranium i s recycled i n a l l -..actors, and plutonium i s recycled i n energy-center reactors. (Note: Sketch does n o t f. cover Option 26; see Table 6.1-5.)

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UEDL7801.7&5

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Option 3: I n t h i s option, LEU (235U/238U) converters are o erated outside the secure energy centers and Pu/U converters, Pu-U/U breeders, and 23 U(HE)/Th, 233U/Th, Spent f u e l i s returned t o the centers and Pu/Th HTGRs are permitted i n s i d e the centers. Uranium i s recycled i n a l l the reactors, and plutonium i s recycled f o r reprocessing. (Note: Sketch does not f u l l y cover Option 36; see i n the energy-center reactors. Table 6.1-5.)

Fig. 6.1-2.

Options 2 and 3:

The Plutonium-Uranium Options.


6-16

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nM

MEDL nows.1

Option 4: I n t h i s o p t i o n , LEU (235U/238U) converters and denatured 235U and 2 3 3 u converters a r e operated o u t s i d e t h e energy centers and no r e a c t o r s a r e o p e r a t e d i n s de the centers. Spent f u e l i s returned t o t h e secure energy centers f o r reprocessing. The f i s s i l e uranium i s recycled i n t o t h e converters, but the plutonium i s s t o r e d i n s i d e t h e c e n t e r e i t h e r f o r u l t i m a t e disposal o r f o r f u t u r e use a t an u n s p e c i f i e d date.

Fig. 6.1-3.

Options 4, 5U, and 5T:

Denatured Uranium Options w i t h Converters On

c a l c u l a t e t h e c a p i t a l , f a b r i c a t i o n , and reprocessing costs f o r t h e same u n i t . (Note:

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An

exception t o t h i s philosophy was contained i n a s e t o f cases described i n Appendix D i n which t h e U308 supply was assumed t o be s u f f i c i e n t l y l a r g e so as n o t t o impose a p r a c t i c a l l i m i t on the growth o f t h e nuclear system over t h e planning horizon.

L

f;

I n t h i s case, t h e

decision t o construct--or n o t t o construct--a r e a c t o r concept was based on i t s t o t a l power cost, which o f course included t h e c o s t o f U308 as an increasing f u n c t i o n o f t h e t o t a l amount consumed. Thus, w h i l e t h e a b i l i t y t o conserve U308 d i d e n t e r i n t o t h e decision, i t was n o t t h e s i n g l e dominating f a c t o r . ) An example o f t h e u n c e r t a i n t y involved i n c a l c u l a t i n g t h e t o t a l power c o s t o f a nuclear u n i t i n t h e f u t u r e i s i l l u s t r a t e d i n Fig. 6.1-7.

T h i s f i g u r e was developed by

assigning a reasonable s e t o f u n c e r t a i n t i e s t o t h e c a p i t a l , f a b r i c a t i o n , and reprocessing costs f o r a s e t o f f i v e r e a c t o r concepts w i t h f o u r f u e l options f o r each concept. The

I n a l l cases, t h e costs were assumed t o be mature i n d u s t r y costs d u r i n g t h e p e r i o d 2010 t o 2040 w i t h

actual costs and t h e i r u n c e r t a i n t y a r e discussed i n d e t a i l i n Appendix 6. t h e p r i c e o f U308 increasing from $140/lb t o $180/lb during t h i s period.

The r e a c t o r

concepts shown in t h e f i g u r e are t h e LWR, SSCR, HWR, and HTGR converters and t h e FBR.

u


6-1 7

L u p 8 4 ENRICH~ ...

HM

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I

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HM

HEDL 7W1.782

Option 5U: I n t h i s option, LEU (235U/238U) converters and denatured 235U and 233U converters are operated outside the energy centers and Pu/Th converters are permitted inside the centers. Spent f u e l i s returned t o the secure energy centers f o r reprocessing, The f i s s i l e uranium i s recycled i n t o the outside reactors and the plutonium i n t o the inside reactors. The goal i n t h i s case i s t o minimize the amount of plutonium produced and t o "transmute" a l l t h a t i s produced i n t o 233U i n the energy-center reactors.

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6

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HEOL mo1.m.z

Option 5T: i n t h i s option, LEU (235U/238U) converters and denatured 233U converters are operated outside the energy centers and Pu/Th converters are permitted i n s i d e the centers. Spent f u e l i s returned t o the secure energy centers f o r reprocessing. The f i s s i l e uranium i s recycled i n t o the outside reactors and the plutonium i n t o the i n s i d e reactors. The goal i n t h i s case i s not t o minimize the amount of plutonium produced but t o "transmute" all that is produced t o 233U i n the energy-center reactors.


L

6-18

L

I;

L HEDL7OOl.70.3

HM

I n t h i s option, LEU (235U/238U) converters and denatured 235U and 233U Option 6: converters are operated outside the energy centers and Pu/Th converters and Pu-U/Th breeders (Pu-U cores, Th blankets) are permitted i n s i d e the centers. Spent f u e l i s returned t o the secure energy centers f o r reprocessing. The f i s s i l e uranium i s recycled i n t o the outside reactors and the inside breeders, and the plutonium i s recycled i n t o the i n s i d e converters and breeders. With the reactors used, o n l y a Zight " P U - ~ U - ~ ~ ~ U ~ ' tmsmutation rate i s r e a l ized.

Fig. 6.1-4.

Options 6, 7, and 8:

Denatured Uranium Options w i t h Converters and Breeders.

c

The f u e l c y c l e options assumed f o r t h e converters are as follows: (1) Low-enriched 235U/238U f u e l , r e a c t o r operating on throwaway cycle;

(2) Low-enriched

235U/238U

f u e l , reprocessing and

235U

r e c y c l e permitted;

( 3 ) Pu/U f u e l , reprocessing and Pu and 235U r e c y c l e permitted (LWRs o n l y ) ; (4) Pu/Th f u e l , reprocessing and Pu and z33U r e c y c l e permitted;

(5) Denatured 233U/238U/Th f u e l , reprocessing and 233U and Pu r e c y c l e permitted. For t h e case o f t h e FBR, t h e f u e l options are

( I ) Pu/U f u e l i n core, Th i n blankets, reprocessing and Pu and 233U r e c y c l e permitted; (2) Pu/Th f u e l i n core, Th i n t h e blankets, reprocessing and Pu and 233U r e c y c l e permitted.

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6-1 9

b

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Option 7: I n t h i s option, LEU (235U/238U) converters, denatured 235U and 2 3 3 U converters, and denatured 2 3 3 U breeders are operated outside the energy centers and Pu/Th converters and Pu-U/Th breeders (Pu-U cores, Th blankets) are permitted i n s i d e the centers. Spent f u e l i s returned t o the secure energy centers f o r reprocessing. The f i s s i l e uranium i s recycled i n t o the outside reactors and the inside breeders, and the plutonium i s recycled i n the i n s i d e converters and breeders. With the reactors used, only a l i g h t rrPu-to-233Urrtransmutation rate i s realized. This case represents the f i r s t time a denatured breeder is introduced i n the system.

n

Option 8: I n t h i s option, LEU (235U/238U) converters, denatured 235U and 233U converters, and denatur.ed 2 3 3 U breeders a r e operated outside the energy centers and Pu/Th converters and Pu-Th/Th breeders (Pu-Th cores, Th blankets) are permitted i n s i d e the centers. Spent f u e l i s returned t o the secure energy centers f o r reprocessing. The f i s s i l e uranium i s recycled i n t o the outside reactors and the plutonium i n t o t h e inside reactors. With the reactors used, a heavy " ~ J I A - ~ O - ~transmutation ~~U~~ rate i s realized. Again, a denaturedbreeder i s u t i l i z e d i n the system.


L

6-20 c-

1 yJ* SUPPLY

CONVLRSION

I SRCIFIED CONSTRUCTION

r7'\ U-233 SRCIFIEO MARE1

LlMllLD INTRODUCTION m c m R DESIGNS

OBSOLLSCENT MSIGNS WtDL m - W 3

Fig. 6.1-5. Model f o r Nuclear Systems Assessment Study.

As Fig. 6.1-7

Fig. 6.1-6.

.Nuclear Fuel-Cycle Model.

i l l u s t r a t e s , t h e t o t a l l e v e l i z e d power c o s t o f a r e a c t o r concept i n s o f a r I n partic-

as an intercomparison o f concepts i s concerned i s dominated by t h e u n c e r t a i n t i e s .

u l a r , t h e t o t a l power costs f o r those concepts possessing t h e g r e a t e s t resource saving ( t h e HWR and t h e FBR) e x h i b i t t h e g r e a t e s t u n c e r t a i n t i e s . The e f f e c t o f t h e p r i c e o f U308 i s a l s o Figure 6.1-7 shows t h a t t h e t o t a l power c o s t o f t h e LWR on t h e throwaway c y c l e significant.

L

c L c L

i s s i g n i f i c a n t l y lower i f t h e p r i c e o f U308 i n t h e year o f s t a r t u p i s $40/lb r a t h e r than $140/ 1b. .

The l e v e l i z e d power costs given f o r each r e a c t o r system i n Fig. 6.1-7 were determined

from t h e sum o f t h e discounted values o f t h e cash flows associated w i t h t h e system d i v i d e d by t h e discounted e l e c t r i c a l energy production. The cash flows considered were: (1) c a p i t a l investment, i n c l u d i n g t h e r e t u r n of t h e investment and t h e r e t u r n on t h e investment; (2) f i x e d charges, such as c a p i t a l replacements, nuclear l i a b i l i t y insurance, etc.; ( 3 ) operat i o n and maintenance costs; (4) income taxes; and (5) f u e l expenses. The f i r s t four items are r e l a t i v e l y straightforward, w i t h t h e r e l e v a n t data given i n Appendix B. The f i f t h item, however, m e r i t s some a d d i t i o n a l discussion, p a r t i c u l a r l y as f u e l expenses r e l a t e t o t h e v a l u a t i o n o f t h e bred f i s s i l e material. For these c a l c u l a t i o n s t h e c o s t of bred f i s s i l e m a t e r i a l was taken t o be t h e "shadow price," which i s t h e value o f an a d d i t i o n a l u n i t o f f i s s i l e m a t e r i a l t o t h e p a r t i c u l a r scenario i n question.

The shadow p r i c e c a l c u l a t e d f o r t h e bred f i s s i l e m a t e r i a l i s d i r e c t l y r e l a t e d t o t h e U308 p r i c e s a t and subsequent t o t h e v a l u a t i o n p o i n t i n time. The value o f t h e bred f i s s i l e m a t e r i a l thus increases w i t h increasing U30, p r i c e which i n t u r n increases as a func-

t i o n o f t h e cumulative q u a n t i t y consumed.

For t h e resource-limited scenarios, an a d d i t i o n a l

u n i t o f 233U o r Pu w i l l postpone t h e purchase o f an equivalent amount o f U308, t h e delay having a d o l l a r value due t o t h e use o f discounted cash flows. For those scenarios which a r e n o t resource-limited,

an a d d i t i o n a l u n i t o f bred f i s s i l e m a t e r i a l permits t h e elimina-

t i o n o f an equivalent amount o f U308.

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6-21

TIME FRAME: 2010 TO 2040 U308 PRICE: I40 $p I N YEAR OF STARTUP INCRWING TO 180 S/I,AT THE END OF THE PLANT LIFE

U 3 9 PRICE Gf 40 $/I I N YEAR Of STARTUP REACTOR OPTIONS:

LWR

SSCR

Since t h e v a l u a t i o n o f t h e bred f i s s i l e m a t e r i a l i s r e l a t e d t o the cumulative U308 p r i c e structure, t h e r a t e a t which t h e U308 i s consumed during a p a r t i c u l a r scenario a l s o a f f e c t s t h e time-dependent p r i c e c a l c u l a t e d f o r t h e bred f i s s i l e m a t e r i a l . Rapid consumpt i o n o f t h e resource base (i.e.,

a high energy demand) y i e l d s a r a p i d l y r i s i n g shadow

p r i c e . Such an e f f e c t i s r e a d i l y noticeable i n the c a l c u l a t i o n o f t h e power costs of breeder r e a c t o r s since i t i s possible f o r t h e c r e d i t c a l c u l a t e d f o r t h e bred m a t e r i a l t o exceed t h e p e r i o d ' s charges f o r t h e r e a c t o r ' s inventory.

Thus, t h e n e t f u e l expense f o r c e r t a i n systems producing h i g h l y valued f i s s i l e m a t e r i a l can be negative, r e s u l t i n g i n s i g n i f i c a n t power cost d i f f e r e n c e s when compared t o t h e r e a c t o r systems operating w i t h high-cost n a t u r a l resources. This type o f phenomenon i s i l l u s t r a t e d schematically by Fig. 6.1-8 i n which t h e power costs o f a f a s t breeder and o f an LEU-LWR are p l o t t e d as a f u n c t i o n o f U308 price. The r i s i n g power c o s t o f t h e LWR i s d i r e c t l y a t t r i b u t a b l e t o t h e increasing f u e l expense caused by t h e U308 price.

The d e c l i n i n g f a s t r e a c t o r power c o s t

r e f l e c t s t h e increasing value o f (and hence l a r g e r c r e d i t f o r ) t h e bred m a t e r i a l when compared t o U308-derived f i s s i l e material. The s i t u a t i o n i s s t i l l complicated even i f one considers o n l y t h e conceptually simple case o f t h e throwaway cycle. From Fig. 6.1-9, where f o r s i m p l i c i t y t h e p r i c e of U308 was assumed t o be constant over t h e l i f e o f t h e p l a n t , i t appears t h a t t h e LWR i s t h e l e a s t expensive r e a c t o r when t h e U308 p r i c e i s fess than $60/lb, and t h a t t h e HWR

w i l l be l e s s expensive than t h e LWR when t h e U308 p r i c e i s g r e a t e r than $160/lb. However, an examination o f t h e u n c e r t a i n t i e s leads one again t o t h e conclusion t h a t t h e y dominate t h e problem, and t h a t conclusions based on economic arguments are tenuous a t best.

Thus,

t h e d e c i s i o n was made t o construct o r n o t c o n s t r u c t a nuclear u n i t on t h e basis o f i t s a b i l i t y t o extend t h e U308 supply r a t h e r than on i t s r e l a t i v e cost.


6-22 ' ! 50-

40

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350

u308 Price. $ / l b

Fig. 6.1-8.

Influence of U3O8 Prices on Total Power Costs.

c ONCE-THROUGH OPTION

3

u30e PRICE, SAB HEDL 7805-090.39

Fig. 6.1-9. Total Power Cost o f Various Reactor Systems as a Function of U308 Price (Constant U30s Price with Time; Once-Through Option).


6-23

6.2.

DISCUSSION OF RESULTS FOR SELECTED NUCLEAR POLICY OPTIONS

This s e c t i o n discusses r e s u l t s obtained i n t h i s study f o r a selected s e t o f nuclear system options t h a t t y p i f y t h e r o l e o f nuclear power under d i f f e r e n t nuclear p o l i c y decisions. The i n t e n t i s t o i d e n t i f y the basic issues, t o determine t h e l o g i c a l consequences o f decisions made i n accordance w i t h those issues, and t o d i s p l a y t h e consequences i n an i l l u s t r a t i v e manner.

D e t a i l e d r e s u l t s f o r a l l the nuclear system options o u t l i n e d i n

Section 6.1 are presented i n Appendix C. 6.2.1.

The Throwaway/Stowaway Option

The throwaway/stowaway c y c l e (see Fig. 6.1-1) i s a conceptually simple nuclear system o p t i o n and t h e r e f o r e has been selected as the reference c y c l e against which a l l o t h e r opt i o n s are compared.

__

t

b

I n order t o thorough-

l y understand t h e i m p l i c a t i o n s o f t h e throw-

away cycle, the e f f e c t o f several deployment Mxx)

options u t i l i z i n g t h e various advanced conv e r t e r s on t h e throwaway c y c l e was analyzed

-1 -

Avg. Capacity Factor = 0.67 T a i l s Ccmmosition 0.W20

-

i n detail.

I n general, t h e analysis assumed

a nuclear growth r a t e o f 350 GWe i n t h e year 2000 followed by a n e t increase o f 15 GWelyr, b u t t h e consequences o f a s i g n i f i c a n t reduct i o n i n t h e nuclear growth r a t e were a l s o

r n

$

considered.

m

t h e high-cost and t h e intermediate-cost

I n a d d i t i o n , t h e e f f e c t o f both

U308 supplies was determined.

HTGR

I ?

Fig. 6.2-1. L i f e t i m e U308 Requirements f o r Various Reactors on t h e Throwaway Cycle.

A summary o f t h e 30-yr U308 requirements f o r several reactors on t h e throwaway cycle, i n c l u d i n g an LWR w i t h a f u e l system designed f o r an extended discharge exposure, i s shown i n Fig. 6.2.1.

li

I n each case, t h e

average capacity f a c t o r o f t h e r e a c t o r was assumed t o be 0.67, and t h e t a i l s composit i o n of t h e enrichment p l a n t was assumed t o be 0.0020.

As t h e f i g u r e indicates, a l l t h e

reactors have lower U308 requirements than t h e standard LWR, t h e extended-discharge LWR being 6% lower, t h e SSCR 16% lower, t h e HTGR 23% lower, and t h e s l i g h t l y enriched HWR 39% lower. These U308 requirements were c a l c u l a t e d f o r e s s e n t i a l l y standard designs w i t h o u t elaborate. design optimization.

It i s recognized t h a t design o p t i m i z a t i o n could improve t h e r e a c t o r

performance c h a r a c t e r i s t i c s ; however, t h e goal o f t h i s analysis was n o t t o d e l i n e a t e t h e u l t i m a t e r o l e o f any p a r t i c u l a r r e a c t o r concept based on c u r r e n t performance c h a r a c t e r i s t i c s , b u t r a t h e r t o i d e n t i f y t h e probable r o l e o f each r e a c t o r concept and t h e i n c e n t i v e f o r improving i t s performance c h a r a c t e r i s t i c s .


6-24

The p o t e n t i a l nuclear c o n t r i b u t i o n w i t h LWRs on t h e throwaway cycle, both w i t h and w i t h o u t a f u e l system designed f o r extended exposure' being included, i s shown i n Fig. 6.2-2 f o r t h e high-cost U308 supply. The nuclear c o n t r i b u t i o n passes through a maximum o f approximately 420 GWe i n s t a l l e d capacity i n about 2010 and declines continuously t h e r e a f t e r , t h e system w i t h t h e LWR-EE p r o v i d i n g a s l i g h t l y greater capacity over most o f t h e period.* cumulative capacity constructed throughout t h e planning horizon i s approximately 600 GWe.

The The

maximum i n s t a l l e d capacity i s l e s s than t h e cumulative capacity because new u n i t s must be cons t r u c t e d t o replace those r e t i r e d during t h e period.

The maximum annual U308 requirement i s

72,000 ST/yr and t h e maximum annual enrichment requirement i s 45 m i l l i o n SWU/yr, which can be regarded as excessive.

neither o f

Thus, t h e p r i n c i p a l l i m i t a t i o n , i n t h i s case, i s simply

t h e s i z e o f t h e economic U308 supply.

A more c o s t l y U308 supply would, o f course, imply a smaller maximum i n s t a l l e d capacity o c c u r r i n g e a r l i e r i n time, w h i l e t h e converse would be t r u e f o r a cheaper , U308 supply. As i s shown i n Fig. 6.2.3, i f t h e U308 supply were a f a c t o r of two larger, t h e maximum nuclear c o n t r i b u t i o n would increase from approximately 420 GWe t o approximately 730 GWe and would occur a t about t h e year 2030. If, on t h e other hand, t h e supply were a factor o f t w o smaller, t h e maximum nuclear c o n t r i b u t i o n would decrease t o approximately 250 GWe and would occur i n about t h e year 2000. A cross-plot o f t h e e f f e c t o f t h e U308 Supply on t h e maximum i n s t a l l e d nuclear capacity f o r t h e LWR on t h e throwaway c y c l e i s shown i n Fig. 6.2-4.

It i s noted i n Fig. 6.2-3

t h a t i f t h e U308 supply should be as l a r g e as 6.0

m i l l i o n ST, t h e maximum annual U308 Pequirement would be 120,000 ST/yr and t h e maximum annual enrichment requirement would be 77 m i l l i o n SWUlyr.

Given t h e probable l i m i t a t i o n on

t h e amount o f U308 t h a t could be mined and m i l l e d annually, these annual U308 requirements could be t h e l i m i t i n g f a c t o r . The e f f e c t o f adding an advanced converter (SSCR, HTGR, o r HWR) t o a nuclear power system operating on t h e throwaway c y c l e with t h e high-cost U308 supply i s shown i n The increase i n t h e nuclear c o n t r i b u t i o n f o r each o f t h e advanced converter Fig. 6.2-5. options i s r e l a t i v e l y small. A t most t h e maximum i n s t a l l e d nuclear capacity increases by approximately 30 GWe and t h e year i n which t h e maximum occurs by approximately t h r e e years. Adding t h e SSCR t o an LWR produces a s l i g h t l y g r e a t e r nuclear c o n t r i b u t i o n than adding an HTGR. This may a t f i r s t appear t o be a paradox since t h e l i f e t i m e U308 r e q u i r e ment f o r t h e HTGR i s l e s s than t h a t f o r t h e SSCR (see Fig. 6-2.1),

b u t t h e 4-yr d i f f e r e n c e

i n i n t r o d u c t i o n dates i s s u f f i c i e n t t o o f f s e t t h e d i f f e r e n c e i n U308 requirements. (The d i f ference i s n o t l a r g e enough t o be s i g n i f i c a n t , however.) The reason t h a t so small an increase i n nuclear capacity i s r e a l i z e d by i n t r o d u c i n g t h e ' v a r i o u s converters i s t h a t by t h e time they dominate t h e nuclear system a very s i g n i f i c a n t f r a c t i o n o f t h e U308 supply has already been committed t o t h e standard LWR. This i s i l l u s t r a t e d i n Fig. 6.2-6, where an HWR i n t r o duced i n 1995 does n o t become dominant u n t i l 2010.

It f o l l o w s t h a t i f t h e U308 supply were

l a r g e r w i t h t h e same nuclear growth rate, o r i f t h e nuclear growth r a t e were smaller w i t h t h e same U308 supply, t h e a d d i t i o n o f an advanced converter would have a g r e a t e r impact. This i s i l l u s t r a t e d i n Fig. 6.2-7,

f o r which t h e intermediate-cost U308 supply was assumed, and

-unless a system c o n s i s t i n g o f t h e standard LWR alone i s designated, i t i s t h e LWR system i n c l u d i n g an LWR-EE t h a t i s denoted as 1L and compared w i t h o t h e r systems i n l a t e r sections o f t h i s chapter. However, as pointed o u t here, t h e i n s t a l l e d c a p a c i t i e s o f t h e two LWR systems d i f f e r o n l y s l i g h t l y .


6-25

1mO-

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Fig. 6.2-2. The Nuclear C o n t r i b u t i o n o f LWRs on t h e Throwaway Cycle (High-Cost U308 Supply). ImO

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Fig. 6.2-5. The E f f e c t on t h e Nuclear C o n t r i b u t i o n o f Adding Advanced Converters on t h e Throwaway Cycle (High-Cost U308 Supply)

-

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Fig. 6.2-6. The U308 Commitment versus Time f o r an LWR-HWR System on t h e Throwaway Cycle (High-Cost U308 Supply).

Fig. 6.2-7. The E f f e c t on t h e Nuclear C o n t r i b u t i o n o f Adding Advanced Converters on t h e Throwaway Cycle ( Intermediate-Cost u308 supply)


6-26

Im,

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Fig. 6.2-8. The E f f e c t on t h e Nuclear C o n t r i b u t i o n o f Adding Advanced Converters on t h e Throwaway Cycle (200 GWe i n 2000 p l u s 10 GWe/yr Thereafter) (High-Cost U308 Supply 1

-

The E f f e c t o f Enrichment Fig. 6.2-9. T a i l s Composition on t h e Nuclear C o n t r i b u t i o n w i t h t h e LWR on t h e Throwaway Cycle (High-Cost U308 Supply).

L C?

b

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0 an0

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Fig. 6.2-10. The Enrichment T a i l s Composition as a Function o f Time f o r t h e Reference Case and f o r an Improving T a i l s Strategy. i n Fig. 6.2-8,

The Amounts o f U308 Fig. 6.2-11. Processed Through t h e Enrichment P l a n t s as a Function o f Time f o r t h e LWR on t h e Throwaway Cycle (High-Cost U308 Supply).

f o r which a reduced growth r a t e was assumed.

With t h e intermediate-cost

supply, t h e e f f e c t o f t h e 4-yr d i f f e r e n c e i n i n t r o d u c t i o n dates between t h e SSCR and t h e HTGR i s no longer s i g n i f i c a n t , and t h e HTGR makes t h e g r e a t e r c o n t r i b u t i o n . The e f f e c t o f changing t h e enrichment t a i l s composition upon t h e nuclear c o n t r i b u t i o n w i t h t h e LWR on t h e throwaway c y c l e i s shown i n Fig. 6.2-9 i n which t h e reference case w i t h a constant enrichment t a i l s composition o f 0.0020 i s compared w i t h two o t h e r cases: one i n which t h e enrichment t a i l s composition decreases l i n e a r l y from 0.0020 i n 1980 t o 0.0005 i n

L L

2010 and remains constant t h e r e a f t e r ; and another i n which t h e t a i l s composition s i m i l a r l y decreases and i n a d d i t i o n t h e t a i l s s t o c k p i l e accumulated p r i o r t o 2010 i s mined a t a l a t e r date w i t h a t a i l s composition o f 0.0005.

The decreasing enrichment t a i l s composition, shown

i n Fig. 6.2-10, i s t h e i n d u s t r y average, and hence t h e improving t a i l s s t r a t e g y - i m p l i e s lowe r i n g t h e t a i l s composition o f t h e gasequs d i f f u s i o n p l a n t s beginning i n 1980. I n addition,

t h e s t r a t e g y i m p l i e s a continual t r a n s i t i o n toward an i n d u s t r y based upon an enrichment process capable o f operating a t an average t a i l s composition o f 0.0005.

L


6-27

The e f f e c t o f applying t h e improving t a i l s s t r a t e g y t o a nuclear system based on t h e throwaway c y c l e i s t o increase t h e maximum i n s t a l l e d nuclear capacity by approximately 60 GWe Mining t h e t a i l s and t o delay t h e maximum by approximately f i v e years (see Fig. 6.2-9). s t o c k p i l e accumulated p r i o r t o 2010 does n o t s i g n i f i c a n t l y change t h e r e s u l t . The reason t h a t mining t h e past t a i l s s t o c k p i l e does n o t produce a s i g n i f i c a n t l y l a r g e r nuclear c o n t r i b u t i o n i s explained by Fig. 6.2-11,

which shows t h e cumulative amount o f U308 processed

through t h e enrichment p l a n t s as a f u n c t i o n o f time. t h e amount o f U,Ob

The amount i s considerably l e s s than

committed a t any given time, as shown i n Fig. 6.2-6.

It i s important t o

note t h a t t h e amount o f U308 a c t u a l l y processed through t h e enrichment p l a n t s p r i o r t o 1990 i s r e l a t i v e l y small, and a t t h i s time t h e t a i l s composition f o r t h e improving t a i l s s t r a t e g y has been decreasing l i n e a r l y f o r 10 yr.

Thus, most o f t h e U308 i n t h e improving t a i l s case

i s processed a t lower t a i l s compositions, and mining t h e past s t o c k p i l e does n o t produce a s i g n i f i c a n t improvement. The most dramatic e f f e c t associated w i t h t h e improving t a i l s o p t i o n i s t h e increase i n t h e maximum annual enrichment requirement.

As i n d i c a t e d i n Fig. 6.2-9,

t h e maximum annual U308 requirement f o r t h i s o p t i o n i s 67,000 ST/yr, annual enrichment requirement i s 92 m i l l i o n SWU/yr.

w h i l e t h e maximum

Thus, t h e p r i n c i p a l l i m i t a t i o n i n t h i s

case would be t h e a v a i l a b i l i t y o f enrichment capacity. The u t i l i z a t i o n and movement o f f i s s i l e m a t e r i a l per GWe o f i n s t a l l e d capacity i n t h e year 2035 f o r each o f t h e converter options i s shown i n Fig. 6.2-12a-d, high-cost U308 supply.

assuming t h e

These figures represent a snapshot o f t h e system i n time and i n c l u d e

t h e f i r s t core loadings f o r u n i t s s t a r t i n g up i n t h e year 2036.

As can be seen, t h e U308 con-

sumption f o r Case 1L i n t h e year 2035 i s approximately 142 ST U308/GWe, w i t h t h e LWRs having an extended discharge exposure comprising 92% o f t h e i n s t a l l e d capacity. When t h e LWRs are followed by SSCRs (Case l S ) , t h e annual U308 consumption i s 135 ST U308, w i t h t h e SSCR comp r i s i n g 74% o f t h e i n s t a l l e d capacity.

The f r a c t i o n a l i n s t a l l e d capacity o f t h e SSCR i s l e s s

than t h a t o f t h e extended-exposure LWR i n Case 1L because t h e extended-exposure LWR i s i n t r o duced'in 1981 w h i l e t h e SSCR i s n o t introduced u n t i l 1991. I n general, t h e f r a c t i o n a l i n s t a l l e d capacity of a r e a c t o r concept i n t h e year 2035 w i l l decrease monotonically as t h e i n t r o d u c t i o n date f o r t h e concept increases. S i m i l a r l y , t h e f r a c t i o n a l i n s t a l l e d nuclear c a p a c i t y of a r e a c t o r concept w i l l increase monotonically as i t s U308 requirement decreases. When t h e LWRs are f o l l o w e d by HTGRs (Case lG.), t h e U308 consumption i n t h e year 2035 i s 133 ST U308/GWe, w i t h t h e HTGR comprising 54% o f t h e i n s t a l l e d capacity. The annual U308 consumption i s lower than i n Case 1s because t h e U308 requirement o f t h e HTGR i s l e s s than t h a t o f t h e SSCR (see Table 6.1-2 and Fig. 6.2-1). The f r a c t i o n a l i n s t a l l e d capacity of t h e HTGR i s l e s s than t h a t o f t h e SSCR i n t h e Case 1s because t h e SSCR i s introduced i n 1991 w h i l e t h e HTGR i s n o t introduced u n t i l 1995. When HWRs f o l l o w t h e LWRs (Case l H ) , U3O8 consumption i n year 2035 i s approximately 106 ST U308/GWe and t h e HWR comprises 79% o f t h e i n s t a l l e d capacity. and t h e HTGR i n Case 1 G have t h e same i n t r o d u c t i o n date.

The HWR i n t h i s case

The HWR, however, has a lower

U308 requirement and hence t h e t o t a l i n s t a l l e d nuclear c a p a c i t y i s g r e a t e r w i t h t h i s


6-28

. .

HEDL 7805-090.34b - 532 51 Kg U 1,761 Kg HM

4 7.2

-

141.6 ST U308

U5CE)b-J

I

2,907 Kg H M

0.08 GWe CF = 60.3

lo3 swu

I

lo3 swu

[17,mo 543 Kg Kg U2%

)

u L

THROWAWAY

ENRICH.

80.1

L

0.92 G W e CF = 60.3

28,730 Kg H M

I L i;

Case 1L: LWRs Only; High-Cost U308 Supply.

165 Kg U235 5,650 Kg HM

HEDL 7805-090.34a 0.26 GWe CF = 60.3

23.1

lo3 SWU

THROWAWAY

53.7

L I!

lo3 swu

L 410 Kg U235 15,870 Kg H

U5(LE)/U 0.74 GWe

23.750 Kg HM

CF = 60.3

(b)

Case 1s:

LWRs Followed by SSCRs; High-Cost U308 Supply.

Fig. 6.2-12. Utilization and Consisting of Converters Operating Except f o r Case l L , which u t i l i z e s here and i n subsequent systems are

Movement of Fissile Material i n Nuclear Systems on Throwaway/Stowaway Cycle (year 2035). (Note: the extended exposure LWR, a l l LWRs included standard LWRs.)

I;

L F

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.Li

c


6-29

bd

Fig. 6.2-12

(cont.)

urn

291 Kg ,9,967 Kg HM

4 40.9 X;$

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3 8

SWU

ENRICH.

51.2 X - l $ SWU

t 260 Kg Urn -3,495 Kg HM

Y

L

5,146 Kg HM

-

CF40.3 HEDL 7805-090.31

(c)

Case 1G:

LWRs Followed by HTGRs; High-Cost U308 Supply.

18.6 x

Id SWU

J

THROWAWAY

13.9

\

HEDL 78Md90.58

(d)

Case 1H:

LWRs Followed by HWRs; High-Cost U308 Supply.


6-30

reactor.

Since t h i s increase i s due simply t o t h e c o n s t r u c t i o n o f a d d i t i o n a l HWRs, t h e

f r a c t i o n a l i n s t a l l e d capacity o f t h e HWR i s increased comensurately. I n summary, using t h e assumptions contained i n t h i s study, t h e f o l l o w i n g conclusions can be drawn about t h e behavior o f a nuclear power system operating on t h e throwaway option: (1) The e f f e c t o f deploying an advanced converter i n 1995, under t h e assumption o f 350 GWe i n t h e year 2000 and 15 GWe/yr t h e r e a f t e r w i t h t h e high-cost U308 supply, would be small. (2) I f t h e U308 supply a v a i l a b l e below $160/lb should be l a r g e r than 3 m i l l i o n ST, o r i f t h e nuclear growth should be smaller than assumed above, then t h e e f f e c t o f deploying t h e advanced converter would be l a r g e r .

( 3 ) The e f f e c t o f reducing t h e enrichment t a i l s composition i s somewhat l a r g e r than t h a t o f deploying an advanced converter under t h e assumed conditions. (4) The dominant v a r i a b l e f o r t h e nuclear power system on t h e throwaway cycle i s

t h e U308 supply; a U308 supply e i t h e r t w i c e as l a r g e o r t w i c e as small i s o f g r e a t e r consequence than any o f t h e e f f e c t s discussed above.

6.2.2.

L h'

L L L

c

Converter System w i t h Plutonium Recycle

I n order t o assess t h e o p t i o n o f plutonium r e c y c l e i n converters iwas assumed -..at a reprocessing c a p a b i l i t y would be a v a i l a b l e i n 1991. (This assumption does n o t argue t h a t t h e 'reprocessing capacity would be economically a t t r a c t i v e o r d i v e r s i o n - r e s i s t a n t , but I n t h i s option the classimerely t h a t i t would be t e c h n o l o g i c a l l y f e a s i b l e by t h i s date.) c a l plutonium r e c y c l e was m o d i f i e d somewhat by r e j e c t i n g converters w i t h self-generated r e c y c l e i n f a v o r o f converters w i t h complete plutonium loads.

T h i s has t h e advantage o f

reducing t h e number o f r e a c t o r s t h a t must be placed i n t h e energy centers and comnensuratel y increases t h e number o f r e a c t o r s t h a t can be placed o u t s i d e t h e centers. The i n d i v i d u a l r e a c t o r concepts and t h e i r l o c a t i o n s are shown i n Fig. 6.1-2

G h'

(Option 2).

A comparison o f t h e nuclear c o n t r i b u t i o n o f t h e LWR w i t h plutonium r e c y c l e t o t h a t o f t h e LWR on t h e throwaway c y c l e (Fig. 6.2-13) shows t h a t w i t h r e c y c l e t h e maximum ins t a l l e d nuclear c a p a c i t y i s increased from approximately 420 GWe t o approximately 600 GWe and t h e time a t which t h e maximum occurs i s increased from about year 2010 t o about year 2020 (high-cost UJ08 supply).

The maximum annual U308 requirement for t h i s case i s

67,000 ST/yr and t h e maximum annual enrichment requirement i s 46 m i l l i o n SWU/yr.

These

L


6-31

IWO

I

I

I

I

I

I

c*Sf 2L -THE LWI WITH PWIONIUM RfCYCU

THE LWR WllH PWTONIUM RECYCLE

W I M U M ANNUAL RECUIREMEM:

yo.

-

67 '

I -

IL

2

--

I ' c,

1980

1 m

Zi

Id IT/*-

2003

-

-

LWR WITH PLUTONIUM RECYCLE

LWR ON THE THROWAWAY

mo

nno

2mo

20(0

2mo

Fig. 6.2-13. The E f f e c t on t h e Nuclear C o n t r i b u t i o n of Recycling Plutonium i n LWRs (High-Cost U308 Supply). IWO

I

I

I

I

Fig. 6.2-14. R e l a t i v e Nuclear C o n t r i butions o f LWRs Located I n s i d e (LWR-Pu) and Outside (LWR-U) Energy Centers (High-Cost u308 supply)

-

requirements do n o t d i f f e r s i g n i f i c a n t l y from

I

-Hf LWR WIIH PWIONIUM RECYCLE

those o f t h e LWR on t h e throwaway c y c l e (see Fig. 6.2-2)

because t h e nuclear growth pro-

j e c t i o n was s p e c i f i e d t o be 350 GWe i n t h e

u

1 '

b

year 2000 plus 15 GWe/yr thereafter. Thus, t h e primary e f f e c t o f reprocessing i s t o a l l o w t h e nuclear system t o grow beyond t h e 400-GWe l e v e l even though a s c a r c i t y o f U308 e x i s t s

c

I IPBO

IWO

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zom

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fou)

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a t costs below $160/lb.

Viewed.differently,

t h e primary e f f e c t o f reprocessing i s n o t

YUR

t o support t h e c o n s t r u c t i o n o f a d d i t i o n a l Fig. 6.2-15. The Effect o f U308 Supply on t h e Nuclear C o n t r i b u t i o n o f t h e LWR w i t h Plutonium Recycle (Case 2L).

nuclear u n i t s i n t h e e a r l i e r years when U308 i s i n p l e n t i f u l supply.

The i n s t a l l e d nuclear c a p a c i t y t h a t must be l o c a t e d i n t h e energy centers as a f u n c t i o n o f time i s shown by t h e lower curve i n Fig. 6.2-14, t h e d i f f e r e n c e between t h e two curves i n d i c a t i n g t h e nuclear capacity t h a t can be made a v a i l a b l e outside t h e centers. The maximum c a p a c i t y which must be l o c a t e d i n t h e energy centers i s approximately 260 GWe, w h i l e a maximum o f 400 GWe can be a v a i l a b l e outside t h e center. For approximately t h r e e decades (from t h e year 2000 t o t h e year 2030), over 300 GWe can be a v a i l a b l e outside t h e centers.

The use o f plutonium r e c y c l e t o a l l o w t h e nuclear system t o grow beyond t h e

400-GWe l e v e l as t h e U308 supply becomes scarce i s v i v i d l y i l l u s t r a t e d i n Fig. 6.2-14. Note t h a t t h e number o f u n i t s loaded w i t h plutonium increases s i g n i f i c a n t l y as t h e i n s t a l l e d capacity exceeds t h e 400-GWe l e v e l and t h a t they comprise an increasing f r a c t i o n of the t o t a l i n s t a l l e d c a p a c i t y i n l a t e r years.


6-32

-

12

87 Kg f i s Pu 138 Kg $ 3 14,780 Kg HM

Kg HM

I â&#x20AC;&#x2DC;9.

, 1 ST

L L

â&#x20AC;&#x2DC;3O8

HEDL 7806-090.30 Fig. 6.2-16. U t i l i z a t i o n and Movement o f F i s s i l e M a t e r i a l i n a Nuclear System Consisting o f LWRs Operating w i t h Plutonium and/or Uranium Recycle (Case 2L, HighCost U308 supply) (Year 2035).

The e f f e c t o f t h e intermediate-cost U308 supply on t h e LWR plutonium r e c y c l e case i s shown i n Fig. 6.2-15.

With 6.0 m i l l i o n ST U308 below $160/lb,

t h e maximum nuclear

c o n t r i b u t i o n would increase from approximately 600 GWe i n t h e year 2020 t o approximately 960 GWe i n t h e year 2045. Thus, t h e U308 supply i s again t h e dominant variable. The maximum annual U308 requirement would be 110,000 ST/yr and t h e maximum annual enrichment requirement would be 72 m i l l i o n SWU/yr. These annual requirements would c o n s t i t u t e t h e

a,

p r i n c i p a l l i m i t a t i o n o f t h e system.

The u t i l i z a t i o n and movement o f f i s s i l e m a t e r i a l per GWe o f i n s t a l l e d c a p a c i t y f o r t h e LWR w i t h plutonium r e c y c l e i s shown i n Fig. 6.2-16. Again t h i s f i g u r e represents a snapshot o f t h e system i n time ( i n t h e year 2035) and includes both t h e f i r s t core loading

il

f o r those r e a c t o r s t h a t are s t a r t i n g up and t h e l a s t core discharge f o r those r e a c t o r s t h a t a r e s h u t t i n g down.

The annual U308 consumption i n 2035 i s 59 ST U308/GWe, and t h e LWR

u t i l i z i n g plutonium comprises 54% o f t h e i n s t a l l e d capacity.

Approximately 368 kg o f

f i s s i l e plutonium i n f r e s h f u e l per GWe o f i n s t a l l e d c a p a c i t y per year must be handled w i t h i n t h e energy centers f o r t h i s case.

(Note:

Simply i d e n t i f y i n g t h e amount o f

f i s s i l e plutonium i n f r e s h f u e l t h a t must be handled i s n o t analogous t o determining t h e d i v e r s i o n r e s i s t a n c e o f the system.

L

While t h e amount o f f i s s i l e plutonium being handled

may be important, t h e s t a t e and l o c a t i o n o f t h e f i s s i l e plutonium and t h e procedures used t o handle i t are more important i n assessing t h e d i v e r s i o n r e s i s t a n c e o f a system.)

L


c -I

.

t

6-33

I n summary, a converter s t r a t e g y based on LWRs w i t h plutonium r e c y c l e could supply a maximum nuclear c o n t r i b u t i o n o f 600 GWe w i t h a U308 supply o f 3.0 m i l l i o n ST a t below $160/lb.

This i s 180 GWe more than t h e maximum nuclear c o n t r i b u t i o n obtained w i t h t h e LWR

on t h e throwaway cycle; however, i t i s l e s s than t h e maximum nuclear c o n t r i b u t i o n o f 730 GWe obtainable on t h e throwaway c y c l e w i t h a U308 supply o f 6.0 m i l l i o n ST a t beTow $160/lb. Also, converter s t r a t e g y based on LWRs w i t h plutonium r e c y c l e w i l l r e q u i r e t h a t as much as 260 GWe be located i n t h e energy centers.

I '

ct

6.2.3.

Converter System w i t h Plutonium Throwaway

Under Option 4 (see Fig. 6.1-3)

i t i s assumed t h a t t h e nuclear p o l i c y i s t o d e f e r use

--

o f plutonium u n t i l some i n d e f i n i t e future date and t o operate a l l converters on low-enriched

L

o r denatured uranium.

I

The a c t i v i t i e s i n t h e energy center are thus l i m i t e d t o reprocessing,

uranium f u e l f a b r i c a t i o n , and plutonium storage.

As shown i n Fig. 6.2-17,

w i t h t h e high-

L

c o s t U308 supply, t h e nuclear c o n t r i b u t i o n i n t h i s case reaches a maximum o f approximately 590 GWe i n about 2020, which i s a s i g n i f i c a n t increase over t h a t o f the(UtPu) throwaway

t

plutonium recycle.

case, and, i n f a c t , i s q u i t e comparable t o t h e maximum nuclear capacity obtained w i t h However, the reactors employed minimize t h e production o f plutonium and

t h e r e f o r e t h e amount u l t i m a t e l y thrown away.

This, coupled w i t h t h e f a c t t h a t 233U i s worth

s l i g h t l y more than 239Pu i n a thermal reactor, allowed the system w i t h plotonium throwaway t o u l t i m a t e l y achieve t h e same nuclear c o n t r i b u t i o n as t h e system w i t h plutonium recycle. The maximum annual U308 and enrichment requirements were found t o be 80,000 ST/yr and 69 m i l l i o n SWU/yr.

This ore requirement i s 20% greater than t h a t f o r t h e case of LWR plutonium

recycle, and t h e enrichment requirement i s 50% greater.

-

i

The increases can be d i r e c t l y a t -

t r i b u t e d t o t h e U308 and enrichment requirements o f t h e denatured LWR loaded w i t h 15% 235U i n As i l l u s t r a t e d i n Table 6.1-2, t h e l i f e t i m e U308 and enrichment requirements o f t h i s 23*U. r e a c t o r are 24% and 64% g r e a t e r than t h e same requirements f o r t h e standard LWR. The e f f e c t o f t h e intermediate-cost U308 supply f o r t h i s case i s shown i n Fig. 6.2-18.

The maximum nuclear c o n t r i b u t i o n increases from approximately 590 GWe i n about

year 2020 t o approximately 980 GWe i n about year 2045.

Again t h e c o n t r i b u t i o n o f t h e

system i s comparable t o t h a t o f t h e LWR plutonium r e c y c l e case, and again t h e maximum annual U308 and enrichment requirements, 105,000 ST/yr and 100 m i l l i o n SWU/yr,

respec-

t i v e l y , w i l l represent t h e p r i n c i p a l l i m i t a t i o n s o f t h e system. The u t i l i z a t i o n and movement o f f i s s i l e m a t e r i a l per GWe of i n s t a l l e d capacity f o r Case 4L i n t h e y e a r 2035 are shown i n Fig. 6,2-19. The U308 consumption, i n c l u d i n g t h e f i r s t core loadings and l a s t discharges, i s 32 ST U308/GWe. The standard LWR loaded w i t h approximately 3% enriched 235U comprises 5% o f t h e i n s t a l l e d nuclear capacity, t h e denatured LWR loaded w i t h 15% enriched 235U comprises 39%, and t h e denatured LWR loaded w i t h 11% 233U i n


6-34

1HE LWR WITH FISSILE U U N I U M PCYCLE AND PLUTONIUM THROWAWAY

A

WXIMUM ANNUAL RfWIREMENlS:

ENRICHMENT

Fig. 6.2-17. The Effect on the Nuclear Contribution of LWRs Operating with Fissile Uranium Recycle and Plutonium Throwaway (High-Cost U308 Supply).

-32.0

Fig. Supply on Operating Plutonium

- im .

iB

i

lNTERMEDlATE COST U$B SUWLV 160. I $ S l BELOW S 1 M LO,

6.2-18. The Effect of the U308 the Nuclear contribution of LWRs with Fissile Uranium Recycle and Throwaway (Case 4L).

ST U30e

HEDL 7805-090.33

Fig. 6.2-19. Utilization and Movement o f Fissile Material in a Nuclear System Consisting of LWRs Operating with Fissile Uranium Recycle and Plutonium Throwaway (Case 4L, High-Cost U308 Supply) (Year 2035).

L L


6-35

238U

comprises 57%.

The p r i n c i p a l advantage associated w i t h t h i s o p t i o n i s t h a t a l l nuclear

u n i t s can be l o c a t e d outside t h e energy centers, which means t h a t t h e amount o f f i s s i l e p l u tonium i n f r e s h f u e l t h a t must be handled i n t h i s system i s zero. This advantage i s n o t w i t h o u t cost, however; i t requires t h e development o f an i n d u s t r y capable o f reprocessing s i g n i f i c a n t q u a n t i t i e s o f f u e l containing thorium and r e f a b r i c a t i n g s i g n i f i c a n t q u a n t i t i e s o f f u e l containing 232U.

I n order t o successfully implement t h i s option, one must develop a

nuclear i n d u s t r y i n which approximately 95% o f t h e reprocessing capacity i n t h e year 2035 i s capable of handling fuel containing thorium and 57% o f t h e f a b r i c a t i o n capacity i s capable o f handling f u e l containing 232U. I n sumary, ifemployed j u d i c i o u s l y , a converter s t r a t e g y based on t h e LWR can be developed which can discard a l l f i s s i l e plutonium and s t i l l supply a maximum nuclear cont r i b u t i o n of 590 GWe w i t h a U3O8 supply o f 3.0 m i l l i o n ST below $160/lb. This i s essent i a l l y i d e n t i c a l t o t h a t of the c l a s s i c a l LWR plutonium r e c y c l e w i t h t h e same U308 supply.

L

With a U308 supply of 6.0 m i l l i o n ST below $160/lb, t h e system could supply a maximum nuclear c o n t r i b u t i o n o f 980 GWe; however, as pointed o u t above, considerable development work would be r e q u i r e d on f u e l design and f a b r i c a t i o n . 6.2.4.

Converter System w i t h Plutonium Production Minimized; P u - ~ o - 33U ~ "Transmutation"

L

L rIi

t b

c L h:

An i n h e r e n t disadvantage i n the plutonium throwaway o p t i o n discussed above i s t h a t t h e f i s s i l e plutonium produced i n t h e system i s never u t i l i z e d . Therefore, i t was considered des i r a b l e t o analyze an o p t i o n i n which f i s s i l e plutonium produced i n a s i m i l a r system i s used t o produce 233U f o r t h e dispersed reactors. The 233U producer would be a converter w i t h a plutonium-thorium core. This converter would, o f course, be l o c a t e d i n an energy center, w h i l e t h e o t h e r reactors would be located outside t h e center as shown i n Fig. 6.1-3 (Option 5U). I t i s important t o note t h a t w h i l e t h i s o p t i o n u t i l i z e s a l l t h e p l u t o n i u n produced i n t h e system, i t minimizes the amount of pZutoniwn that is produced. This requires t h e development o f r e a c t o r concepts designed s p e c i f i c a l l y t o minimize plutonium production. The nuclear c o n t r i b u t i o n o f t h i s o p t i o n u t i l i z i n g LWRs o n l y (Case 5UL) reaches a maximum o f approximately 700 GWe s h o r t l y before year 2030 (see Fig. 6.2-20). Thus, u t i l i z i n g t h e plutonium produced i n t h e system increases the maximum nuclear c o n t r i b u t i o n by approximately 100 GWe over t h a t o f t h e o p t i o n with plutonium throwaway; i t a l s o produces a delay i n t h e maximum o f about e i g h t years (compare w i t h Fig. 6.2-17). The maximum annual

u&

and enrichment requirements f o r t h i s o p t i o n are 75,000 ST/yr and 65

m i l l i o n SWUlyr, r e s p e c t i v e l y , each being approximately 6% l e s s than t h a t r e q u i r e d for Option 4. The amount of t h e system's i n s t a l l e d nuclear capacity t h a t must be located i n t h e energy center i s shown i n Fig. 6.2-21 as a f u n c t i o n of time. This o p t i o n i s d i s t i n g u i s h e d by t h e f a c t t h a t t h e maximum capacity t h a t must be located i n a secure region does n o t exceed 100 GWe a t any time d u r i n g t h e planning horizon.

The amount t h a t may be l o c a t e d outside t h e


6-36

energy center ranges from approximately 300 GWe i n t h e year 2000 t o approximately 600 GWe i n the year 2025.

The disadvantage o f t h i s o p t i o n i s t h a t t h e h i g h energy support r a t i o

( t h e amount o f capacity t h a t can be l o c a t e d outside t h e energy center d i v i d e d by t h e amount t h a t must be located i n t h e center) cannot be maintained i n d e f i n i t e l y . I n f a c t , t h e energy support r a t i o decreases continuously as t h e end o f t h e U308 supply i s approached. ImO THE LWR WITH PURONIUM MlNUllUTlON AND U I I L I U T I O N

at.m

t

z m

-

-

,a

-

-

LWR WITH *unoNIvM MINIMIUTIONAN0,UIILIUlION~ AN0 UIlLlUllON

I I I I I CASE N L -THE LWR WITH WJlONIUM MINIMI7ATION AN0 LlllLlUllON

I;

I

1

em-

rmt

I,

L

6

0 1980

I I990

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2010

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2010

1010

YEAR

=

1010

2t60

YEA#

Fig. 6.2-20. The E f f e c t o f Minimizing t h e Production and Use o f Plutonium i n LWRs (High-Cost U308 Supply).

CASE SJL

-

f

"'

kd

L

THE LWR WITH PLUTONIUM

THE LWR FOLLWEO M AN ADVANCED CONVfRlER WITH ?LUlONIUM MINIMIUltON AND UIlllZAllON

M4XIMUM ANNUL REOUlRtMtNl

tl

*IC" COST ",Os SUPPL" 10 106SlOtLOI"SlWLB

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OI9,

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1990

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1020

1010

1030

1010

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s o

YEAR

Fig. 6.2-22. The E f f e c t o f U308 Supply on t h e Nuclear C o n t r i b u t i o n o f LWRs Operating w i t h Plutonium Minimization and U t i 1i z a t i o n .

The E f f e c t on t h e Nuclear Fig. 6.2-23. C o n t r i b u t i o n o f Adding Advanced Converters Operating w i t h Plutonium M i n i m i z a t i o n and U t i l i z a t i o n (High-Cost U308 Supply).

The h i g h energy support r a t i o could be maintained f o r a longer p e r i o d o f time, however,

ift h e U3O8 supply were l a r g e r . Figure 6.2-22 shows t h a t doubling t h e U308 supply would increase t h e maximum nuclear c o n t r i b u t i o n o f t h e system from approximately 700 GWe i n year 2039 t o over 1000 GWe i n year 2050. Since t h e maximum energy support r a t i o occurs a t about t h e same-time as t h e maximum nuclear c o n t r i b u t i o n , i t can be assumed t h a t w i t h t h e increased U & supply a l a r g e energy support r a t i o could be maintained as f a r i n t o t h e f u t u r e as year 2050. Given t h e U308 supply, i t would appear t h a t t h e p r i n c i p a l l i m i t a t i o n f o r t h i s o p t i o n

would be t h e maximum annual U308 and enrichment requirements, which are 115,000 ST/yr and

90 m i l l i o n SWU/yr, r e s p e c t i v e l y .

L I: I'

k.


6-37

The e f f e c t upon t h e nuclear c o n t r i b u t i o n o f adding advanced converters w i t h t h e LWRS i s shown i n Fig. 6.2-23 f o r t h e high-cost U308 supply. The HWR has t h e l a r g e s t effect, increasing t h e nuclear con,$ribution of t h e system t o approximately 810 GWe i n year 2035. The l a r g e r e f f e c t of t h e advanced converters i n t h i s o p t i o n compared t o t h e i r e f f e c t i n

HEDL 7805490.57

Fig. 6.2-24. U t i l i z a t i o n and Movement of F i s s i l e M a t e r i a l i n an LWR Nuclear System Minimizing t h e Production and Use o f Plutonium (Case 5UL, High-Cost U308 Supply) (Year 2035). t h e throwaway o p t i o n f o r t h i s ore supply i s p r i m a r i l y due t o t h e f a c t t h a t reprocessing i s a v a i l a b l e i n t h i s case. The a v a i l a b i l i t y o f reprocessing e f f e c t i v e l y increases t h e amount of U3O8 a v a i l a b l e a f t e r t h e advanced converters are introduced and t h e r e f o r e increases t h e amount of U308 upon which t h e advanced converters can employ t h e i r resource savings. The u t i l i z a t i o n and movement i n year 2035 o f f i s s i l e m a t e r i a l per GWe o f i n s t a l l e d * c a p a c i t y f o r t h e system u t i l i z i n g LWRs o n l y (Case 5UL) i s shown i n Fig. 6.2-24. The annual

U&

consumption i s approximately 36 ST U&/GWe.

The LWR transmuting plutonium t o 233U

i s supplied w i t h approximately 170 kg o f f i s s i l e plutonium i n f r e s h f u e l per GWe o f i n s t a l l e d capacity and i t comprises 13% o f t h e i n s t a l l e d capacity. This can be compared t o t h e c l a s s i c a l case o f plutonium r e c y c l e i n which approximately 54% o f t h e i n s t a l l e d capacity must *The movement o f f i s s i l e m a t e r i a l i n a l l cases i s a f u n c t i o n o f time, Furthermore, i t i s a f f e c t e d by f i r s t - c o r e charges and l a s t - c o r e discharges (which a r e included i n Fig. 6.2-24 and subsequent s i m i l a r f i g u r e s ) . The f i s s i l e balance f o r a decaying (or growing) system d i f f e r s s i g n i f i c a n t l y from t h a t o f a s t a t i c system.


6-38

be located i n energy centers and 368 kg o f f i s s i l e plutonium i n f r e s h f u e l per GWe o f i n s t a l l ed capacity must be handled each year i n those centers. This i s n o t meant t o imply t h a t a decrease i n t h e amount o f nuclear capacity which must be placed i n secure regions i s synonymous w i t h an increase i n diversion-resistance. Neither i s i t meant t o imply t h a t a decrease i n the amount of f i s s i l e plutonium which must be handled as f r e s h f u e l i s synonymous w i t h an increase i n p r o l i f e r a t i o n resistance. I f e i t h e r o f these items i s desirable, however, t h i s o p t i o n minimizing t h e production and use o f plutonium does o f f e r a s i g n i f i c a n t increase i n t h e energy support r a t i o and a s i g n i f i c a n t decrease i n t h e amount o f f r e s h - f u e l plutonium t h a t must be handled. I t i s important t o note t h a t t h e deployment o f the plutonium m i n i m i z a t i o n and

u t i l i z a t i o n o p t i o n would r e q u i r e t h e development o f a nuclear i n d u s t r y capable o f reprocessing f u e l containing thorium and r e f a b r i c a t i n g f u e l containing 232U. As Fig. 6.2-24 indicates, o n l y o n e > r e a c t o r p r o v i d i n g 3% o f t h e i n s t a l l e d capacity i n year 2035 does n o t u t i l i z e thorium. Thus, i n order t o s u c c e s s f u l l y implement t h i s option, 97% o f t h e reprocessing capacity i n year 2035 must be capable o f handling f u e l containing thorium, and 51% o f t h e f a b r i c a t i o n capacity must be capable o f handling f u e l containing 232U. I n sumnary, a converter s t r a t e g y based on t h e LWR which minimizes t h e amount of plutonium produced, b u t uses t h a t which i s produced, could supply a maximum nuclear cont r i b u t 4 o n of 700 GWe w i t h t h e high-cost U308 supply. This i s approximately 100 GWe g r e a t e r than t h e maximum nuclear c o n t r i b u t i o n obtained i n t h e case o f plutonium throwaway and f i s s i l e uranium recycle.

The s t r a t e g y does, however, r e q u i r e t h a t approximately

100 GWe be l o c a t e d i n an energy center.

With t h e intermediate-cost U308 supply, t h e system

could make a maximum nuclear c o n t r i b u t i o n o f more than 1000 GWe. I n e i t h e r case, t h e development of f u e l designs capable o f minimizing t h e amount of plutonium produced and a l s o t h e development o f a nuclear i n d u s t r y capable o f hand1 i n g thorium-based f u e l s must be developed. 6.2.5.

Converter System w i t h Plutonium Production Not ~ ~ U Minimi zed; P u - ~ o - ~ "Transmutation"

This o p t i o n d i f f e r s from t h e preceding o p t i o n

i n t h a t t h e dispersed reactors are n o t designed t o minimize t h e amount o f plutonium produced. Thus more plutonium i s handled as fresh f u e l and more i s "transmuted" i n t o 233U. Again a converter w i t h a plutonium-thorium core i s located i n t h e energy center, and o t h e r reactors are l o c a t e d o u t s i d e t h e center (see Fig. 6.1-3, Option 5T). Figure 6.2-25

shows t h a t t h e nuclear c o n t r i b u t i o n f o r t h i s o p t i o n using LWRs o n l y

(Case 5TL) reaches a maximum o f approximately 640 GWe s h o r t l y before year 2025. The maximum c o n t r i b u t i o n i s l e s s than t h e 700-GWe maximum i n t h e preceding case p r i m a r i l y because o f t h e d i f f e r e n t amounts o f f i s s i l e plutonium u t i l i z e d i n t h e two systems.

Since 239Pu i s worth l e s s i o a thermal r e a c t o r than e i t h e r 235U o r 233U, t h e system which minimizes t h e amount o f p l u tonium should (and does) make a s l i g h t l y l a r g e r nuclear c o n t r i b u t i o n .


6-39

The f r a c t i o n o f t h e i n s t a l l e d nuclear capacity which f o r t h i s case must be l o c a t e d The maximum i s approximately

i n energy centers i s shown i n Fig. 6.2-26 as a f u n c t i o n o f time.

--

I;

L L

120 GWe, which i s s l i g h t l y greater than t h a t f o r t h e previous case. The amount o f nuclear capacity a v a i l a b l e f o r l o c a t i o n outside energy centers ranges from approximately 300 GWe i n t h e year 2000 t o approximately 500 GWe i n t h e year 2025. The maximum annual U308 and enrichment requirements are 65,000 ST/yr and 45 m i l l i o n SWU/yr, respectively. These are q u i t e s i m i l a r t o t h e maximum annual requirements f o r t h e case of t h e LWR w i t h c l a s s i c a l plutonium r e c y c l e (see Fig. 6.2-13).

The disadvantage o f t h i s o p t i o n i s t h a t t h e energy support r a t i o decreases continuously as t h e end o f t h e U308 supply i s approached. Figure 6.2-27 i n d i c a t e s t h a t i f a U308 supply o f 6.0 m i l l i o n ST below $160/lb were a v a i l a b l e , t h e system would continue t o grow

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LWR WITH PU~ONIUH

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U S E 5lL

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TDAMMIIIATION

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1

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LWR WTH PLUTONIUM TDAMMUTATION /

- THROIIAWAY LWR ON THE CYCE

YEAR

YEAR

Fig. 6.2-25. The E f f e c t on the Nuclear C o n t r i b u t i o n o f "Transmuting" P1u t o n i um Produced i n LWRs t o 233U (High-Cost U308 Supply)

-

IOX

I

I

I

LWR WllH PLUTONIUM IRAMMUTATION

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Fig. 6.2-26. R e l a t i v e Nuclear C o n t r i butions o f LWRs Located I n s i d e (Pu/Th) and Outside (Denatured LWRs) Ener y Centers (Plutonium "Transmuted" t o 2 3 U) (High-Cost U308 SUPPlY) *

4

u n t i l approximately year 2050, and thus t h e h i g h energy support r a t i o associated w i t h t h i s o p t i o n could be maintained much longer. The maximum annual U308 and enrichment requirements i n t h i s case are 109,000 ST/yr and 77 m i l l i o n SWU/yr, r e s p e c t i v e l y . Thus, again we have an o p t i o n f o r which t h e p r i n c i p a l l i m i t a t i o n would be t h e annual ore and enrichment requirements.

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Fig. 6.2-27. The E f f e c t o f U308 Supply on t h e Nuclear C o n t r i b u t i o n o f LWRs 5n System with Plutonium "Transmutation" (Case 5TL). comprises 18%o f t h e i n s t a l l e d capacity.

The u t i l i z a t i o n and movement o f f i s s i l e m a t e r i a l per GWe of i n s t a l l e d capacity f o r Case 5TL i n t h e year 2035 are shown i n Fig. 6.2-28. The annual U3O8 consumption i s approximately 68 ST U,08/GWe,

and the LWR u t i l i z i n g plutonium

Approximately 260 kg o f f i s s i l e plutonium per GWe o f


i; 6-40

i n s t a l l e d capacity must be handled as f r e s h f u e l each year w i t h i n t h e energy centers. This can be compared t o the c l a s s i c a l case o f plutonium r e c y c l e i n which 56% o f t h e i n s t a l l e d capacity i s located i n t h e energy centers and 368 kg o f f i s s i l e plutonium i s handled as f r e s h f u e l each year.

Thus, using t h e plutonium t o produce

233U r e s u l t s i n a s i g n i f i c a n t reduction

i n t h e amount o f i n s t a l l e d capacity t h a t must be located i n secure regions, and i t a l s o reduces t h e amount o f f i s s i l e plutonium t h a t must be handled as f r e s h f u e l each year.

L L L

-

REFABRICATION

67.8 ST U308

L HEDL 7805490.56

U t i l i z a t i o n and Movement o f F i s s i l e M a t e r i a l i n an LWR Nuclear System Fig. 6.2-28. "Transmuting" Plutonium t o 2 3 3 U (Case 5TL, High-Cost U308 Supply) (Year 2035).

As f o r t h e preceding option, t h e h i g h energy support r a t i o associated w i t h t h i s case

fl

r e q u i r e s t h e development o f a nuclear i n d u s t r y capable o f reprocessing s i g n i f i c a n t amounts o f f u e l containing thorium and r e f a b r i c a t i n g s i g n i f i c a n t amounts o f f u e l c o n t a i n i n g 232U, although these amounts are considerably smaller. As Fig. 6.2-28 indicates, t h e LWR loaded with approximately 3%enriched 235U comprises 62% o f t h e i n s t a l l e d capacity i n year 2035,

~

t h e LWR loaded w i t h Pu i n Th comprises 18%, and t h e LWR loaded w i t h 12% 233U i n 2 3 * U comprises 20%. Thus approximately 34% o f t h e reprocessing capacity must be capable o f handling f u e l containing thorium and 20% o f t h e f a b r i c a t i o n capacity must be capable o f handling f u e l cont a i n i ng 232U.

In sumnary, a converter s t r a t e g y based on t h e LWR which "transmutes" a l l plutonium t o 233U could supply a maximum nuclear c o n t r i b u t i o n o f 640 GWe w i t h t h e high-cost U308 supply, o f which about 120 GWe would be l o c a t e d i n energy centers. While t h e nuclear cont r i b u t i o n f o r t h i s case i s somewhat l e s s than f o r t h e case i n which t h e production o f plutonium i s minimized, i t does n o t r e q u i r e t h e development o f new r e a c t o r concepts and i t w i l l r e q u i r e handling smaller amounts o f 233U.

L I' #Ur


b

6-41

6.2.6

Converter-Breeder System w i t h L i g h t Plutonium "Transmutation"

The r e s u l t s presented i n t h e preceding sections have demonstrated t h a t nuclear power systems based on converter reactors w i l l u l t i m a t e l y be l i m i t e d by t h e q u a n t i t y of economically recoverable uranium. While a l a r g e r U308 resource base w i l l a l l o w l a r g e r systems t o develop, t h e converse i s a l s o true.

Since t h e U308 resource base has always

been somewhat uncertain, t h e deployment o f f a s t breeder reactors has t r a d i t i o n a l l y been considered as t h e method by which the consequences o f t h i s u n c e r t a i n t y would be minimized. Thus, i t has h i s t o r i c a l l y been assumed t h a t by deploying'FBRs nuclear power systems would outgrow t h e c o n s t r a i n t s n a t u r a l l y imposed by t h e U308 resource base.

-

L

L

I n t h e o p t i o n discussed here (Option 6), an FBR w i t h a plutonium-uranium core and a thorium b l a n k e t i s l o c a t e d i n the energy center t o produce 233U which i s then used i n denatured converter r e a c t o r s outside t h e center. Because a higher plutonium "transmutation" r a t e could be obtained w i t h a plutonium-thorium core i n the FBR, t h i s o p t i o n i s r e f e r r e d ~U" rate. t o as having a Zight " P u - ~ o - ~ ~ transmutation

The i n d i v i d u a l r e a c t o r concepts

contained i n t h i s o p t i o n are shown i n Fig. 6.1-4. The nuclear c o n t r i b u t i o n associated w i t h t h i s o p t i o n when a l l t h e converters u t i l i z e d a r e LWRs (Case 6L) i s shown i n Fig. 6.2-29.

I n t h i s case, even w i t h t h e high-cost U308

supply, t h e system i s capable o f maintaining a n e t a d d i t i o n r a t e o f 15 GWe/yr throughout t h e planning horizon

L-

i

-

i.e.,

from 1980 through 2050.

The a b i l i t y o f t h e nuclear system

t o maintain t h i s n e t a d d i t i o n r a t e i s a d i r e c t consequence o f t h e compound system doubling time o f t h e FBR, which, i n t h i s case, i s 13 yr. This doubling time i n t u r n i s a d i r e c t consequence o f t h e FBR having a Pu-U core. I n t h i s o p t i o n t h e i n s t a l l e d nuclear capacity which must be located i n energy centers increases as a f u n c t i o n o f time t o approximately 560 GWe i n year 2050 (see Fig. 6.2-30). The most r a p i d increase occurs between 2010 and 2020 as the number o f FBRs on l i n e i n creases s i g n i f i c a n t l y . The amount o f nuclear capacity a v a i l a b l e f o r i n s t a l l a t i o n outside the centers increases from approximately 300 GWe i n year 2000 t o over 500 GWe i n year 2050. I n i t i a l l y , t h e LWR loaded w i t h approximately 3% enriched 235U i s t h e p r i n c i p a l r e a c t o r a v a i l a b l e , b u t as the U308 i s depleted, i t i s replaced by the LWR loaded w i t h 11%233U i n 238U. This i s i l l u s t r a t e d i n Fig. 6.2-31, which a l s o i n d i c a t e s t h a t t h i s o p t i o n i s

e

capable o f maintaining an energy support r a t i o greater than u n i t y throughout the planning horizon. The maximum annual U308 and enrichment requirements f o r t h i s case are 62,000 !jT/yr and 44 m i l l i o n SWUlyr, r e s p e c t i v e l y .

These annual requirements do n o t d i f f e r s i g n i f i c a n t l y from those obtained with t h e LWR on the throwaway cycle, t h e reason being t h a t i n e i t h e r

case, t h e goal o f t h e nuclear power system i s t o maintain a n e t a d d i t i o n r a t e o f 15 GWe/yr provided t h i s increase can be sustained by the U308 supply.

The maximum i n s t a l l e d capacity


I; 6-42

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Fig. 6.2-30. R e l a t i v e Nuclear C o n t r i butions o f Reactors Located I n s i d e (Pu-Fueled) and Outside (Denatured LWRs) Energy Centers (High-Cost U308 Supply).

Fig. 6.2-29. The Nuclear C o n t r i b u t i o n o f an LWR-FBR System w i t h L i g h t Plutonium "Transmutation1' (High-Cost U308 Supply).

wf

L

f o r t h e LWR loaded w i t h approximately 3% enr i c h e d 235U i n e i t h e r case i s approximately

TMMMUI*TION

420 GWe.

I '

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However, i n t h i s option, as t h e i n -

s t a l l e d capacity o f t h e 235U-loaded LWRs decreases, t h e energy c e n t e r FBRs produce i n creasing amounts o f 233U f o r t h e denatured LWRs, and thus t h e t o t a l i n s t a l l e d nuclear capacity cont i n u e s t o increase a t a n e t r a t e o f 15 GWe/yr. 0I9e4

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The amount o f f i s s i l e plutonium t h a t must be handled i n t h e energy centers as f r e s h f u e l Approxieach year i s shown i n Fig. 6.2-32.

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Fig. 6.2-31. R e l a t i v e Nuclear C o n t r i butions o f Each Reactor Type i n LWR-FBR System w i t h L i g h t Plutonium "Transmutation" (High-Cost U308 Supply).

mately 620 kg o f f i s s i l e plutonium per GWe must be handled i n t h i s case, as compared t o approxi-

mately 170 kg o f f i s s i l e plutonium i n f r e s h f u e l per GWe each year f o r t h e case o f plutonium minimization and u t i l i z a t i o n . Thus, i t appears t h a t t h e a b i l i t y t o maintain an energy support r a t i o g r e a t e r than u n i t y w h i l e simultaneously adding 15 GWe/yr w i l l necessitate handling more

L

f i s s i l e plutonium i n f r e s h f u e l i n t h e energy centers. As pointed o u t ' i n previous cases, the a b i l i t y t o maintain a high energy support r a t i o requires t h e development o f a nuclear i n d u s t r y capable o f reprocessing*fuel containing thorium and r e f a b r i c a t i n g fuel containing 232U. loaded w i t h approximately 3% enriched

235U

I n t h i s o p t i o n i n the year 2035, the LWR comprises approximately 282 o f t h e i n s t a l l e d

capacity, t h e FBR comprises 48%, and the LWR loaded w i t h 11% 233U i n 238U comprises 24%. Upon examining t h e f l o w o f thorium and uranium metal associated w i t h these reactors, i t can be seen that 38% o f t h e reprocessing capacity must be capable o f handling f u e l cont a i n i n g thorium and 27% o f the f a b r i c a t i o n i n d u s t r y must be capable o f handling f u e l containing

2321).

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6-43

The annual consumption of U308 i n 2035 was found t o be approximately 32 ST U308/GWe.

i]

This consumption r a t e w i l l decrease continuously as the 235U-10aded LWR i s replaced w i t h t h e 233U-loaded LWR.

-

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t -

31.8 ST U308

L HEDL 780-090.32

U t i l i z a t i o n and Movement of F i s s i l e M a t e r i a l i n an LWR-FBR Nuclear Fig. 6.2-32. System w i t h a L i g h t " P u - ~ o - ~ ~ ~Transmutation UI' Rate (Case 6L, High-Cost U308 Supply) (Year 2035).

I n summary, a s t r a t e g y based on an FBR w i t h a Pu-U core and a thorium blanket could supply a n e t a d d i t i o n r a t e of 15 GWe/yr t o t h e year 2050 and beyond w i t h a U308 supply o f 3 m i l l i o n ST below $160/lb.

The i n s t a l l e d nuclear capacity i n 2050 would be 1100 GWe,

w i t h 560 GWe, o r approximately 50% o f t h e i n s t a l l e d capacity, l o c a t e d i n secure energy centers. Approximately 27% o f t h e f a b r i c a t i o n capacity must be capable o f handling f u e l containing 232U. Thus, w h i l e a nuclear system based on an FBR w i t h a Pu-U core and a

u I.

thorium blanket can supply 15 GWe/yr f o r an i n d e f i n i t e p e r i o d o f time, i t simultaneously r e q u i r e s t h a t a s i g n i f i c a n t amount o f nuclear capacity be l o c a t e d i n secure regions.


6-44

6.2.7.

Converter-Breeder System w i t h Heavy Plutonium "Transmutation"

The preceding discussion i n d i c a t e s t h a t a nuclear power system t h a t includes an FBR having a Pu-U core and producing 233U i n a thorium blanket can m a i n t a i n an energy support r a t i o g r e a t e r than unity w h i l e simultaneously adding 15 GWe/yr t o t h e i n s t a l l e d capacity throughout t h e planning horizon. The p o s s i b i l i t y e x i s t s , however, t h a t a nuclear power system t h a t includes an FBR having a Pu-Th core and a thorium blanket would r e s u l t i n a ~ ~ U r a t e which would maintain an energy support r a t i o s i g n i heavy P u - ~ o - ~ transmutation

f i c a n t l y g r e a t e r than u n i t y over t h e same p e r i o d o f time.

The p r i n c i p a l problem associated

w i t h a nuclear system based on an FBR w i t h a Pu-Th core i s t h a t t h e breeding r a t i o o f t h e breeder, and hence t h e breeding r a t i o o f t h e e n t i r e system, tends t o be low.

Therefore,

t h e e f f e c t o f adding t o t h e system an FBR operating on denatured 233U t o augment t h e production was a l s o investigated. system are shown i n Fig. 6.1-4

233U

The i n d i v i d u a l r e a c t o r concepts contained i n t h i s

(Option 8).

The nuclear c o n t r i b u t i o n associated w i t h t h i s o p t i o n (Case 8L, tvith denatured breeder) i s compared t o t h a t o f t h e LWR on t h e throwaway c y c l e f o r t h e high-cost U308 supply i n Fig. 6.2-33. The system i s capable o f maintaining a n e t a d d i t i o n r a t e of 15 GWe/yr throughout t h e planning horizon.

Id

L L ij

L L L

The i n s t a l l e d nuclear capacity which f o r Case 8L must be l o c a t e d i n energy centers i s shown i n Fig. 6.2-34 as a f u n c t i o n o f time. The maximum i s l e s s than 300 GWe througho u t t h e planning horizon.

The amount a v a i l a b l e f o r l o c a t i o n outside t h e energy centers

ranges from approximately 300 GWe i n t h e y e a r 2000 t o approximately 800 GWe i n t h e y e a r 2050. This can be compared t o Option 6 f o r which t h e nuclear capacity t h a t must be l o c a t e d i n secure regions increases continuously t o approximetely 560 GWe i n 2050. Thus, a nuclear system containing FBRs w i t h Pu-Th cores p l u s FBRs w i t h denatured

233U cores i s

capable o f maintaining a very high energy support r a t i o f o r an i n d e f i n i t e p e r i o d o f time. I t does require, however, t h a t reactors t h a t are n e t producers o f f i s s i l e m a t e r i a l be

l o c a t e d i n energy centers. The u t i l i z a t i o n and movement o f f i s s i l e m a t e r i a l i n year 2035 f o r Case 8L and t h e small U308 supply are shown i n Fig. 6.2-35.

The LWR loaded w i t h approximately 3% enriched

235U comprises approximately 13% o f t h e i n s t a l l e d capacity, t h e denatured 235U LWR comprises 233U LWR comprises 8%, and t h e denatured FBR comprises 38%. The denatured 235U LWR i s being r a p i d l y phases o u t o f t h e nuclear system i n year 2035, w h i l e t h e denatured 233U LWR i s being

L;

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approximately 12%, t h e energy c e n t e r FBR comprises approximately 29%, t h e denatured

r a p i d l y phased in.

This i s i n d i c a t e d i n Fig. 6.2-35

by t h e f a c t t h a t t h e heavy metal d i s -

charge f o r t h e denatured 235U LWR i s considerably g r e a t e r than t h e heavy metal charge, w h i l e t h e heavy metal charge f o r t h e denatured 233U LWR i s considerably g r e a t e r than t h e heavy metal discharge. The former i s i n d i c a t i v e o f f i n a l core discharges, w h i l e t h e l a t t e r

i s i n d i c a t i v e of f i r s t core loadings.

L f'

L 6f ' '

w


6-45

1mO

I

I

I

I

1

CASE BL -THE FBR WITH M A W PLUTONIUM IUNSMVTATION

ImO

I

W E 8L

I

1

I

I

- IHE FER WITH HEAW PLUTONIUM TUMMUTATlON

/f

/

L IC

YEAR

MAR

Fig. 6.2-33. The Nuclear Contributions o f an LWR-FBR System w i t h Heavy Plutonium "Transmutation" (High-Cost U308 Supply).

I

t'; Id I

Fiq. 6.2-34. R e l a t i v e Contributions of Reaciors Located I n s i d e (Pu-Fueled) and Outside (Denatured LWRs and FBRs) Energy Centers (High-Cost U308 Supply).

I n t h i s o p t i o n t h e annual consumption o f U308 i s approximately 25 ST U308 i n year 2035, decreasing t h e r e a f t e r as the LWRs loaded w i t h 235U a r e replaced by t h e LWRs loaded w i t h 233U.

Approximately 430 kg o f f i s s i l e plutonium per GWe o f i n s t a l l e d capacity must

be handled as f r e s h f u e l each year w i t h i n energy centers, somewhat l e s s than the 620 kg t h a t must be handled i n Option 6. The a b i l i t y t o maintain a h i g h energy support r a t i o w h i l e simultaneously adding 15 GWeIyr again r e q u i r e s t h e development o f a nuclear i n d u s t r y capable o f reprocessing f u e l containing thorium and r e f a b r i c a t i n g f u e l containing 232U. Figure 6.2-35 shows t h a t 65% of t h e reprocessing capacity i n year 2025 must be capable o f handling f u e l c o n t a i n i n g thorium and t h a t 31% o f t h e r e f a b r i c a t i o n capacity must be capable o f handling f u e l containing

232U.

The e f f e c t o f d e l e t i n g the denatured FBR from t h e system i s shown i n Figs. 6.2-36 and 6.2-37. Figure 6.2-36 shows t h a t w i t h o u t the denatured FBR t h e i n s t a l l e d nuclear c a p a c i t y reaches a maximum o f approximately 840 GWe i n about 2035 and declines continuously thereafter.

The reason f o r t h i s , o f course, i s t h a t w i t h o u t t h e denatured FBR t h e system has

a n e t breeding r a t i o o f l e s s than u n i t y .

I;

Therefore, w h i l e t h e system can m u l t i p l y t h e f i s s i l e supply s i g n i f i c a n t l y , i t cannot continue t o grow i n d e f i n i t e l y . The nuclear capacity t h a t must be l o c a t e d i n energy centers f o r t h e modified Case 8L i s shown i n Fig. 6.2-37.

This c a p a c i t y does n o t exceed 140 GWe throughout t h e planning horizon.

The amount o f capacity a v a i l a b l e f o r l o c a t i o n outside t h e secure regions ranges from approximately 300 GWe i n t h e year 2000 t o approximately 700 GWe i n year 2035.

2: f Âś .-

E

I n summary, a s t r a t e g y based on an FBR w i t h a Pu-Th core and a thorium blanket can supply a n e t a d d i t i o n r a t e o f 15 GWe/yr t o year 2050 and beyond provided a denatured breeder i s included i n t h e system.

I f t h e denatured breeder i s n o t included, then the maximum

nuclear c o n t r i b u t i o n would be approximately 840 GWe.

The amount o f nuclear capacity t h a t

must be l o c a t e d i n secure regions does n o t exceed 140 GWe i n t h i s case.


6-46

19 Kg fir Pu

0.38 GWe CFe.2 I 180 Kg fir Pu 304 Ka Urn 3758 K; Th 8309 Kg HM

swu

L

25.2 ST

4

13 Kg fir Pu

3589 Kg Th 4522 Kg HM

0.12 GWe CFd.3

2309 Kg Th 2917 Kg HM

HEDL 786490.29

Fig. 6.2-35. Utilization and Movement o f F i s s i l e Material i n an LWR-FBR Nuclear System w i t h Heavy "Pu-t0-*33U". Transmutation Rate (Case 8L, High-Cost U308 Supply) (Year 2035).

1WI

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I

I

I

I

I

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I

I

I

I

I

I

I

T H i F B I WITH HEAW WJTONIIYH TPAMMUIATION AND WITHOVT A DTMIURED B l E C f I

YO,

SUPPLY

- 3.0 . 18

IT

DENATUED BEEDEI

L

L

Q I

F B I WITH HUW

c c

-

L Yâ&#x201A;ŹAR

YEAR

Fig. 6.2-36. Effect on Nuclear ContriF i g . 6.2-37. Relative Nuclear Contrib u t i o n o f Eliminating Denatured Breeder from butions o f Reactors Located Inside (Pu-Fueled) LWR-FBR System w i t h Heavy Plutonium "Transand Outside (Denatured LWRs) Energy Centers mutation." (Case 8L Minus Denatured Breeder) (Case 8L Minus Denatured Breeder) (High-Cost (High-Cost U308 Supply). u 308 supply)

L L


6-47

c

6.3.

CONCLUSIONS

The p r i n c i p a l conclusions developed during t h e course of t h i s study are summarized

i n Tables 6.3-1,

6.3-2,

and 6.3.3.

From t h e preceding discussion and Table 6.3-1,

t h e f o l l o w i n g conclusions are drawn

f o r various nuclear systems operating on t h e throwaway cycle:

(1) With a U308 supply o f 3.0 m i l l i o n ST below $160/lb, t h e maximum i n s t a l l e d capacity w i t h t h e standard LWR on t h e throwaway c y c l e would be approximately 420 GWe, and t h i s would occur i n about year 2006.

z

( 2 ) A reduction i n the U308 requirement o f a l l LWRs commencing operation i n 1981 and t h e r e a f t e r by 6% would n o t s i g n i f i c a n t l y increase t h e maximum i n s t a l l e d capacity. Thus, f o r t h e case of t h e LWR on t h e throwaway cycle, t h e e f f o r t should be on improvements i n U3O8 u t i l i z a t i o n s i g n i f i c a n t l y g r e a t e r than 6% f o r LWRs commencing operation a f t e r 1981 o r on improvements which can be r e t r o f i t t e d i n t o e x i s t i n g LWRs.

Table 6.3-1. Summary o f Results f o r Nuclear Power Systems Operating on t h e Throwaway/Stowaway Cycle

Option

Technology Development Requirement

Maximum Nuclear Contribution (GWe 1

Year o f Maximum Contribution

High-Cost U308 supply Standard LWR Improved LWR LWR p l u s advanced converter LWR w i t h improved t a i l s composition

None LWR w i t h extended d i s charge exposure SSCR, HTGR, o r HWR

420 430

2006 2010

450

2012*

Advanced enrichment process

500

2015

Intermediate-Cos t u308 Supply Standard LWR

Successful U308 explorat i o n program

730

2030

LWR p l u s advanced Converter

SSCR, HTGR, o r HWR; a l s o successful U 308 expl o r a t i o n program

850

2035

*Depends on advanced converter concept and i t s i n t r o d u c t i o n date.


6-48

Table 6.3-2.

Summary o f Results f o r Nuclear Power Systems U t i l i z i n g LWR Converters w i t h and w i t h o u t FBRs ( w i t h Recycle)

Option

Technology Development Requirement

7-

Maximum Nuclear C o n t r i b u t i o n Total F r a c t i o ? o f GWE Year Gb!e i n Energy Center

High-Cost U308 Supply Pu r e c y c l e (2L)

Reprocessing , r e f a b r i c a t ion

-2020

600

Pu throwaway (4L)

Advanced f u e l design, reprocessing

-2020

590

-

Pu p r o d u c t i o n minimized. Pu-to-2'3U "transmutation" (5UL)

Advanced f u e l design, reprocessing

~2030

700

0.15

Pu p r o d u c t i o n n o t minimized, Pu-to-z.J3U "transmutation" (5TL)

Advanced f u e l design, reprocessing

-2025

640

0.21

FBRs added, li h t Pu transmutation 76L)

Advanced f u e l design, reprocessing, FBR

%2050

>1100

-0.56*

FBRs added, heavy Pu transmutation (7L)

Advanced f u e l design, reprocessing, FBR

.2050

>1100 ( w i t h denat. FBR)

-0.27*

~2035

850 (w/o denat. FBR)

4.16

960

-

-0.40

(w/o denat. FBR)

Intermediate-Cost U308 Supply Pu r e c y c l e (2L) Pu throwaway (4L)

Reprocessing, r e f a b r i c a t i o n Advanced f u e l design, reprocessing

~2045 -2045

Pu p r o d u c t i o n minimized, Pu-to-233U "transmutat i o n " (5UL)

Advanced f u e l design, reprocessing

Pu p r o d u c t i o n n o t minimized, P u - ~ o - * ~ ~"transmutation" U (5TL)

Advanced f u e l design, reprocessing

*

980

-

>2050

>loo0

-

-2050

1020

-

L

L ii

L

I n y e a r 2050.

(3) The deployment o f an advanced converter beginning i n 1995 w i l l n o t s i g n i f i c a n t l y increase t h e maximum i n s t a l l e d capacity i f t h e U308 supply i s l i m i t e d t o 3.0 m i l l i o n ST below $160/lb.

This i s p r i m a r i l y due t o t h e f a c t t h a t a s i g n i f i c a n t amount

o f t h e U308 supply has been committed t o t h e standard LWR p r i o r t o t h e advanced converter a t t a i n i n g a l a r g e f r a c t i o n o f t h e i n s t a l l e d capacity. If t h e U308 supply should be as l a r g e as 6.0 m i l l i o n ST below $160/lb, considerably l a r g e r .

(4)

then t h e e f f e c t o f t h e advanced converter i s

An advanced enrichment process capable o f economically reducing t h e t a i l s compo-

s i t i o n t o 0.0005 c o u l d have a g r e a t e r e f f e c t than improvements i n LWR U308 u t i l i z a t i o n or t h e deployment o f an advanced converter.

L


6-49

Sumnary o f Fuel Cycle Reauirements f o r Nuclear Power Table 6.3-3. Systems U t i 1 i z i n g LWR Converters w i t h and w i t h o u t FBRs ( w i t h Recycle; High-Cost U308 Supply)

Option

Fraction o f I n s t a l l e d Nuclear Capacity Permitted Outside Energy Center i n Year 2025

Fraction o f Refabrication Capacity t o Handle 232U i n Year 2035

Pu recycle

0.61

0

0

Pu throwaway Pu production minimized; Pu-t0-23~U "transmutation" Pu production not minimized; Pu-to-23W "transmutation" FBRs added. l i g h t Pu transmutation

1 .oo

0.95

0.57

0.85

0.97

0.53

0.79

0.34

0.20

0.56

0.38

0.27

0.76

0.65

0.31

FBRs added, heavy Pu transmutation

c

Fraction o f Reprocessing Capacity t o Handle Th i n Year 2035

(5) The e f f e c t o f an e x p l o r a t i o n program successful enough t o r e l i a b l y increase t h e U308 resource base t o 6.0 m i l l i o n ST below $160/lb would be considerably g r e a t e r than any o f t h e above.

Thus, when analyzing t h e throwaway option, t h e s i z e o f the U308 resource base and t h e u n c e r t a i n t y associated w i t h i t dominate t h e analysis. From t h e discussion i n Section 6.2 and Tables 6.3-2

and 6.3-3,

t h e f o l l o w i n g conclu-

sions are drawn f o r LWR and LWR-FBR systems operating w i t h recycle: (1) With t h e high-cost U308 supply, t h e e f f e c t of plutonium r e c y c l e i n LWRs would be t o increase t h e i n s t a l l e d nuclear capacity t o 600 GWe, and t h i s would occur i n about year 2020.

T h i s would require, however, t h a t as much as 40% o f t h e nuclear capacity be

l o c a t e d i n t h e energy centers.

I f t h e U308 supply should be as l a r g e as 6.0 m i l l i o n ST

below $160/lb, t h e maximum i n s t a l l e d nuclear capacity would be 960 GWe, and t h i s would occur i n about year 2045.

G

(2) I f a l l plutonium were thrown away b u t f i s s i l e uranium were r e f a b r i c a t e d and reloaded, t h e maximum i n s t a l l e d nuclear capacity could be as l a r g e as 590 GWe w i t h t h e

high-cost U308 supply.

A t t a i n i n g 590 GWe, however, requires t h e development o f f u e l I n a d d i t i o n , i t requires t h e designs which minimize t h e amount o f plutonium produced. development o f an i n d u s t r y i n which as much as 95% o f t h e reprocessing capacity i s devoted t o f u e l c o n t a i n i n g thorium and as much as 57% of t h e r e f a b r i c a t i o n capacity i s devoted t o f u e l c o n t a i n i n g 2321.1. (3) I f t h e plutonium produced i n t h e system described immediately above were ref a b r i c a t e d and reloaded, t h e maximum i n s t a l l e d nuclear capacity would increase t o approxi-

L

mately 700 GWe, which i s an increase i n t h e maximum o f approximately 110 GWe.


6-50

(4) I f a l l plutonium produced were transmuted t o 2 3 3 U b u t no attempt was made t o minimize t h e amount o f plutonium produced, the maximum i n s t a l l e d nuclear capacity could be as l a r g e as 640 GWe w i t h t h e high-cost U30a supply.

As much as 21% o f the i n s t a l l e d

nuclear capacity would have t o be l o c a t e d i n secure energy centers, however, and i t would r e q u i r e t h a t 34% o f the reprocessing capacity be devoted t o f u e l c o n t a i n i n g thorium and 20% o f t h e r e f a b r i c a t i o n capacity be devoted t o f u e l containing 233U. (5) I f a nuclear system u t i l i z i n g an FBR w i t h a Pu-U core and a thorium b l a n k e t were developed, t h e system c o u l d maintain a n e t a d d i t i o n r a t e o f 15 GWe/yr i n d e f i n i t e l y . The i n s t a l l e d nuclear capacity, i n t h i s case, could be as high as 1100 GWe i n y e a r 2050; however, 56% o f t h i s capacity would have t o be l o c a t e d i n secure energy centers.

Also, approximately

38% o f t h e reprocessing capacity would have t o be devoted t o f u e l containing thorium and 27% o f t h e r e f a b r i c a t i o n capacity would have t o be devoted t o f u e l c o n t a i n i n g 232U. (6) I f a nuclear system u t i l i z i n g an FBR w i t h a Pu-Th core and a thorium b l a n k e t were developed, t h e maximum i n s t a l l e d capacity would depend upon t h e performance c h a r a c t e r i s t i c s o f the denatured design r e c e i v i n g f u e l from t h e FBR.

I f t h i s design were a denatured breeder,

t h e nuclear system would be capable o f adding 15 GWe/yr i n d e f i n i t e l y . If, however, the design were a denatured LWR, then t h e i n s t a l l e d nuclear capacity would increase t o approximately 850 GWe i n about year 2035 and decrease t h e r e a f t e r . I n a d d i t i o n t o t h e r e s u l t s and conclusions presented i n t h i s chapter, d e t a i l e d r e s u l t s f o r a l l the nuclear p o l i c y options c a l c u l a t e d are t a b u l a t e d i n Appendix C. Also, as ment i o n e d e a r l i e r , a separate a n a l y s i s performed under t h e assumption o f an u n l i m i t e d U308

L

L L L L L G

supply b u t w i t h the nuclear power systems i n competition w i t h c o a l - f i r e d p l a n t s i s described i n Appendix D.

1;:

Chapter 6 References

1.

R. D. Nininger, "Remarks on Uranium Resources and Supply," I n d u s t r i a l Forum, New York, March 7, 1978.

2.

John Klemenic, D i r e c t o r , Supply Analysis D i v i s i o n , Grand Junction O f f i c e , DOE Uranium and Enrichment Division, "Production Capabi 1it y ,'I October 1978,

3.

John Klemenic and David Blanchfield, Mineral Economist, Grand Junction O f f i c e , "Product i o n C a p a b i l i t y and Supply," paper presented a t Uranium I n d u s t r y Seminar, October 26-27, 1977, Grand Junction, Colorado; proceedings pub1ished as GJO-108( 77).

4.

Fuel Cycle 78, Atomic

"Uranium Enrichment Services A c t i v i t y Financial Statements f o r Period Ending September

30, 1977," p. 13, Schedule C, 0R0-759. 5.

"AEC Gaseous D i f f u s i o n P l a n t Operations," 0R0-684, USAEC (January 1972).

6.

"Data on New Gaseous D i f f u s i o n Plants,"

7.

See also, T. M. Helm, M. R. Shay, R. W. Hardie, and R. P. Omberg, "Reactor Design C h a r a c t e r i s t i c s and Fuel Inventory Data," TC-971 , Hanford Engineering Development Laboratory (September 1977).

0R0-685, USAEC ( A p r i l 1972).

L L I:


CHAPTER 7 OVERALL ANALYSIS OF DENATURED FUEL SYSTEMS

rid

Chapter Out1 i n e

L

7.0.

Introduction,

7.1.

P r o l i f e r a t i o n - R e s i s t a n t C h a r a c t e r i s t i c s o f Denatured

T . J. B u r n s , ORNL

233U

Fuel ,

C. M . N e w s t e a d , BNL, a n d T . J. B u r n s , ORNL

7.1.1. 7.1.2. 7.1.3. 7.1.4.

L

I s o t o p i c B a r r i e r o f Fresh Fuel Gamma-Radiation B a r r i e r o f Fresh Fuel Spent Fuel F i s s i l e Content Conclusions

-.

i

7.2.

Y

Impact o f Denatured 2 3 3 U Fuel on Reactor Performance and Selection: w i t h Other Fuel Cycles, T . J. B u r n s , ORNL

L

7.2.1. 7.2.2. 7.2.3. 7.2.4.

r'

7.3.

L

Comparison

Thermal Reactors Fast Reactors Symbiotic Reactor Systems Conclusions

Prospects f o r Implementation and Commercialization o f Denatured 2 3 3 U Fuel Cycle, J. C . C l e v e l a n d a n d T . J. B u r n s , ORNL

7.3.1. 7.3.2. 7.3.3. 7.4.

Possible Procedure f o r Implementing and Commercializing t h e Denatured Fuel Cycle Considerations i n Commercializing Reactors Operating on A1 t e r n a t e Fuels Conclusions

Adequacy o f Nuclear Power Systems U t i l i z i n g Denatured 233U Fuel f o r Meeting E l e c t r i c a l Power Demands, M . R . S h a y , D . R . H a f f n e r , w. E . B l a c k , T . M . H e l m , R . W . Hardie, a n d R . P . O m b e r g , HEDL

. -

u

A -

7.4.1. 7.4.2. 7.4.3. 7.4.4. 7.4.5. 7.4.6.

7.5.

The A n a l y t i c a l Method Data Base Results f o r Price-Limited Uranium Supplies Results f o r Unconstrained Resource A v a i l a b i l i t y Systems Employing Improved LWRs and Enrichment Technology Conclusions

Tradeoff Analysis and Overall Strategy Considerations, ORNL

7.5.1. 7.5.2. 7.5.3.

No-Recycle Options Recycle Options O v e r a l l Conclusions and Recommendations

T . J. B u r n s a n d I. S p i e w a k ,


c


7-3 I

d 7.0.

r--

1

> w

INTRODUCTION

T. J. Burns Oak Ridge National Laboratory

I

r-

k

The assessment o f any proposed f u e l c y c l e must o f necessity consider various t o p i c s t h a t a f f e c t the f e a s i b i l i t y and v i a b i l i t y o f the p a r t i c u l a r cycle.

Moreover, an assessment

o f a p a r t i c u l a r f u e l c y c l e must consider t h e r e l a t i v e m e r i t s o f t h e f u e l c y c l e compared t o other p o t e n t i a l l y a v a i l a b l e f u e l c y c l e options.

This study o f t h e denatured 2 3 3 U f u e l c y c l e

has addressed various aspects o f the c y c l e i n the preceding chapters:

the p r o l i f e r a t i o n -

r e s i s t a n t c h a r a c t e r i s t i c s o f the c y c l e ( i n Chapter 3 ) ; the impact o f denatured 2 3 3 U f u e l on t h e performance o f several types o f r e a c t o r s ( i n Chapter 4); the implementation and commercial i z a t i o n aspects o f t h e denatured f u e l c y c l e ( i n Chapter 5) ; and the economic/resource i m p l i c a t i o n s o f t h e c y c l e ( i n Chapter 6).

I n each o f these chapters, t h e assessment o f t h e

denatured 233U c y c l e was l i m i t e d p r i m a r i l y t o the s p e c i f i c aspect under consideration.

In

t h i s chapter t h e d e t a i l e d r e s u l t s o f the assessment are summarized and integrated, and the p o t e n t i a l t r a d e o f f s possible between t h e various considerations are addressed.

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L id

.

r-

t

E

I n addition,

recommendations f o r f u r t h e r study o f c r u c i a l aspects o f the denatured 2 3 3 U f u e l c y c l e are

.

made.


7-4

7.1.

PROLIFERATION-RESISTANT CHARACTERISTICS OF DENATURED 233U FUEL

i;

C. M. Newstead Brookhaven National Laboratory and

T. J. Burns Oak Ridge National Laboratory

As has been s t a t e d i n e a r l i e r chapters, t h e primary goal o f t h e denatured f u e l c y c l e i s t o permit t h e r e c y c l e o f f i s s i l e f u e l s i n dispersed reactors i n a manner c o n s i s t e n t w i t h n o n p r o l i f e r a t i o n considerations. I n t h i s s e c t i o n t h e p r o l i f e r a t i o n - r e s i s t a n t characteri s t i c s o f t h e denatured 2 3 % f u e l c y c l e that have been described i n d e t a i l i n Chapter 3 are summarized, and t h e i r s i g n i f i c a n c e w i t h respect t o both n a t i o n a l p r o l i f e r a t i o n and subnational t e r r o r i s m i s noted. I n general, these c h a r a c t e r i s t i c s d e r i v e from t h r e e d i s t i n g u i s h i n g features o f t h e denatured f u e l cycle: (1) t h e i n t r i n s i c i s o t o p i c b a r r i e r o f t h e f r e s h denatured f u e l , (2) t h e gamma r a d i a t i o n b a r r i e r associated with t h e 232U i m p u r i t y present i n thorium-derived f u e l , and (3) the low chemically separable f i s s i l e content o f t h e spent denatured f u e l . 7.1.1.

2%

L L

I s o t o p i c B a r r i e r o f Fresh Fuel

The i s o t o p i c b a r r i e r o f t h e f r e s h f u e l i s created by t h e a d d i t i o n o f t h e denaturant t o the

L

238U

f i s s i l e f u e l , i t s purpose being t o preclude t h e use of t h e 233U

d i r e c t l y i n a nuclear weapons program. Although the thorium present i n most proposed denatured f u e l s could be chemically removed, t h e sepqrated uranium would have t o o low a f i s s i l e content f o r i t t o be d i r e c t l y usable i n a p r a c t i c a l nuclear device.

By contrast,

L

t h e o t h e r p o t e n t i a l f u e l c y c l e r e l y i n g on recycled m a t e r i a l , t h e Pu/U cycle, would r e q u i r e o n l y a chemical separation t o e x t r a c t weapons-usable m a t e r i a l d i r e c t l y from power r e a c t o r f u e l . The i s o t o p i c b a r r i e r in. denatured f u e l i s n o t an absolute b a r r i e r , however, since any isotope separation (i .e.,

enrichment) technique can be used t o circumvent it.

Depending upon i t s technological resources, a n a t i o n may have o r may develop separation facilities.

On t h e other hand, i t i s u n l i k e l y t h a t a subnational group would possess

i s o t o p i c separation c a p a b i l i t i e s and thus t h e i s o t o p i c b a r r i e r i n h e r e n t i n denatured f u e l

f.â&#x20AC;&#x2122;

would provide considerable p r o t e c t i o n against t e r r o r i s t nuclear a c t i v i t i e s . As i s pointed o u t i n Section 3.3.4

and Appendix A, enrichment technology has made

g r e a t s t r i d e s i n recent years and i s p r e s e n t l y undergoing r a p i d f u r t h e r development.

Ten

years ago t h e o n l y operational enrichment f a c i l i t i e s were based on t h e gaseous d i f f u s i o n technique, a method r e q u i r i n g a l a r g e expenditure o f energy and a l a r g e p l a n t t o be economic.

Today t h e gas c e n t r i f u g a t i o n technique, which requires a s i g n i f i c a n t l y lower

energy consumption than t h e gaseous d i f f u s i o n method, i s a v a i l a b l e and i s p r a c t i c a l w i t h small-scale plants. For example, t h e URENCO consortium i s c u r r e n t l y operating c e n t r i fuge enrichment p l a n t s o f 50 tonnes per year capacity a t Capenhurst i n t h e United Kingdom and a t Almelo i n The Netherlands.

The URENCO c e n t r i f u g e represents an economic

design b u i l t by t e c h n o l o g i c a l l y advanced countries (England, The Netherlands, Germany)

I;

L


7-5

L 1

w i t h o u t b e n e f i t o f U.S.

experience.

For a m i l i t a r y program, economics would n o t be an

o v e r r i d i n g c r i t e r i o n and could be s a c r i f i c e d i n f a v o r o f a more moderate l e v e l o f technology.

Moreover, the open l i t e r a t u r e contains s u f f i c i e n t i n f o r m a t i o n concerning t h e

c e n t r i f u g e designs t o guide mechanically competent engineers w i t h access t o adequate facilities.

R e p l i c a t i o n o f an economic design would r e q u i r e a somewhat higher l e v e l o f

technology than prototype construction. The f o l l o w i n g p a r t i c u l a r p o i n t s regarding t h e enrichment o f denatured

233U

fuel

should be noted: Because of t h e lower mass o f 233U, separating 2 3 % from 238U would r e q u i r e o n l y 9/25 o f the e f f o r t r e q u i r e d t o separate 2 3 5 U from 238U, assuming equal feed enrichments. Since the f a s t c r i t i c a l mass o f 2 3 % i s l e s s than t h a t o f 235U, l e s s enrichment capacity would be r e q u i r e d t o produce a 2 3 % weapon from 233U/238U feed than would be r e q u i r e d t o produce a 235U weapon from 235Uf238U feed, again assuming equal enrichments o f t h e feed m a t e r i a l . The higher t h e enrichment o f the source m a t e r i a l , the l e s s separative work t h a t would

L

have t o be done t o upgrade t h e m a t e r i a l t o 90% enrichment. For example, e n r i c h i n g n a t u r a l uranium t o a 10% l e v e l consumes 90% o f the separative work required t o achieve a 90% l e v e l .

It i s t o be noted t h a t the enrichment o f denatured

23% fuel

i s approximately 12%, whereas the enrichment o f c u r r e n t l y used LWR 235U f u e l i s around 3-4%. With respect t o items (2) and (3), a rough comparison can be made o f the feed requirements and t h e number o f c e n t r i f u g e s t h a t would be necessary t o produce 90% enriched m a t e r i a l from various f u e l s i n one year (normalized t o 1 kg o f product): Number of Centrifuges Required Feed Required

12% 2 3 3 u

8

0.3 kg SWUfyr Capacity 55

20% 2351)

5

50

3

30 178

292 779

17 46

Fuel

u

(kg)

3.2% 235U Natural Uranium

5 kg SWUfyr Capacity 3

The above values do n o t consider measures t o e l i m i n a t e t h e 232U contamination and they assume t h a t a reasonable t a i l s assay w i l l be maintained (%0.2% 235U). I f a higher t a i l s assay were acceptable, t h e number of centrifuges could be reduced b u t t h e feed m a t e r i a l r e q u i r e d would be increased. One year, o f course, i s a long time when compared t o a p e r i o d o f weeks t h a t would be needed t o o b t a i n approximately 10 kg o f plutonium by chemically reprocessing two t o three spent LWR-LEU fuel elements.

I t would be possible t o speed up the process time f o r

t h e c e n t r i f u g e method e i t h e r by increasing t h e i n d i v i d u a l machine capacity, by adding a d d i t i o n a l centrifuges,

o r by operating a t a higher t a i l s assay.

Increasing t h e capacity


7-6

would be q u i t e d i f f i c u l t and would r e q u i r e increasing technological s o p h i s t i c a t i o n ; however, adding c e n t r i f u g e s would r e q u i r e only t h a t t h e same device be d u p l i c a t e d as many times as necessary. Increasing t h e t a i l s assay would r e q u i r e more feed m a t e r i a l . F i n a l l y , i n considering t h e p o t e n t i a l circumvention o f t h e i s o t o p i c b a r r i e r , i t i s important t o a n t i c i p a t e t h e enrichment technologies t h a t could e x i s t i n 20 t o 25 years

1

I I

t h e time when t h e denatured f u e l c y c l e could be deployed.

-

Technologically advanced

countries already have the necessary technological base t o design and c o n s t r u c t c e n t r i fuges, and many p r e s e n t l y developing countries may have acquired t h e technology base by I

t h a t time.

1

Countries w i t h a p r i m i t i v e technology are u n l i k e l y t o use t h i s route, since

even with the f i n a n c i a l assets and t e c h n i c a l l y competent personnel they would have t h e

.

d i f f i c u l t task o f developing t h e r e q u i s i t e support f a c i l i t i e s .

Other p o t e n t i a l isotope

separation techniques are under development i n many countries.

Laser isotope separation

(LIS), plasma techniques, aerodynamic methods, chemical techniques, and electromagnetic separation methods c u r r e n t l y show varying degrees o f promise. these methods i s discussed i n Appendix A.

, I

I

The c u r r e n t s t a t u s o f

It i s impossible t o p r e d i c t the u l t i m a t e

success o r f a i l u r e o f these a l t e r n a t i v e methods, and hence t h e i s o t o p i c separation c a p a b i l i t y which might e x i s t i n 25 years i s even more d i f f i c u l t t o estimate. Current estimates f o r t h e U.S. development program i n LIS and plasma methods suggest t h a t i t w i l l

I

be a t l e a s t t e n years before such methods could be operative on a working i n d u s t r i a l

I

basis, even w i t h a h i g h l y s o p h i s t i c a t e d R&O e f f o r t .

7.1.2. The production of

Gamma-Radiation B a r r i e r o f Fresh Fuel

233U

r e s u l t s i n t h e concomitant production o f a small b u t radio-

a c t i v e l y s i g n i f i c a n t q 2 a n t i t y o f 232U through the 232Th(n,2n)

r e a c t i o n [and t h e 230Th(n,y)

r e a c t i o n i f 230Th i s present i n t h e thorium].

As the 232U decays through 228Th and i t s daughter products, t h e gama a c t i v i t y o f t h e 233U-containing f u e l s increases, thus p r o v i d i n g a r a d i a t i o n b a r r i e r much more intense than i s found i n o t h e r f r e s h f u e l s . While chemical processing

c o u l d be employed t o remove t h e

23211

decay products, such a procedure would p r o v i d e a r e l a t i v e l y

low r a d i o a c t i v i t y f o r o n l y 10-20 days, since f u r t h e r decay o f t h e 232U present i n t h e f u e l would provide a new population o f 228Th and i t s daughters, the a c t i v i t y o f which would cont i n u e t o increase i n i n t e n s i t y f o r several years. The concentration o f

232U

i n t h e r e c y c l e f u e l i s u s u a l l y characterized as so many

p a r t s per m i l l i o n (ppm) o f 2 s U i n t o t a l uranium. 232Th(n,2n)

reaction, t h e

232u

Due t o t h e threshold nature of the

concentration v a r i e s w i t h t h e neutron spectrum o f t h e I t a l s o v a r i e s w i t h t h e amount of recycle. For 12%

r e a c t o r i n which i t i s produced. 23%

denatured fuel, t h e

produced

233U

Section 3.1.3).

232U

concentration ( i n ppm U) ranges from 250 ppm f o r LWR-

t o a maximum o f 1600 ppm f o r c e r t a i n LMFER-derived denatured f u e l s (see I f the l a t t e r m a t e r i a l were enriched t o produce weapons-grade material,

t h e 232U concentration would be approximately 8000 ppm, and thus t h e m a t e r i a l would be highly radioactive.


7-7

While t h e r a d i a t i o n f i e l d would introduce complications i n t h e manufacture o f a weapon, p a r t i c u l a r l y f o r a t e r r o r i s t group, t h e r e s u l t i n g dose r a t e s would n o t provide an absolute b a r r i e r (see Section 3.3.5). As mentioned above, i t would be p o s s i b l e t o clean up t h e f i s s i l e m a t e r i a l so t h a t i t was r e l a t i v e l y free of r a d i a t i o n f o r a p e r i o d o f 10 t o

20 days. A l t e r n a t i v e l y , p r o v i d i n g s h i e l d i n g and remote handling would a l l o w t h e r a d i a t i o n b a r r i e r t o be circumvented; however, c o n s t r u c t i o n and/or a c q u i s i t i o n o f t h e shielding, remote handling equipment, etc.,

could increase t h e r i s k o f d e t e c t i o n o f a c o v e r t pro-

gram before i t s completion. N o n - f i s s i l e m a t e r i a l i n c l u d e d i n t h e weapon would a l s o provide some s h i e l d i n g d u r i n g d e l i v e r y , and a d d i t i o n a l shadow s h i e l d i n g t o p r o t e c t the operator o f t h e d e l i v e r y v e h i c l e and t o f a c i l i t a t e the loading operations could be devel oped.

L

I n another approach, t h e 232U could be separated from t h e 233U by i n v e s t i n g i n a r a t h e r l a r g e cascade o f over some 3000 centrifuges, p o s s i b l y i n c l u d i n g 228Th cleanup t o l i m i t t h e r a d i a t i o n contamination o f t h e centrifuges.

A w i l l i n g n e s s t o accept c e r t a i n operational disadvantages would permit t h e radiation-contaminated m a t e r i a l t o be processed i n t h e cent r i f u g e s provided they were shielded and some p r o v i s i o n was made f o r remote operation.

a,

and t h e c u r r e n t l y employed f r e s h LEU f u e l would have e s s e n t i a l l y none a t a l l . 7.1.3.

-

L i

--

k;

By

comparison, clean mixed oxide Pu/U f u e l would have a much l e s s s i g n i f i c a n t rad.iation problem

Spent Fuel F i s s i l e Content

Spent denatured f u e l contains t h r e e possible sources o f f i s s i l e m a t e r i a l : unburned 233U; 2 3 P a which decays t o

23S;and

Pu produced from t h e 238U denaturant.

Use o f t h e

uranium contained i n t h e spent denatured f u e l i s s u b j e c t t o a l l t h e considerations outl i n e d above and would a l s o be hindered by the fission-product contamination (and r e s u l t a n t r a d i a t i o n ) i n h e r e n t i n spent r e a c t o r f u e l . As was noted i n Section 3.3.4, t h e r e l a t i v e l y long h a l f - l i f e o f 233Pa (27.4 days) could permit the production o f weapons-grade m a t e r i a l v i a chemical separation o f t h e 2j3pa; however, such a procedure would r e q u i r e t h a t chemical separation be i n i t i a t e d s h o r t l y upon discharge from t h e r e a c t o r (while r a d i a t i o n l e v e l s are very high) t o minimize the amount of 233Pa which decays t o 2 3 3 U w h i l e s t i l l contained i n t h e 238U denaturant. Moreover, since the discharge concentration o f 233Pa i s t y p i c a l l y 5% o f t h a t o f

233U,

a considerable heavy metal processing r a t e would be r e q u i r e d

t o recover a s i g n i f i c a n t q u a n t i t y o f 233Pa (and hence 23%) w i t h i n t h e time frame a v a i l able.

The plutonium concentration i s comparable t o t h a t o f 23?a,

by decay.

b u t very l i t t l e i s l o s t

Hence, t h e spent f u e l can be allowed t o cool f o r some time before reprocessing.

It would seem, therefore, t h a t i f denatured 233U spent f u e l were d i v e r t e d i t would be

L

p r i m a r i l y f o r i t s plutonium content. Any f u e l c y c l e u t i l i z i n g 238U i n e v i t a b l y leads t o some plutonium production.

t IJ L

Compared t o t h e LEU c y c l e and t h e Pu/U cycle, the denatured 233U f u e l c y c l e reduces t h e plutonium production by (1) employing as l i t t l e 238U as necessary t o achieve t h e denatur ng objective, and ( 2 ) r e p l a c i n g the displaced 238U w i t h 232Th t o enhance the I


7-8

production o f "denaturable" 3%. The plutonium production rates for various reactors operating on conventional and denatured f u e l cycles are discussed i n Chapter 4 and summarized i n Table 7.1-1, where the Light-Water Reactor (LWR) i s represented by the pressurized-water reactor (PWR); the SSCR (Spectral-Shift-Control l e d Reactor) i s a modified PWR; the heavy-water reactor (HWR) i s assumed t o be a s l i g h t l y enriched CANDU; the High-Temperature Gas-Cooled Reactor (HTGR) i s taken t o be the F o r t S t . Vrain plant; and the High-Temperature Reactor (HTR) o f the Pebble-Bed Reactor (PBR) type i s represented by the West German design. Plutonium discharge data f o r Fast Breeder Reactors (FBRs) represented by the Liquid-Metal Fast Breeder Reactor (LMFBR) are included f o r comparison. t h a t the denatured f u e l cycle f o r the HWR gives the greatest reduction i n plutonium production between the regular and denatured cycles. The HTGR has about the same absolute plutonium production f o r the denatured f u e l cycle as the HWR and i n both cases the plutonium amounts are rather small. The HTR-PBR i s best i n

L

I t i s q u i t e c l e a r from Table 7.1-1

L

absolute minimum plutonium production, y i e l d i n g only 14 kg/GWe-yr and even l e s s i n a h i g h l y optimized design. Table 7.1-1. F i s s i l e Plutonium Discharge f o r Various Reactor and Fuel Cycle Combinations (Capacity Factor = 0.75) F i s s i l e Pu Discharge (kg/GWe-yr) LEU Cycle Pu/U Cycle Denatured Cycle LWR

7 74

SSCR

196 183b 72 63

858'

63

72 HWR (CANDU) 32 HTGR 36 .HTR-PBR . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14. . . 991 347 LMFBR ;Plutonium burner. S l i g h t l y enriched CANDU.

For the LWR, SSCR and HWR the percentage o f the discharge plutonium t h a t I s f i s s i l e plutonium i s approximately the same f o r the denatured cycle as f o r the LEU cycle. For the HTGR and PBR, the f i s s i l e plutonium percentage i s only %39%f o r the denatured cycle (compared t o 56% f o r the LEU cycle). Further, the discharge plutonium from the HTGR and PBR, and also from the HWR, i s more f o u r than t h a t from the LWR o r HTR, and HWR t o obtain a given t i o n r e s t r a i n t associated w i t h

d i l u t e d w i t h other heavy material by a f a c t o r o f three t o SSCR. Thus, more material must be processed i n the HTGR, amount o f plutonium, which provides an a d d i t i o n a l p r o l i f e r a spent f u e l discharged from these reactors. However, the

on-line r e f u e l i n g feature o f the CANDU, and also o f the PBR, may be a disadvantage from a p r o l i f e r a t i o n viewpoint since low-burnup f u e l could be removed and weapons-grade plutonium extracted from it. On the other hand, premature discharge o f low-burnup f u e l from t h e reactors would i n c u r economic penalties.

L L I'


7-9

Viewed solely from the plutonium production viewpoint, the order of preference in terms of higher proliferation resistance for the various denatured reactor candidates to be employed a t dispersed s i t e s i s as follows: HTR-PBR, HWR, HTGR, LWR, and SSCR. However, other factors must also be addressed in evaluating the candidate reactors, one of which i s that t h e i r plutonium production maintains the symbiosis of a system t h a t includes plutonium-fueled 233U producers i n secure energy centers. This plutonium being consumed within the center as i t i s recovered from the spent fuel would limit the amount of plutonium available f o r possible diversion. While such an energy center could also be implemented f o r the Pu/U cycle, the denatured cycle would permit the dispersal of a larger fraction of the recycle-based power generation capability. Hence, the number and/or s i z e of the required energy centers m i g h t be markedly reduced r e l a t i v e t o the number required by the Pu/U cycle. 7.1.4.

Conclusions

The proliferation-resistant characteristics of the denatured 233U fuel cycle derive from i t s i n t r i n s i c isotopic barrier, i t s gamma radiation barrier, and i t s relatively low content of chemically separable f i s s i l e material in spent fuel: The isotopic denaturing of the denatured 233U cycle would provide a significant technical barrier (although n o t an absolute one) that would decrease with time a t a rate which i s country-specific. Technologically primitive countries will find i t an imposing barrier relative t o other routes. Countries that have the technological expertise to develop isotope separation capabilities will have the technology required to circumvent t h i s barrier; however, they will also have the option of utilizing possible indigeneous natural uranium or low enriched 235U fuel as alternate feed materials. The denatured 233U cycle imposes a significant radiation barrier due t o the 232U daughter products i n the fresh fuel as an inherent property of the cycle. Such a radiation f i e l d increases the e f f o r t required to obtain weapons-usable material from fresh denatured reactor fuel. While the amount of plutonium discharged in the denatured 233U fuel cycle i s significantly less than in either the Pu/U cycle or the LEU cycle, the presence of plutonium i n the cycle (even though i t i s i n the spent f u e l ) does represent a proliferation concern. Conversely, i t also represents a resource potentially useful i n a symbiotic power system employing denatured fuel. The concept of a safeguarded energy center provides a means of addressing this duality i n t h a t the f i s s i l e plutonium can be burned in the center t o produce a proliferationr e s i s t a n t fuel.

I n summary, the denatured 233U fuel cycle offers a technical contribution t o prol i f e r a t i o n resistance. However, the fuel cycle must be supplemented w i t h political and institutional arrangements also designed t o discourage proliferation.


7-1 0

7.2.

IMPACT OF DENATURED 233U FUEL ON REACTOR PERFORMANCE AND SELECTION: COMPARISON WITH OTHER FUEL CYCLES

T. J. Burns Oak Ridge National Laboratory

The discussion i n Chapter 4 has shown t h a t the impact o f t h e denatured 233U fuel c y c l e on the performance o f the various r e a c t o r s considered i n t h i s study i s l a r g e l y due t o differences i n t h e nuclear p r o p e r t i e s o f 233U and 232Th r e l a t i v e t o those o f 239Pu (and 235U) and 238U, r e s p e c t i v e l y . f u e l than e i t h e r 239Pu o r

235U,

For thermal systems, 233U i s a s i g n i f i c a n t l y b e t t e r both i n terms o f energy production and i n terms o f the

conversion r a t i o * t h a t can be attained.

For f a s t systems, however, t h e s u b s t i t u t i o n o f 233U-based f u e l s f o r 239Pu-based f u e l s r e s u l t s i n a somewhat poorer r e a c t o r performance,

p a r t i c u l a r l y with respect t o t h e breeding ratio.*

I n t h i s s e c t i o n t h e performance o f t h e

various r e a c t o r s operating on the denatured 233U f u e l c y c l e i s compared w i t h t h e i r performance on other f u e l cycles.

I n a d d i t i o n , the dependence o f the denatured 233U f u e l

c y c l e on a u x i l i a r y fuel cycles f o r an adequate supply o f 233U i s discussed.

Because o f t h i s

dependence, r e a c t o r s f u e l e d with denatured 233U must be operated i n symbiosis with r e a c t o r s t h a t produce 233U.

These l a t t e r reactors, r e f e r r e d t o as transmuters, may be e i t h e r thermal

r e a c t o r s o r f a s t reactors.

The p a r t i c u l a r r e a c t o r s selected f o r operation as transmuters and

those chosen t o operate on denatured 233U f u e l w i l l depend on several f a c t o r s , two o f the most important being t h e resource requirements o f t h e n d i v i d u a l r e a c t o r s and t h e energy growth c a p a b i l i t y r e q u i r e d of t h e symbiotic system.

The i n f l u e n c e o f these various f a c t o r s i s

p o i n t e d o u t i n the discussion below. 7.2.1

Thermal Reactors

I n comparing t h e performance o f thermal r e a c t o r s operating on denatured 233U f u e l w i t h t h e i r performance on other fuels, i t i s u s e f u l t o d i s t i n g u i s h between two generic f u e l c y c l e types:

those t h a t do n o t r e q u i r e concurrent reprocessing ( t h a t i s , once-through systems) and those t h a t do. Although t h e denatured 233U f u e l c y c l e cannot i t s e l f be employed as a once-through system, t h e implementation o f the MEU(235)/Th once-through c y c l e i s a l o g i c a l f i r s t step t o t h e implementation o f the denatured

233U

cycle.

Thus both

once-through and r e c y c l e scenarios are considered here f o r thermal reactors. Once-Through Systems Two f u e l cycles o f i n t e r e s t t o t h i s study can be implemented w i t h o u t concurrent reprocessi n g c a p a b i l i t y : t h e LEU c y c l e and t h e MEU(235)/Th cycle.

*

The LEU c y c l e i s , o f course, already used

The conversion ratio and breeding ratio are both defined as t h e r a t i o o f t h e r a t e a t which f i s s i l e m a t e r i a l i s produced t o t h e r a t e a t which f i s s i l e m a t e r i a l i s destroyed a t a s p e c i f i c p o i n t i n time ( f o r example, a t t h e midpoint o f t h e e q u i l i b r i u m cycle). The term conversion r a t i o i s a p p l i e d t o those r e a c t o r s f o r which t h i s r a t i o i s l e s s than 1, which i s u s u a l l y t h e case f o r thermal reactors, w h i l e t h e term breeding r a t i o i s a p p l i e d t o those r e a c t o r s f o r which t h i s r a t i o i s g r e a t e r than 1, which i s u s u a l l y t h e case f o r f a s t reactors (i .e., breeders).


7-1 1

r o u t i n e l y i n LWRs and small-scale f a b r i c a t i o n o f MEU(235)/Th fuels f o r LWRs might be a t t a i n able w i t h i n 2

-

- 3 years.

However, i t i s pointed out t h a t the once-through c y c ~ chas two

-

variants throwaway and stowaway and i n c e r t a i n systems (for example, the PWR, as noted that is, below). t h e MEU(235)/Th cycle might be economic only from a stowaway standpoint

L

-

only i f a reprocessing c a p a b i l i t y i s eventually envisioned. Table 7.2-1

sumnarizes the U308 and separative work requirements estimated f o r PWRS

HWRs, HTGRs, PBRs, and SSCRs operating as once-through systems on both the LEU and the MEU(235)/Th cycles. Several i n t e r e s t i n g points are evident from these data. The LEU-HWR requires the smallest resource comnitment (as w e l l as the smallest SWU requirement). The conventional PWR requires a s i g n i f i c a n t l y greater resource comnitment and l a r g e r SWU requirements f o r the MEU/Th once-through cycle than f o r t h e LEU once-through cycle and hence no i n c e n t i v e e x i s t s f o r the MEU/Th cycle on PWRs i f only the throwaway option i s considered.

S i g n i f i c a n t l y , however, both o f the gas-cooled graphi te-moderated reactors, the

HTGR and the PBR, require smaller U308 commitments f o r t h e MEU/Th once-through cycle than f o r the LEU case. Moreover, f o r both o f these reactors, the SWU requirements f o r the MEU/Th cycle are n o t s i g n i f i c a n t l y d i f f e r e n t from those f o r the LEU cycle; i n f a c t , f o r t h e PBR, the MEU/Th cycle i s s l i g h t l y l e s s demanding than the LEU cycle. These e f f e c t s are p r i m a r i l y due t o the high burnup design o f both the HTGR and the PBR. A t the higher burnup l e v e l s o f the gas-cooled reactors, most o f the 233U produced i n the MEU/Th cycle i s burned i n s i t u and contributes s i g n i f i c a n t l y t o both the power and the conversion r a t i o . I t i s also i n t e r e s t i n g t o note that, while n o t considered i n Table 7.2-1, the unique design o f the PBR would permit recycle o f the f e r t i l e elements without intervening reprocessing and thus

would f u r t h e r reduce both the ore and SWU requirements f o r t h e MEU/Th cycle.

[Note:

The

data given i n Table 7.2-1 f o r PWRs considers only current comnercially deployed designs. Studies now underway i n the WE-sponsored Nonproliferation A l t e r n a t i v e Systems Assessment Program (NASAP) i n d i c a t e t h a t LWR modifications t o reduce uranium requirements are feasible. Similarly, much o f t h e other reactor data are subject t o design refinement and uncertainties, as w e l l as t o f u t u r e optimization f o r s p e c i f i c roles.] Table 7.2-1.

30-Year Uranium and Separative Kork Requirgments f o r Once-Through LEU and MEU(235)/Th Fuel Cycles*' Uranium Requirement (ST U308/GWe)

Reactor

L i; i;

LEU

Separative Work Requirement (MT SWU/GWe)

MEU/Th

LEU

MEU/Th

PWR HWR

5989 3563

8360 8281

3555 666

7595 7521

HTGR

4860

4515

3781

4143

4184' 7920

3891 3010

3669 7160

PBR SSCR

4289 5320

a -75% capacity factor; no c r e d i t f o r end-of-1 i f e core inventories; ,0.2% t a i l s . "The data presented i n t h i s t a b l e are consistent w i t h the data subm t t e d by t h e U.S. t o INFCE ( I n t e r n a t i o n a l Nuclear Fuel Cycle Evaluation) f o r the cases i n which corresponding reactors are considered. C Does n o t include recycle o f f e r t i l e elements without intervening r processing.


I;'

7-1 2

Ifthese once-through systems are operating on the throwaway option, the f i s s i l e material discharged i n t h e i r spent f u e l elements i s deemed unusable; i n f a c t , no value i s assigned t o the spent f u e l i n once-through f u e l cycle accounting. Thus, i n t h i s case the most resource-efficient once-through f u e l cycle i s the one t h a t requires the lowest f i s s i l e charge per u n i t power. If, however, a c a p a b i l i t y f o r reprocessing the spent f u e l i s eventually envisioned (i.e., i f the throwaway option becomes a stowaway option), then the q u a n t i t y o f f i s s i l e material i n the spent f u e l becomes an important consideration. E s t i mates o f the amounts o f the various f i s s i l e materials discharged by each reactor type operating on both t h e LEU cycle and the MEU(235)/Th cycle are given i n Table 7.2-2. Table 7.2.2.

I; i;

L

30-Year Charge and Discharge Quantit i e s f o r Once-Through Fuel Cyclesa

~

235u

Reactor

PWR

HllR HTGR PBR SSCR

Charge

24.72 17.53 19.49 18.09 22.25

233U

-

..

MTIGWe F i s s i 1e Dischargeb Total 23% Puf Fissile

Cumulative Net F i s s i l e Consumption

c

LEU Cycle 6.45

5.22

11.67

13.05

1.77

5.49

7.26

10.37

3.25

2.16

2.79 5.46

1.89 5.88

5.41 4.68 11.34

14.08 13.41 10.91

21.45 25.11 4.35 4.32

12.38 7.52 13.64 12.23

MEU(235)/Th Cycle PWR HWR HTGR PBR

33.83 32.63 17.99 16.55

7.80 14.28 2.31 2.73

11.52 10.08 1.35

2.13 0.75 0.69

1.17

0.42

a

bAt 75% caDacitv factor. Estimated from -equilibrium cycle. For the PUR and HWR, the use o f the MEU/Th f u e l c y c l e r a t h e r than the LEU fuel cycle r e s u l t s i n a s i g n i f i c a n t increase i n the amount o f f i s s i l e material contained i n the spent f u e l . It should be noted, however, t h a t t h i s increase i s p r i m a r i l y the r e s u l t

o f higher feed requirements (i-e., 235U commitment). I n contrast, converting from the LEU cycle t o the MEU/Th cycle does n o t m a t e r i a l l y a f f e c t the n e t consumption o f the gascooled HTGR and PBR (although i t dramatically a f f e c t s the types o f f i s s i l e material present i n t h e i r spent f u e l ) . The r e l a t i v e l y low values f o r the discharge q u a n t i t i e s f o r the gas-cooled reactors i s the r e s u l t o f two effects: a lower i n i t i a l loading; and a design t h a t i s apparently based on higher burnup, which i n t u r n reduces t h e amount o f f i s s i l e material discharged. F i n a l l y , i t i s t o be remembered t h a t the resources represented by t h e spent f u e l inventory are recoverable only when the spent f u e l i s reprocessed, whereas the U308 commitment i s necessary throughout t h e operating l i f e t i m e o f t h e reactor. Thus, i n a sense, t h e spent f u e l resource must be discounted i n time t o order t o assess the best system from a resource u t i l i z a t i o n basis.

c L L


'I

t

7-1 3

The i s o t o p i c composition of the spent f u e l i n v e n t o r i e s i s a l s o o f i n t e r e s t from a p r o l i f e r a t i o n standpoint. For b o t h the LEU and t h e MEU/Th once-through f u e l cycles, the f i s s i l e uranium content of the spent f u e l i s denatured ( d i l u t e d w i t h 238U) and hence i s protected by t h e i n h e r e n t i s o t o p i c b a r r i e r .

I:

Thus the plutonium i n t h e f u e l would be t h e

f i s s i l e m a t e r i a l most s u b j e c t t o diversion. The use o f the MEU/Th c y c l e i n place o f the LEU c y c l e sharply reduces t h e amount o f plutonium produced (by 60-80%, depending on r e a c t o r type), and f o r both cycles t h e q u a n t i t y o f plutonium produced i n t h e gas-cooled reactors i s s u b s t a n t i a l l y l e s s than t h a t produced i n the o t h e r r e a c t o r types. Recycle Systems

I f r e c y l i n g o f t h e f i s s i l e m a t e r i a l i n t h e thermal reactors i s permitted, then 233U (and plutonium) produced i n the MEU(235)/Th i s recoverable on a schedule d i c t a t e d by t h e production r a t e o f the system. Table 7.2-3 gives estimates o f t h e n e t l i f e t i m e consumption and production o f various f i s s i l e m a t e r i a l s f o r the MEU(235)/Th f u e l c y c l e under the assumption t h a t t h e c a p a b i l i t y f o r uranium r e c y c l e i s a v a i l a b l e . ated does not r e f l e c t t h e 235U l o s t t o the enrichment t a i l i n g s . ) the MEU(233)/Th f u e l c y c l e estimates are a l s o provided.

(The 235U consumption t a b u l For comparison purposes,

The most s t r i k i n g aspect o f

Table 7.2-3 i s t h e apparent 30% reduction o f f i s s i l e consumption achieved w i t h the 233U system, i n d i c a t i n g t h e higher value o f 233U as a thermal r e a c t o r f u e l . I n f a c t , the t r u e e x t e n t o f t h i s e f f e c t i s masked somewhat since a l a r g e f r a c t i o n o f the recycled f u e l f o r the 235U makeup case i s i n f a c t 233U.

I t should a l s o be noted t h a t t h e MEU(233)/Th c y c l e generally r e s u l t s i n

a smaller n e t plutonium production, even though t h e degree o f denaturing i s l e s s (i.e., 238U f r a c t i o n o f uranium charged i s higher).

Table 7.2-3. Estimated Net 30-Year F i s s i l e Consumption and Production f o r MEU/Th Cycles w i t h Uranium Recyclea MT/GWe With 235U l o a d i n g and Makeup Reactor

235U Consumption

F i s s i l e Pu Production

9.1

1.89

3.1

0.96

10.4

7.7

1.09

8.7

2.56

5.9

2.44

12.5 4.5

HTGR PBR SSCR

-

A t 75% capacity factor.

k:

233U Consumption

2.85 0.90 1.13

PWR

HWR

a

F i s s i l e Pu Production

With 233U Loading and Makeup

-

-

-

the


I 7-14

As has been stated e a r l i e r , the consideration o f an MEU/Th cycle t h a t u t i l i z e s 233U makeup presumes the existence o f a source o f the r e q u i s i t e 233U. Although the 233U i n the spent f u e l elements would be recovered, the amount would be inadequate t o maintain the system and an exogenous source must be developed. One means f o r generating 233U i s by using a Pu/Th-oxide-fueled thermal reactor. Table 7.2-4 summarizes some p e r t i n e n t r e s u l t s f o r the various thermal reactors operating on the Pu/Th cycle. I t should be noted t h a t the HTGR case given i n Table 7.2-4 i s f o r a case i n which the f u l l core i s refueled every 5 yr and i s n o t optimized f o r 2 3 % production. Thus, much o f the 233U bred during t h i s period i s consumed i n providing power, and the transmutation e f f i c i e n c y (tons o f plutonium "transmuted" i n t o tons o f 233U) i s s i g n i f i c a n t l y reduced r e l a t i v e t o the PWR and SSCR. The transmutation e f f i c i e n c y o f 0.40 f o r the PWR and SSCR i s also rather poor, however, compared t o the'l.20 Production o f 233U v i a plutonium-consuming value f o r a Pu/Th-fueled FER (see Section 4.5). transmuters i s more s u i t e d t o f a s t reactors. On the other hand, i t i s recognized t h a t Pu/Thfueled therma reactors could provide an i n t e r i m source o f 23311.

Table 7.2-4.

Net 30-Year F i s s i l e Conspption and Production f o r Pu/Th Cycles

L; I; L

L L

MTIGWe Reactor

F i s s i l e Pu Consumption

PWR HWR~

20.7 19.84

HTGR PER SSCR

233U Output

Transmutatton Efficiency

8.16 31.76

0.394 0.593

16.5

3.03

-

0.184

23.8

9.63

0.405

-

f:

-

a A t 75% capacity factor, using e q u i l i b r i u m cycle bvalues. From data i n Table 6.1-3. 7.2.2.

Fast Reactors

I n t h i s study f a s t reactors have been considered as possible candidates f o r two roles: as power reactors operating on denatured 233U f u e l ; and as transmuters burning plutonium t o produce 233U. With LMFBRs used as the model, the denatured FBRs were analyzed f o r a range o f 233U/U enrichments t o parameterize the impact o f the f u e l on the reactor performance (see Section 4.5), and t h e transmuter FBRs were analyzed both f o r a P u / ~ ~ * U core d r i v i n g a Tho2 blanket and f o r a Pu/Th system i n which the thorium was included i n both the core and the blanket.

b;

The specified 233U/U enrichment i s a c r u c i a l parameter f o r the denatured f a s t reactors. Increasing the allowable enrichment permits more thorium t o be used i n the f u e l .e., reduces the material and hence allows the reactors t o be more s e l f - s u f f i c i e n t (i

ii

L f'

I'

.L;


7-1 5

U

L

r e q u i r e d 233U makeup).

Increasing the 2 3 3 U enrichment a l s o reduces the amount o f f i s s i l e

plutonium contained i n t h e discharged f u e l , which i s obviously d e s i r a b l e from a safeguards viewpoint. However, increasing the 233U f r a c t i o n a l s o increases t h e v u l n e r a b i l i t y o f the denatured f u e l t o i s o t o p i c enrichment, e f f e c t i v e l y f o r c i n g a compromise between p r o l i f e r a t i o n concerns regarding the f r e s h f u e l versus p r o l i f e r a t i o n concerns regarding the spent f u e l . The lowest enrichment f e a s i b l e f o r t h e denatured LMFBR systems analyzed :iez i n the range o f 11-14%. Such a system would u t i l i z e U02 as f u e l and would r e q u i r e s i g n i f i c a n t amounts of 233U as makeup since t h e plutonium i t produced could n o t be r e c y c l e d i n t o i t .

L

The "breeding" r a t i o components o f c e r t a i n denatured LMFBRs as a f u n c t i o n o f 233U f The r a t i o o f 233U produced t o Pu produced i s very enrichment are shown i n Table 7.2-5. This sugs e n s i t i v e t o the s p e c i f i e d degree o f denaturing i n the range o f 12-20% 233U/U.

, increased

gests t h a t s i g n i f i c a n t performance improvements may be p o s s i b l e (i .e.

233U produc-

t i o n and decreased 239Pu production) f o r r e l a t i v e l y small increases i n t h e denaturing criteria.

L

O f course, t h e o v e r a l l "breeding" r a t i o o f the denatured LMFBR i s s i g n i f i c a n t l y

degraded below t h a t f o r t h e reference P U / ~ ~c y~c Ul e (see Table 4.5-1

Table 7.2-5.

Denatured LMFBR Mid-Equilibrium Cycle Breedi ng R a t i o Components*

"Breeding" Component

233U

23311

Enric hment

i n Chapter 4).

Pu "Breeding" Component

%12%

0.41

0.71

Overall "Breeding" Ratio 1.12

20%

0.70

0.39

1.09

40%

0.90

0.15

1.05

100%

1.02

-

1.02

*Using values from Section 4.5-1. A more recent s t u d y [ P r o l i f e r a t i o n Resistant Large Core Design Study (PRLCDS)] i n d i c a t e s t h a t s u b s t a n t i a l improvements i n t h e FBR performance i s possible. Because o f t h e superior breeding p o t e n t i a l o f a 239Pu-fueled system r e l a t i v e t o a 233U-fueled system i n a f a s t neutron spectrum, t h e f a s t r e a c t o r i s i d e a l l y s u i t e d t o t h e r o l e o f a plutonium-fueled transmuter.

Moreover, i n c o n t r a s t t o t h e thermal trsnsniuters,

t h e f a s t r e a c t o r s r e s u l t i n a n e t o v e r a l l f i s s i l e m a t e r i a l gain.*

L ld

1

i"

Two types o f FBR transmuters have been analyzed f o r the c l a s s i c a l homogeneous FBR core c o n f i g u r a t i o n (a c e n t r a l homogeneous core surrounded by f e r t i l e blankets).

I n the f i r s t , the usual Pu/238U-fueled core was assumed w i t h a Tho2 r a d i a l b l a n k e t ( a l s o a Tho2 a x i a l

blanket i n one case).

I n t h e second type a Pu/Th core was assumed.

t h e n e t production data f o r t y p i c a l f a s t transmuters of each type. gain/cycle w i t h t h e

Table 7.2-6 summarizes The o v e r a l l f i s s i l e

core i s s i g n i f i c a n t l y higher than t h a t w i t h the Pu/Th core,

t h e r e s u l t being t h a t the "breeding" r a t i o i s n o t n o t i c e a b l y reduced from t h e breeding r a t i o f o r t h e reference

cycle.

The production o f 233U i n the Pu/Th r e a c t o r i s

approximately a f a c t o r o f 4 higher, b u t t h i s i s achieved as a r e s u l t o f " s a c r i f i c i a l " consumption o f plutonium.

*

Thus, these two r e a c t o r types r e f l e c t a t r a d e o f f between 233U

As noted i n Chapter 4.5, s i g n i f i c a n t u n c e r t a i n t i e s are associated w i t h t h e fast-neutron cross sections f o r 2 3 3 U and Th.


c

7-16

and o v e r a l l f i s s i l e production (i.e., Table 7.2-6.

Core Material

p o t e n t i a l growth r a t e ) .

E q u i l i b r i u m Cycle Net F i s s i l e Production f o r Pot e nt ia 1 LMFBR Transmuters*

Reactor A x i a l Blanket Material

Net F i s s i l e Production - (kg/GWe yp)

-

Radial Blanket Materi a1

Pu

233U

L L

Fissile

*Using values from Section 4.5-1 ( ~ 7 5 %capacity f a c t o r ) . A more r e c e n t study [ P r o l i f e r a t i o n Resistant Large Core Design Study (PRLCDS)] i n d i cates t h a t s u b s t a n t i a l improvements i n the FBR performance i s possible.

I n a d d i t i o n t o t h e systems u t i l i z i n g t h e c l a s s i c a l homogeneous core c o n f i g u r a t i o n , systems u t i l i z i n g a heterogeneous core c o n f i g u r a t i o n (i.e., interspersed f i s s i l e and f e r t i l e regions) were examined as a p o s s i b l e means o f improving t h e performance o f f a s t r e a c t o r s operating on a l t e r n a t e f u e l cycles. The s u b s t i t u t i o n o f d i f f e r e n t coolants and f u e l forms (i.e., carbides and metals versus oxides) were a l s o considered. The n e t e f f e c t o f these changes i s t o increase t h e f u e l volume f r a c t i o n i n the r e a c t o r core, harden t h e spectrum, or, i n some cases, both. The advanced f a s t r e a c t o r concepts show s i g n i f i c a n t improvement regarding t h e breeding r a t i o (and doubling time) r e l a t i v e t o t h e c l a s s i c a l design when operating on a l t e r n a t e f u e l cycles; however, the performance o f t h e a l t e r n a t e f u e l cycles i s s t i l l degraded over t h a t of the same r e a c t o r type operating on U the P u / ~ ~ * cycle. 7.2.3.

Symbiotic Reactor Systems

As has been s t a t e d throughout t h i s r e p o r t , i n considering denatured 233U r e a c t o r systems i t i s assumed t h a t t h e denatured r e a c t o r s w i l l operate as dispersed power systems supported by fuel c y c l e services and r e a c t o r transmuters l o c a t e d i n secure energy centers. When the system i s i n f u l l operation no external source o f f i s s i l e m a t e r i a l i s supplied; t h a t i s , t h e system i s self-contained. I n i t i a l l y t h e resource base (i.e., n a t u r a l uranium) can be used t o provide a source of 233U f o r implementing t h e denatured 233U f u e l c y c l e [ v i a t h e MEU(235)/Th cycle]; however, a s h i f t t o plutonium-fueled transmuters w i l l e v e n t u a l l y be required. During t h i s t r a n s i t i o n period, t h e system can be characterized by t h e r a t e a t which t h e resource base i s consumed (see Chapter 6). I n order t o compare t h e long-term p o t e n t i a l o f various r e a c t o r systems under t h e r e s t r i c t i o n s imposed by t h e denatured f u e l cycle, two system parameters have been developed: (1) t h e energy support r a t i o , defined as t h e r a t i o o f dispersed r e a c t o r power r e l a t i v e t o t h e energy center ( o r c e n t r a l i z e d ) power, and (2) t h e i n h e r e n t growth p o t e n t i a l o f t h e system. Since both t h e growth r a t e and t h e energy support r a t i o i n v o l v e f i s s i l e mass flows, they are i n t e r r e l a t e d . I n o r d e r t o unambiguously determine both parameters, t h e i n h e r e n t system growth r a t e i s determined a t t h e asymptotic value o f t h e support r a t i o , a value which can be viewed as t h e " n a t u r a l " operati n g r a t i o o f t h e system.

I]

t;,

n

c c

I]

L

c

c


i

7-1 7

Three generic types o f symbiotic r e a c t o r systems can be envisioned by considering

ii

various combinations o f thermal converters and f a s t breeders f o r t h e dispersed (D) and energy center (S) reactors: thermal (D)/thermal (S), thermal (D)/fast(S), and fast(D)/fast(S). I n order f o r t h e generating capacity o f a system t o increase w i t h time w i t h o u t an external supply o f f i s s i l e material, a n e t gain of f i s s i l e m a t e r i a l (of some type) must occur.' Thus, t h e growth p o t e n t i a l of t h e thermal (D)/thermal(S) system i s i n h e r e n t l y negative; t h a t i s , t h e i n s t a l l e d nuclear capacity must decay as a function of,time since t h e o v e r a l l conversion

I

r a t i o i s l e s s than 1. The thermal(D)/fast(S) system, however, does have t h e p o t e n t i a l f o r growth since t h e n e t f i s s i l e gain o f t h e f a s t component can be used t o o f f s e t fhe f i s s i l e

I,

f a s t ( S ) ] and t h e growth r a t e c l e a r l y e x i s t s f o r t h i s system, since maximizing t h e support

l o s s o f t h e thermal reactors. I

However, a tradeoff between t h e support r a t i o [thermal(D)/

r a t i o w i l l mean t h a t n e t fissile-consuming reactors w i l l c o n s t i t u t e the major f r a c t i o n o f t h e system and t h e growth r a t e w i l l be d e t r i m e n t a l l y a f f e c t e d . The f a s t ( D ) / f a s t ( S ) system provides a g r e a t deal more f l e x i b i l i t y i n terms o f t h e allowable energy support r a t i o and i n h e r e n t growth rate.

1 I L L k

To i l l u s t r a t e t h e tradeoff between the growth p o t e n t i a l and t h e support r a t i o , t h e "operating envelopes" shown i n Fig. 7.2-1

have been generated using denatured PWR data

from Section 4.1 and LMFBR transmuter data from Section 4.5.1.

Each envelope represents t h e locus of permissible s y m t i o t i c parameters (growth r a t e , support r a t i o ) f o r t h e system considered,' i.e., t h e permissible combinations of growth r a t e and support r a t i o f o r each s p e c i f i c r e a c t o r combinations.

A t p o i n t s A, B, and C on t h e curves, the transmuter used .is

respectively, t h e c l a s s i c a l (Pu/U)O, reference system w i t h a U02 r a d i a l blanket, a (Pu/U)02 system w i t h a Tho, r a d i a l blanket, and a (Pu/Th)02 system w i t h a Tho2 r a d i a l blanket. A t each p o i n t along t h e curves connecting p o i n t s A, B, and C, t h e transmuter i s a combination o f t h e two r e a c t o r s defined by t h e end p o i n t s o f each curve segment (see key i n upper r i g h t hand corner). Points within t h e envelope correspond t o combinations o f t h e t h r e e transmuters i n d i f f e r e n t proportions. The lower envelope i n Fig. 7.2-la

(repeated i n Fig. 7.2-lb)

i l l u s t r a t e s the tradeoff

f o r t h e denatured PWRs and LMFBR transmuters, and t h e upper envelope d e p i c t s t h e f a s t / f a s t analogue i n which t h e denatured PWR i s replaced by an ~ 1 2 % denatured LRFPR. As indicated, t h e fast(D)/fast(S) symbiotic system provides a higher growth r a t e f o r a given energy supp o r t r a t i o , and, moreover, t h e growth r a t e i s always p o s i t i v e . The upper envelope i n Fig. 7.2-lb represents t h e corresponding case using 20% denatured LMFBRs.

L i

In a l l cases t h e f a s t r e a c t o r data u t i l i z e d were taken from Section 4.5.1; t h a t i s , homogeneous LMFBR cores were assumed. The use of a heterogeneous core f o r t h e transmuter r e a c t o r would have t h e e f f e c t o f d i s p l a c i n g t h e curves i n Fig. 7.2-1 upwards and t o t h e right.

The employment of an advanced converter (a h i g h conversion r a t i o thermal r e a c t o r )

would have a s i m i l a r e f f e c t on t h e thermal/fast curve.


7-1 8

ORNL-DWG 7R-19Anl

I; L

pc SEGMENT AB: SEGMENT BC: SEGMENT AC:

(a) TRANSMUTER x.A + ( l - X ) * B TRANSMUTER = X - B + (l-X).C TRANSMUTER X - A + ( l - X ) * C 0 5 x 5 1

c LMFBR T r a n s r n u t e r s

+ 12% D e n a t u r e d LMFBRs

.0

0.5 IB

5 ENERGY SUPPORT RATIO

c REACTOR A: REACTOR 8: REACTOR C *

(PU/U)02 + U02 RB (Pu/U)02 + Tho2 RB fPu/Th)02 + Tho2 RB

C

I:

ORNL-DWG 78-19809 SEGMENT AB: SEGMENT BC: SEGMENT AC:

A

+

TRANSMUTER = X - A + TRANSMUTER = X * B + TRANSMUTER = X - A + 0 5 x 5 1

L tl Lz

LMFBR T r a n s r n u t e r s

Zni c ( Y

w

E: a

4.S ENERGY SUPPORT RATIO

REACTOR A: REACTOR B:

9 B

REACTOR C:

Fig. 7.2-1.

(Pu/U)02 + U02 RB

C

(Fu/U)02 + Tho2 RB (Pu/lh)02 + Tho2 RB

Operating Envelopes f o r Symbiotic Systems U t i l i z i n g LMFBR Transmuters.

L

t:


~~

li

7-1 9

7.2.4.

i

Conclusions

Since o p t i m i z a t i o n o f t h e various reactors f o r t h e p a r t i c u l a r f u e l c y c l e considered was beyond t h e scope of t h i s study, t h e r e s u l t s presented above a r e s u b j e c t t o several uncert a i n t i e s . Nevertheless, c e r t a i n general conclusions on t h e impact o f t h e various f u e l cycles on r e a c t o r performance are b e l i e v e d t o be v a l i d :

i

For once-through throwaway systems, the various systems studied are ranked i n order o f resource u t i l i z a t i o n as follows: t h e HWR on the LEU cycle; t h e HTGR and HTR-PBR on e i t h e r t h e LEU c y c l e o r on the MEU/Th cycle; and t h e SSCR and PWR on t h e LEU cycle. On t h e MEU/Th c y c l e t h e SSCR and PWR r e q u i r e more uranium than they do on t h e LEU c y c l e and hence do n o t m e r i t f u r t h e r consideration f o r once-through operation.

L

0

For once-through stowaway systems, i n which t h e f i s s i l e m a t e r i a l i n t h e spent f u e l i s expected t o be recovered a t some f u t u r e date, the r e l a t i v e ranking of t h e systems would depend on the u l t i m a t e d e s t i n a t i o n o f the

L i

f i s s i l e material.

I f f u t u r e nuclear power systems are t o be thermal

r e c y c l e systems, then e a r l y emphasis should be placed on r e a c t o r s and

L

f u e l cycles t h a t have a high 2 3 3 U discharge. Ift h e f u t u r e systems are t o be f a s t r e c y c l e systems, then emphasis should be placed on reactors and f u e l cycles t h a t w i l l provide a plutonium inventory.

& 0

For r e c y c l e systems u t i l i z i n g thermal reactors, t h e p r e f e r r e d basic f i s s i l e m a t e r i a l i s 233U. However, implementation o f a 233U f u e l c y c l e w i l l r e q u i r e an exogenous source of the f i s s i l e m a t e r i a l ; therefore, i t i s l i k e l y t h a t t h e MEU(235)/Th c y c l e would be implemented f i r s t t o i n i t i a t e the produc-

I '

t i o n o f 233U.

i

i

h i

i

h

I'

235U and the 233U would be recycled; thus the

system would evolve towards t h e MEU(233)/Th cycle, which i s the denatured 233U c y c l e as defined i n t h i s study. However, i t i s t o be emphasized t h a t these reac-

kr

ii L

Both the unburned

t o r s w i l l n o t produce enough 233U t o s u s t a i n themselves and separate 233U production f a c i l i t i e s must be operated. A Pu/Th-fueled thermal r e a c t o r has been considered as a 0

233U

production f a c i l i t y .

For r e c y c l e systems u t i l i z i n g f a s t reactors, t h e p r e f e r r e d basic f i s s i l e m a t e r i a l i s 239Pu. Using 233U as the primary f i s s i l e m a t e r i a l o r p l a c i n g thorium i n t h e core sharply reduces t h e breeding performance o f f a s t reactors. However, f a s t r e a c t o r s using plutonium fuel and thorium blankets would be e f f i c i e n t 2 3 3U production f a c i l it i e s .


7-20

~

0

The inherent symbiotic nature o f the denatured 233U f u e l cycle (i.e.,

dispersed

reactors fueled w i t h denatured 233U and supported by energy-center reactors fueled w i t h Pu) mandates a tradeoff analysis of growth p o t e n t i a l versus energy support r a t i o ( r a t i o o f power produced outside the energy center t o the power produced i n s i d e the center), assuming no external source o f f i s s i l e material.

For thermal/thermal

systems, the growth p o t e n t i a l i s negative. Fast/thermal systems would permit some o f the net f i s s i l e gain (i.e., growth p o t e n t i a l ) o f the f a s t reactors t o be s a c r i f i c e d f o r a higher energy support r a t i o . growth potential.

Fast/fast systems would provide the highest

Factors other than those a f f e c t i n g reactor performance would

also influence the choice o f reactors for the system,as has been discussed i n Chapters

5 and 6. Section 7.2.

1.

Reference

T. J . Burns and J . R. White, "Preliminary Evaluation o f Alternate Fuel Cycle Options Options U t i l i z i n g Fast Breeders," ORNL-5389 (1978).

L b

c h


7-21

PROSPECTS FOR IMPLEMENTATION AND COMMERCIALIZATION OF DENATURED 233U FUEL CYCLE

7.3.

J. C. Cleveland and T. J. Burns Oak Ridge National Laboratory Chapter 5 has discussed the reactors i n which denatured 2 3 % might be deployed, as we1 1 as t h e accompanying f u e l r e c y c l e f a c i l i t y requirements , and has presented schedules o f d ployment t h a t a r e based s o l e l y on t h e minimum time estimated t o be required t o solve These schedules, which have been used i n t h e nuclear power system

t e c h n i c a l problems.

evaluations presented i n Chapter 6, were developed i n discussions between Hanford Engineering Development Laboratory (HEDL) Argonne National Laboratory (ANL) , Oak Ridge Combustion Engineering (CE) , and t h e Department o f Energy (DOE)

National Laboratory (ORNL)

s p e c i f i c a l l y as a bounding case f o r assessing t h e maximum b e n e f i t s t h a t could be obtained by employing denatured

23%

fuel.

As a r e s u l t , t h e schedules a r e n o t e n t i r e l y c o n s i s t e n t

w i t h those t h a t have been developed subsequently i n t h e N o n p r o l i f e r a t i o n A1 t e r n a t i v e Systems Assessment Program (NASAP). not differ significantly,

While t h e i n t r o d u c t i o n dates o f the lead p l a n t s do

the NASAP scenarios p r e d i c t a much slower deployment o f

commercial reactors. The r e a c t o r i n t r o d u c t i o n dates and deployment schedules used i n t h i s study were based on the f o l l o w i n g assumptions: ~ 1 yr 0 t o develop/commercialize new f u e l design %14 yr t o develop/comercialize modified r e a c t o r design ~ 1 yr 8 t o develop/commercialize new advanced converter design

. ~ 2 4yr t o develop/commercialize new breeder design The r e s u l t i n g i n t r o d u c t i o n dates f o r t h e various reactors a r e as l i s t e d below, where t h e i n t r o d u c t i o n date i s defined as the date o f s t a r t u p o f t h e f i r s t u n i t , r e a c t o r deployment t h e r e a f t e r being l i m i t e d t o a maximum i n t r o d u c t i o n r a t e * ' by biennium o f 1, 2, 4,... 1969 1987

reactors:

- LWRs operating on -

LEU fuel LWRs operating on "denatured 235U'' f u e l (i.e.,

MEU(235)/Th)

- LWRs operating on denatured 233U, Pu/U, and Pu/Th f u e l s - SSCRs operating on LEU, denatured 233U, o r Pu/Th f u e l s 1995 - HWRs operating on any o f several proposed f u e l s 1995 - HTGRs operating on any o f several proposed f u e l s 2001 - FBRs operating on P U N , Pu/Th, o r denatured 233U f u e l s 1991

1991

Since t h e above i n t r o d u c t i o n dates are those estimated t o be the e a r l i e s t possible dates t h a t t e c h n i c a l problems could be resolved, i t i s c l e a r t h a t they cannot be achieved w i t h o u t s u b s t a n t i a l i n i t i a t i v e s and strong f i n a n c i a l support from t h e U.S., Government.

*

The i n t r o d u c t i o n r a t e o f any new technology i s l i k e l y t o be l e s s than t h e maximum r a t e noted above, since t h e c o n s t r u c t i o n market l o s s r a t e o f an established technology i s l i m i t e d t o 10% per year and t o t a l nuclear capacity a d d i t i o n s cannot exceed 15 GWelyr. 233U systems are f u r t h e r constrained because t h e number o r 233U-burning p l a n t s t h a t can be operated i s l i m i t e d by t h e 23%1 production r a t e .


7-22

Even w i t h government support, achieving t h e postulated schedules would be a d i f f i c u l t undertaking and would e n t a i l considerable r i s k since i t would be impossible t o f u l l y demonstrate an a l t e r n a t e r e a c t o r concept before c o n s t r u c t i o n on t h e i n i t i a l comnercial s i z e u n i t s has t o begin.

I'

A minimum o f s i x years would be r e q u i r e d t o c o n s t r u c t a nuclear

u n i t , and a minimum o f t h r e e years would be r e q u i r e d p r i o r t o c o n s t r u c t i o n f o r R&D and l i c e n s i n g approval. (It c u r r e n t l y takes 10 t o 12 y r t o l i c e n s e and c o n s t r u c t LWRs i n t h e U.S.)

A t l e a s t two a d d i t i o n a l years o f operation o f t h e demonstration u n i t would be

necessary t o e s t a b l i s h s a t i s f a c t o r y r e a c t o r performance.

Thus t h e e a r l i e s t time a new

r e a c t o r concept could be demonstrated i s i n the 1991-1995 p e r i o d indicated, and that assumes t h a t a commitment t o proceed has been made by 1980.

Because o f design, l i c e n s i n g ,

and c o n s t r u c t i o n schedules, t h e f i r s t comnercial u n i t s would have t o be ordered w e l l i n advance of t h e operation o f t h e i n i t i a l demonstration r e a c t o r t o achieve t h e b u i l d u p r a t e s assumed i n t h i s study.

I n order t o achieve such commitments p r i o r t o t h e f i r s t

successful demonstration, government support would have t o extend through t h e i n i t i a l commercial u n i t s i n a d d i t i o n t o t h e lead p l a n t .

The new r e a c t o r c y c l e would a l s o have t o

be perceived as economically advantageous t o a t t r a c t t h e p o s t u l a t e d number o f customers. A1 though several o f these r e a c t o r l f u e l options (e.g.,

PuITh LWRs, denatured advanced

L L

c

converters, etc.) a r e based on t h e use o f recycled f i s s i l e m a t e r i a l , i t should be emphasized t h a t comercial-scale reprocessing i s n o t n e c e s s a r i l y r e q u i r e d on t h e same time scale as t h e i n t r o d u c t i o n of t h e r e c y c l e f u e l types because t h e demand f o r r e c y c l e f i s s i l e m a t e r i a l may be q u i t e modest d u r i n g t h e i n i t i a l i n t r o d u c t i o n phase. I n t h e a n a l y s i s presented i n Chapter 6, many o f t h e new f u e l types are, i n f a c t , introduced before t h e associated f u e l

1;

reprocessing i s f u l l y developed, i t being assumed t h a t p i l o t o r prototype-plant scale reprocessing would be adequate t o support t h e i n i t i a l phase of deployment o f f u e l recycle. Hence, although commercial reprocessing o f 233U-containing f u e l s i s n o t p r o j e c t e d u n t i l around t h e t u r n o f t h e century, l i m i t e d i n t r o d u c t i o n o f denatured 233U f u e l i s p e r m i t t e d as e a r l y as 1991. A f u r t h e r argument i s t h a t commercial-scale reprocessing f o r t h e a l t e r n a t e f u e l s would n o t be f e a s i b l e u n t i l t h e backlog o f spent fuel r e q u i r e d f o r p l a n t s t a r t u p had accumulated and t h e number o f r e a c t o r s u t i l i z i n g r e c y c l e d f u e l could assure continued operation o f commercial-scale f a c i l i t i e s . On the o t h e r hand, f o r 33U-containing spent f u e l elements t o be a v a i l a b l e even f o r p i l o t - p l a n t processing, i t i s e s s e n t i a l t h a t e a r l y i r r a d i a t i o n o f thorium i n r e a c t o r s be implemented. I n Section 7.3.1

a possible procedure f o r implementing and e v e n t u a l l y commercializing t h e denatured 233U c y c l e i s discussed. Included i s a scenario which would provide f o r t h e

c bt

L

e a r l y i n t r o d u c t i o n o f thorium f u e l i n t o c u r r e n t l i g h t - w a t e r r e a c t o r s and a l l o w an o r d e r l y progression t o t h e u t i l i z a t i o n o f denatured

23%

f u e l i n breeders.

The major considera-

t i o n s i n commercializing these various r e a c t o r s operating on a l t e r n a t e fuels, and i n p a r t i c u l a r on denatured 233U f u e l , are sumnarized i n Section 7.3.2.

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7-23

L I

L

t L i

7.3.1. Possible Procedure f o r Implementing and Commercializing t h e Denatured Fuel Cycle On t h e b a s i s o f t h e above assumptions, and t h e discussion i n Section 5.1, i t i s obvious t h a t t h e o n l y r e a c t o r s t h a t could operate on denatured 233U f u e l i n t h e near term (by 1991) would be LtlRs. Two p o s s i b i l i t i e s e x i s t f o r producing 233U f o r LWRs p r i o r t o t h e i n t r o d u c t i o n o f commercial f u e l reprocessing. One involves t h e use o f "denatured 235U" f u e l (i.e., MEU(235)/Th) i n LWRs, thereby i n i t i a t i n g t h e production o f 233U. However, t h i s scheme s u f f e r s from very h i g h f i s s i l e inventory requirements associated w i t h f u l l thorium loadings i n LWRs (see Section 4.1). A second option involves t h e use o f p a r t i a l thorium loadings i n LWRs. I n t h i s o p t i o n Tho2 i s introduced i n c e r t a i n l a t t i c e l o c a t i o n s and/or MEU(235)/Th f u e l i s used i n o n l y a f r a c t i o n o f the f u e l rods, the remaining f u e l rods being conventional LEU f u e l rods.

This scheme s i g n i f i c a n t l y reduces the f i s s i l e

i n v e n t o r y penalty associated w i t h f u l l thorium loadings i n LWRs and operational b e n e f i t s as w e l l (see Section 4.1).

f o r BWRs may o f f e r

Also, the p a r t i a l thorium loadings would

a l l o w experience t o be gained on t h e performance o f thorium-based f u e l s w h i l e generating s i g n i f i c a n t q u a n t i t i e s o f 233U. E i t h e r o f t h e above options f o r producing 233U w i l l probably r e q u i r e some form of government i n c e n t i v e since t h e Us08 and separative work

L

requirements (and associated costs) w i l l increase w i t h t h e amount o f Th u t i l i z e d i n t h e once-through throwawaylstowaway modes i n LWRs.

i

Although a reprocessing c a p a b i l i t y would be r e q u i r e d t o recover the bred

33U from

thorium f u e l s , such a c a p a b i l i t y would n o t be r e q u i r e d f o r t h e q u a l i f i c a t i o n and demonstration o f thorium-based f u e l , which i n i t i a l l y would employ 235U r a t h e r than 23%.

1

L

As has been p o i n t e d o u t above, the operation o f LWRs w i t h MEU(235)/Th o r w i t h p a r t i a l thorium loadings could be accomplished during t h e n e x t decade w h i l e t h e development and demonstration o f t h e needed f u e l c y c l e f a c i l i t i e s f o r t h e implementation o f t h e denatured 233U c y c l e are pursued.

I n i t i a l l y the spent f u e l could be stored i n r e p o s i t o r i e s i n

secure f u e l storage centers which would represent a growing s t o c k p i l e o f 2 3 % and plutonium. A d d i t i o n a l f u e l c y c l e s e r v i c e f a c i l i t i e s , such as i s o t o p i c separation, reprocessing, f u e l r e f a b r i c a t i o n and p o s s i b l y waste i s o l a t i o n , could be introduced i n t o these centers as t h e need develops.

As p o i n t e d o u t above, these c o u l d i n i t i a l l y be p i l o t - p l a n t - s c a l e f a c i l i t i e s

followed by l a r g e r prototypes and then comnercial-scale p l a n t s .

It has been estimated

( i n Section 5.2) t h a t commercialization o f a new reprocessing technology would r e q u i r e 12 t o 20 yr and t h e commercialization o f a new r e f a b r i c a t i o n technology would r e q u i r e 8 t o 15 yr. With t h e deployment o f t h e p i l o t - s c a l e reprocessing and r e f a b r i c a t i o n f a c i l i t i e s , recovery o f Pu and U from spent f u e l and t h e subsequent r e f a b r i c a t i o n o f Pu/Th and denatured 233U/Th f u e l s could be demonstrated w i t h i n t h e center. Pu/Th LWRs* could then

As used i n t h i s report, *That i s , thermal transmuters o f an LWR design (see Section 4.0). a transmuter i s a r e a c t o r (thermal o r f a s t ) which burns one f u e l and produces another ( s p e c i f i c a l l y , a r e a c t o r t h a t burns Pu t o produce 233U from Th).


7-24

introduced w i t h i n the centers t o provide an a d d i t i o n a l means f o r 233U production, as well as additional power production. Concurrently, 233U (and unburned 235U) recovered from MEU(235)/Th o r from p a r t i a l thorium loadings could be u t i l i z e d i n denatured 233U fueled LWRs introduced a t dispersed sites. Later, 233U recovered from the Pu/Th fueled LWRs could also be u t i l i z e d t o f u e l dispersed reactors. A t t h i s p o i n t the f i r s t phase o f a nuclear power system t h a t includes reactors operating both i n energy centers and a t d i s persed locations outside the centers would be i n e f f e c t . During t h i s phase, which i s represented i n Fig. 7.3-la, the research and development t h a t w i l l be requirea t o deploy Pu-fueled FBR transmuters w i t h

L

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thorium blankets i n t h e energy centers LWR-LEU

Q

F

[

DENATURED LWR

Q

W R W REPROCESSING THOREX L LWWhITh REPROCESSING

E I

' '\

,e--\

LWR-LEU

\

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1

1 - 1

WREX REPROCESSING THOREX REPROCESSING F I R hRh I

b.

.

DENATURED

With these advance preparations

0 I I

INITIALPHASE

/

'--4'

could be pursued.

DENATURED LWR ADVANCED CONVERTER

1

DENE:FIED

'n

INTERMEDIATE PHASE

DENATURED F I R

W R W REPROCESSING FBR hITh f iF I N A L M W E

Fig. 7.3-1. Three Phases f o r an Evolving Energy Center.

having been made, by the time conventional LEU f u e l i n g i n LWRs begins t o phase out (due t o increasing depletion o f an economical resource base), t h e power system would evolve i n t o a fast/thermal combination employing FBR transmuters and 233U-fueled converters, which by then might include denatured LWRs and advanced converters (SSCRs , HTGRs, or HWRs), depending on the reactor(s) selected f o r development (see Fig. 7.3-lb). Such a system could provide adequate capacity expansion f o r modest energy demand growth; however, i f t h e energy demand i s such t h a t the fast/thermal system i s inadequate, an

a l l - f a s t system including denatured FBRs could be substituted as shown i n Fig. 7.3-lc. The necessity o f the t h i r d phase o f t h e energy center development i s uncertain a t t h i s time, r e f l e c t i n g as i t does assumptions concerning the supply o f economically recoverable U308 and energy demand. I t i s noted t h a t t h i s proposed scheme f o r implementing the denatured f u e l cycle and i n s t i t u t i n g the energy center concept r e l i e s h e a v i l y on two strong technical bases:

c u r r e n t l y employed LWR technology, and the research and development already expended on LMFBRs, which includes the Purex and, t o a lesser extent, the Thorex reprocessing technologies. While a1t e r n a t i v e f u e l cycle technologies o r other types o f reactors w i l l be involved i f they can be demonstrated t o have resource o r economic advantages, the LWRLMFBR scenario has been selected as representative o f the type o f a c t i v i t y t h a t would be required.

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7.3.2.

Considerations in Commercializing Reactors Operating on Alternate Fuels

Although the introduction dates cited above for commercial operation of the various reactors on alternate fuels are considered to be attainable, they can be realized only if the first steps toward commercialization are initiated in the near future under strong and sustained government support. Currently, there is little economic incentive for the private sector to proceed with such development alone. For example, while recent evaluations1s2 of LWRs have indicated the feasibility of using thorium-based fuels with current core and lattice designs, either as reload fuels for reactors already in operation or as both initial and reload fuels for future LWRs, the resource-savings benefit of such fuels relative to once-through LEU fuel cannot be realized in the absence of fuel reprocessing and refabrication services. Moreover, the introduction of thorium into the core will require high initial uranium loadings, so that the fuel costs for the core would increase. Obviously, the lack of strong evidence that fuel recycle services would be available as soon as they were needed would discourage a transition to thorium-based fuels. Alternatively, such services could not be expected to be available commercially until utilization of thorium has been established and a market for these services exists. Thus commercialization of the denatured fuel cycle in LWRs, especially within the time frame postulated in this study, is unlikely unless major government incentives are provided. The government incentives could be in the form of guarantees for investment in the fuel cycle services and/or subsidies for the costs associated with the additional U308 and separative work required for thorium-based fuels or for partial thorium loadings on the once-through cycle. This would also encourage the development of the fuel cycle services by establishing widespread use of thorium-based fuels. The commercial introduction of the required new LWR fuel cycle services could probably be accomplished by allowing a 7-yr lead time for construction of demonstration reprocessing and refabrication plants and an additional 7 yr to construct commercial-size plants. In the meantime, fabrication of MEU(235)/Th fuel or fuel designs involving partial thorium loadings for LWRs could probably be accomplished with existing LEU facilities within 2 to 3 yr (Ref. 3) with an additional 5 to 7 yr required for fuel qualification and/or demonstration. The R&D costs for demonstrating denatured uranium fuel in commercial reactors would be borne by the government. The conercial introduction in the U. S. of the advanced converter concepts (SSCRs, HTGRs, and HWRs) would be more difficult today than was the past commercial introduction of the LWR. Although the introduction in 1958 of the first LWR, the Shippingport reactor, did involve government support, a relatively small investment was required due to its size (~68 MWe). The largest base-load power plants were about 300 MWe when LWRs initially penetrated the comnercial market. Also, during the initial years of deployment of nuclear power, delays due to licensing procedures were considerably shorter, a1 lowing plants to be constructed and brought on-line more rapidly than the current 10- to 12-yr lead time. The longer time causes much larger interest payments and much greater risk of licensing difficulties.


7-26

P r i o r t o comnercial introduction, a demonstration phase o f a new advanced converter concept w i l l be required, and, as has been pointed out i n Chapter 5.1, i t i s assumed here t h a t the demonstration w i l l be on the reactor's reference cycle, which except f o r the HTGR, does not involve thorium.

U t i l i t i e s are u n w i l l i n g t o r i s k the l a r g e investment f o r

commercial-size plants o f 1000 MWe t o 1300 MWe on u n t r i e d concepts.

With the l a r g e

investments necessary f o r demonstration units, s i g n i f i c a n t government support would be required: i.e., a demonstration program i n v o l v i n g government construction o f the i n i t i a l u n i t w i t h government f i n a n c i a l support o f the f i r s t commercial-size p l a n t (1000 MWe t o 1300 MWe). For comercia1 sales t o occur, a vendor would have t o market i t and make the necessary investment t o e s t a b l i s h t h e manufacturing i n f r a s t r u c t u r e . The SSCR i s expected t o draw heavily on e x i s t i n g LWR technology, and i t may even be f e a s i b l e t o operate a conventional PWR i n the s p e c t r a l - s h i f t - c o n t r o l mode by a d d i t i o n o f c e r t a i n equipment.

The f e a s i b i l i t y o f s p e c t r a l - s h i f t - c o n t r o l has already been demonstrated While the p o s s i b i l i t y o f r e t r o f i t t i n g

i n the Belgian VULCAIN experiment (see Section 4.2).

e x i s t i n g l a r g e PWRs t o the SSC mode e x i s t s , f o r reactors going i n t o operation a f t e r the late-l980s, designing PWRs t o accept SSC c o n t r o l a t some l a t e r date i s a more l i k e l y p o s s i b i l i t y . A major impediment t o commercial i n t r o d u c t i o n o f the SSCR i n the U.S. i s l i k e l y t o be the supply o f D20 and government incentive would probably a l s o be required i n t h i s area, as i t w i l l be f o r the deployment o f the CANDU reactor (see below). The technology f o r HTGRs i s already w e l l under way, w i t h a prototype reactor c u r r e n t l y undergoing s t a r t u p t e s t i n g a t F o r t S t . Vrain. P r i o r t o commercial deployment, however, successful operation o f a demonstration HTGR i n the 1000-MWe t o 1300-MWe range would be required.

I n i t i a l l y , HTGRs could operate on the stowaway MEU(235)/Th o r LEU cycle.

Again, comercial-scale reprocessing and r e f a b r i c a t i o n f a c i l i t i e s would n o t be expected u n t i l a demonstrated market f o r such services i s present. The technology f o r HWRs i s also w e l l advanced, w i t h the CANDU reactors fueled w i t h natural uranium already comnercialized

in Canada.

It would be necessary, however, t o

demonstrate t h a t the CANDU w i t h appropriate modifications f o r s l i g h t l y enriched f u e l could be licensed i n the U.S. and produce power a t an acceptable cost. Commercialization of the CANDU i n the U.S. would probably r e q u i r e government action i n three areas:

1.

Transfer o f technology from Canada t o take advantage of CANDU r e a c t o r development and demonstrated performance. U.S.

2. 3.

A1 t e r n a t i v e l y , a demonstration u n i t designed t o

l i c e n s i n g standards would be required.

Government f i n a n c i a l support o f a l a r g e (1000-MWe t o 1300-MWe) CANDU i n the U.S. Development o f D20 production f a c i l i t i e s i n the U.S.

on a l a r g e r scale than

c u r r e n t l y exists. CANDUs operating on thorium-based f u e l s could possibly be introduced simultaneously w i t h the deployment i n the U.S. o f the CANDU reactor concept i t s e l f . Assuming Canadian p a r t i c i p a t i o n , thorium-based f u e l could be demonstrated i n Canadian reactors p r i o r t o the Furthermore, i f by then the LWR thorium f u e l

operation o f a CANDU reactor i n the U.S.


7-27

c y c l e services o f reprocessing and r e f a b r i c a t i o n had been commercially developed, t h e

.

extension o f these services t o CANDU reactors could be b u i l t on t h e e x i s t i n g LWR f a c i l i t y

.. I

b

base. Otherwise, t h e commercial i n t r o d u c t i o n o f these services could n o t be expected u n t i l some time a f t e r i t becomes c l e a r t h a t CANDU reactors w i l l be conunercially deployed

with thorium fuel, thereby i n d i c a t i n g t h e existence o f a market f o r associated f u e l c y c l e services. The i n t r o d u c t i o n dates postulated f o r t h e a l t e r n a t e f u e l c y c l e CANDUs assume t h a t r e q u i s i t e f u e l c y c l e services have already been developed f o r thoriumi n t h e U.S.

f u e l e d LWRs. As pointed o u t i n Section 5.1,

k;

L;

no attempt has been made here t o consider t h e com-

m e r c i a l i z a t i o n prospects o f FBRs since t h e INFCE program ( I n t e r n a t i o n a l Nuclear Fuel Cycle Evaluation) i s c u r r e n t l y studying t h e r o l e o f FBRs i n nuclear power scenarios and t h e i r r e s u l t s should be a v a i l a b l e i n t h e near f u t u r e . I n summary, i t i s apparent t h a t s i g n i f i c a n t b a r r i e r s e x i s t f o r the p r i v a t e sector e i t h e r t o convert LWRs t o thorium-based f u e l s o r t o develop advanced r e a c t o r concepts. While U,08 i s s t i l l r e l a t i v e l y inexpensive, the economics o f a l t e r n a t e r e a c t o r and f u e l c y c l e concepts a t best show marginal savings r e l a t i v e t o t h e LWR and consequently t h e i r

i

b

development and deployment would have t o be h e a v i l y subsidized by t h e government. I n t h e longer term, as t h e p r i c e o f uranium increases due t o d e p l e t i o n o f lower-cost uranium deposits, these a l t e r n a t e concepts could achieve superior economic performance compared t o t h e LWR.

b

The most o p t i m i s t i c i n t r o d u c t i o n dates f o r advanced converters r e s u l t i n a

r e l a t i v e l y small i n s t a l l e d capacity by t h e year 2000, and, as shown i n Chapter 6, t h e impact o f advanced converters on the cumulative U308 consumption by t h e year 2000 would be small.

However, deployment o f a l t e r n a t e r e a c t o r concepts i n t h e time from 1995-2000

could have s i g n i f i c a n t impact on resource use i n t h e p e r i o d 2000-2025.

Except f o r HTGRs,

none o f t h e a1 t e r n a t e r e a c t o r concepts t h a t promise improved resource u t i 1 i z a t i o n has undergone l i c e n s i n g review by the government.

Due t o these f a c t o r s , conversion t o the

denatured f u e l c y c l e and/or i n t r o d u c t i o n o f a l t e r n a t e r e a c t o r concepts on a time scale which can dissuade i n t e r n a t i o n a l tendencies toward conventional plutonium r e c y c l e w i l l

r e q u i r e very s i g n i f i c a n t government involvement and f i n a n c i a l i n c e n t i v e s i n t h e near future.

7.3.3.

Conclusions

From t h e above discussion t h e f o l l o w i n g conclusions can be summarized: 0

1J I

L

The production o f

233U

f o r t h e denatured 233U f u e l c y c l e could be i n i t i a t e d

by i n t r o d u c i n g Th i n t o t h e LWRs c u r r e n t l y operating on t h e once-through cycle. However, t h e r e i s an economic d i s i n c e n t i v e w i t h i n t h e p r i v a t e sector t o convert LWRs t o thorium-based fuels because o f t h e increased costs associated w i t h t h e higher U308 and separative work requirements. Thus commercialization o f t h e denatured f u e l c y c l e i s n o t p l a u s i b l e unless government incentives are provided.

I n i t i a l production o f 233U


7-28

f o r l a t e r recycle could be i n i t i a t e d by the mid-1980's i f such incentives were forthcoming.

Recycle o f 233U on a commercial scale i s not p l a u s i b l e

p r i o r t o the year 2000, however. The i n t r o d u c t i o n o f advanced reactor concepts t h a t would provide s i g n i f i c a n t resource savings beyond the year 2000 w i l l require very l a r g e government support f o r R&D, f o r demonstration f a c i l i t i e s , and f o r lead comnercial

L

plants. I f a r a p i d deployment schedule were required, a d d i t i o n a l resources would have t o be committed t o cover the r i s k s o f e a r l y commercial plants. Fuel service/energy centers whose u l t i m a t e purpose i s t o u t i l i z e plutonium both f o r energy production and f o r 233U production would progress through various phases.

I n i t i a l l y these centers would be f u e l storage f a c i l i t i e s .

With the

i n t r o d u c t i o n o f reprocessing and r e t a b r i c a t i o n i n the center, LWRs located a t dispersed s i t e s would be fueled w i t h denatured 233U.

Concurrently Pu-fueled

thermal transmuters would be deployed w i t h i n the center.

L

Ultimately, t o meet

long-term energy demands, Pu-fueled f a s t transmuters would be introduced w i t h i n the centers. It i s desirable t h a t a f u e l recycle R&D program be i n i t i a t e d f o r denatured f u e l s a t the same time a decision i s made t o f a b r i c a t e thorium-containing fuel f o r large-scale i r r a d i a t i o n i n e x i s t i n g LWRs. P i l o t - s c a l e recycle f a c i l i t i e s could be required w i t h i n seven years a f t e r the i n i t i a t i o n o f a thorium i r r a d i a t i o n program.

Section 7.3 References

1.

N. L. Shapiro, J. R. Rec, R. A. Matzie (Combustion Engineering), "Assessment o f Fuel Cycles i n Pressurized Water Reactors," EPRI-NP-359 (February 1977).

2.

"Assessment o f U t i l i z a t i o n o f Thorium i n BWRs," ORNL/SUB-4380/5 (NEDG-24073), prepared by General E l e c t r i c Company (January 1978).

3.

"The Economics and U t i 1i z a t i o n o f Thorium i n Nuclear Reactors,'' ORNL-TM-6331 (also Technical Annexes 1 and 2, ORNL-TM-6332) prepared by Resource Planning Associates, Inc. (May 1978).

c c t' t

t


7-29

7.4.

ADEQUACY OF NUCLEAR POWER SYSTEMS UTILIZING DENATURED 2 3 3 U FUEL FOR MEETING ELECTRICAL POWER DEMANDS M. R. Shay, D. R. Haffner, W. E. Black, T. M. Helm, R. W. Hardie, and R. P. Omberg Hanford Engineering Development Laboratory

An important measure f o r evaluating a nuclear power system i s whether i t can meet p r o j e c t e d power demands w i t h t h e uranium resources estimated t o be a v a i l a b l e a t an acceptable cost.

This s e c t i o n summarizes the r e s u l t s of analyses performed i n t h i s study t o

determine whether various nuclear power systems u t i 1 i z i n g denatured 2 3 3 U f u e l could meet a p r o j e c t e d power demand o f 350 GWe i n s t a l l e d capacity by t h e year 2000 and a n e t increase o f 15 GWe/year through t h e year 2049, t h e t o t a l capacity i n the year 2050 being 1100 GWe. The analyses were based on a uranium supply model shown i n Fig. 7.4-1 and i n Table B-7 (Appendix B) , which provides both conservative and o p t i m i s t i c p r e d i c t i o n s o f t h e uranium supply as a f u n c t i o n o f cost. The power systems analyzed are described i n d e t a i l i n Chapter 6.

They a r e comprised

o f LEU-LWRs operating i n conjunction w i t h LWRs on o t h e r f u e l cycles o r i n conjunction w i t h one o f t h e three types o f advanced converters (SSCR, HWR, o r HTGR) considered i n t h e study. I n some cases, FBRs a r e included i n the system.

Since t h e maintenance o f p r o l i f e r a t i o n -

r e s i s t a n t power systems was one of the primary concerns, t h e concept o f a secure energy center supporting dispersed reactors was

ORNL-DWG 78-24717

260 240

200

-

I

I

I

I

I

I

I

I

I

I

and denatured fuels.

-

-

used, w i t h the f u e l u t i l i z e d i n t h e d i s persed reactors r e s t r i c t e d t o LEU ( o r SEU)

A r e a c t o r operating on

the denatured 233U f u e l c y c l e i s n o t s e l f sustaining, however, and t h e r e f o r e i t requires an exogenous source o f

23%.

In

the power systems studied, the 2 3 3 U i s

provided by MEU/Th-fueled thermal reactors

-

-In-

-

.

1 t

' 2

'

3

-

Because t h e transmuters have plutonium cores ,

u30e

1

4

1

5

OUANTITY

1

6

(to6

1

7

'

8

"

9

-

These reactors, o f course,

a1 so c o n t r i b u t e t o t h e power generation.

0

transmuters.

-

0

o r p l u t o n i um-fuel ed thermal and/or f a s t

however, they must be l o c a t e d w i t h i n t h e secure energy centers.

(Note:

With t h i s r e s t r i c t i o n

t h e "energy support r a t i o ' ' o f a nuclear system becomes a second important measure o f evaluation, as i s discussed i n Section

+ O t !

tons)

7.2.3.

The energy support r a t i o s f o r t h e

systems described here are given i n Appendix Fig. 7.4-1. Marginal Costs f o r Highand Intermediate-Cost U308 Supply Curves.

C, along w i t h o t h e r d e t a i l e d r e s u l t s from

the analyses. )


7-30

A nuclear power systems evaluation such a s the one performed i n this study requires three basic components. First, the various nuclear power systems to be analyzed must be identified. Second, there must be an analytical model capable of modeling each system i n sufficient detail that differences between the systems can be accurately calculated. And finally, a data base that contains both reactor performance data and economic data must be developed. Sections 7.4.1 and 7.4.2 below give brief descriptions of the model and data base as they were applied to this evaluation. The results of the analyses f o r specified nuclear power systems a r e then sumnarized i n Sections 7.4.3, 7.4.4 and 7.4.5, w i t h the detai 1ed resul t s presented In Appendi x C .

7.4.1.

c I:

The Analytical Method

Two fundamental aspects of the model used i n the analyses r e l a t e t o the nuclear energy demand and the U308 supply, both of which have been specified above. The nuclear energy demand assumed i n the model i s consistent with the current construction plans of utilities through the 1980's. As more nuclear units were required, w i t h the supply of low-cost U308 progressively depleted, i t was assumed t h a t more expensive lower-grade uranium resources would be mined. T h i s was modeled by assuming t h a t the long-run marginal cost of U308 was an increasing function o f the cumulative amount mined. For a particular nuclear pol icy option, the plant construction pattern was therefore governed by economics and/or uranium utilization.

L L

Two different optimizing patterns were used i n the study. In the f i r s t runs economic competition between nuclear fuels and coal was assumed, and the plants were selected t o minimize the levelized cost of power over time. These runs, which a r e presented i n Appendix 0, indicated t h a t nuclear power did not compete well a t U3O8 prices above $160/lb f o r the assumptions used i n t h i s study. Thus f o r the runs of all-nuclear power systems, described i n Chapter 6 and sumnarized here, an attempt was made t o s a t i s f y the demand for nuclear power w i t h the U3O8 available a t a price less than $160/lb U308. Other considerations also affected the selection of the nuclear power plants to be constructed. For example, a reactor t h a t required Pu or 233U could not be constructed unless the projected supply of f i s s i l e material was sufficient throughout the reactor's lifetime. In addition, a nuclear plant design t h a t differed from established technology could be introduced only a t a limited rate. Furthermore, once the manufacturing capability to produce a particular reactor type was well established, the maximum r a t e a t which t h a t reactor type could lose i t s share of the new construction market was limited t o a specified rate. Both the total power cost of each nuclear policy option and the total power cost o f each reactor type available i n each option were calculated. For each reactor type, the total power cost was calculated f o r four components -- capital, operation and maintenance,

L


7-31

L i '

hd -

.-

I '

i

taxes, and f u e l cycle.

The f u e l c y c l e costs were, i n turn, d i v i d e d i n t o seven components

23 %, u r a n i um , t h o r i um, enrichment , p l u t o n i um , f a b r i c a t i o n , and reprocessing

.

--

I t i s t o be noted t h a t t h e power systems c a l c u l a t e d were a l l assumed t o be U.S. based, t h e i n p u t data a l l being o f U.S. o r i g i n . With appropriate i n p u t modifications, however, t h e model could be used f o r o t h e r scenarios. For example, i t could be used t o

analyze t h e p o t e n t i a l f o r t h e deployment o f transmuters both t o produce power i n secure s t a t e s and t o produce 2 3 % f o r export t o s t a t e s wishing t o base t h e i r own power systems on thermal r e a c t o r s w i t h o u t n a t i o n a l reprocessing. 7.4.2.

t L

Data Base

The data r e q u i r e d by t h e model f o r each r e a c t o r type i n c l u d e power l e v e l , annual i s o t o p i c charge and discharge, annual f a b r i c a t i o n requirements, i n t r o d u c t i o n dates, etc. These data are presented i n Tables 6.1-2 and 6.1-3 i n Chapter 6. It i s t o be pointed out, however, t h a t t h e data are f o r reactors o f e s s e n t i a l l y conventional designs, and t h a t the

up8

requirements f o r t h e various r e a c t o r types could be reduced through design optimiza-

t i o n . (Note: The e f f e c t o f o p t i m i z i n g LWRs has been considered i n a separate analysis and i s discussed i n Section 7.4.3 below.). The major parameters i n t h e economic data base used f o r t h i s study are c a p i t a l costs,

i'

uranium costs, f a b r i c a t i o n costs, spent f u e l disposal costs, reprocessing costs, and money

b

costs. The e n t i r e data base, which was developed i n a j o i n t e f f o r t i n v o l v i n g government and i n d u s t r y representatives, i s presented i n Appendix B. 7.4.3.

i -'

b

Results f o r Price-Limited Uranium Supplies

As noted above, t h e denatured nuclear power systems u t i 1 i z e d various combinations of thermal converters and f a s t reactors.

These i n t u r n were examined under s i x fuel cycle options, which a r e summarized i n Table 7.4-1 (Options 4-8). I n a d d i t i o n , t h e same r e a c t o r types were examined under t h r e e reference f u e l c y c l e options

-- a throwaway/stowaway

(Option 1) and two plutonium-uranium options (Options 2 and 3).

option

Four cases were considered

under each option, each case being d i s t i n g u i s h e d by t h e type o f converter being emphasized

LWRs, SSCRs, HWRs, o r HTGRs.

Thus a t o t a l o f 36 d i f f e r e n t nuclear power systems were

analyzed. The maximum nuclear capacity and t h e year i n which t h e maximum occurs f o r each nuclear system s t u d i e d i s shown i n Table 7.4-2 f o r the two uranium supply assumptions (see

L I,

Fig. 7.4-1).

As s t a t e d e a r l i e r , w i t h t h e intermediate-cost supply i t was assumed t h a t 6

m i l l i o n ST o f U308 could be recovered a t costs l e s s than $160/lb, w h i l e w i t h t h e high-cost supply i t was assumed t h a t 3 m i l l i o n ST o f u $ 8 would be available.

--


7-32

Table 7.4-1.

Description of Fuel Cycle Options*

Throwaway/Stowaway Option (see Fig. 6.1-1): Option 1.

LEU converters on once-through cycle.

Plutonium-Uranium Options (see Fig. 6.1-2): Option 2.

Pu/U recycle option; LEU converters outside center; Pu/U converters i n s i d e center; HTGRs i n s i d e center operate on 235U/Th, 233U/Th, and Pu/Th.

Option 3.

Pu/U recycle option; LEU converters outside center; Pu/U converters and breeders i n s i d e center; HTGRs i n s i d e center operate on 235U/Th, 233U/Th, and Pu/U.

Denatured Uranium Options Using Converters Only (see Fig. 6.1-3): Option 4.

Plutonium throwaway option; LEU and denatured 2 3 5 U and 2 3 3 U converters outside center; no reactors i n s i d e center; U only recycled.

Option 5U.

Plutonium minimization option; LEU and denatured 235U and 233U converters outside center; Pu/Th converters i n s i d e center; U and Pu recycl ed.

Option 5T.

Same as 5U without denatured 235U converters.

c

L I: L

Denatured Uranium Options Using Both Converters and Breeders (see Fig. 6.1-4): Option 6.

L i g h t transmutation option; LEU and denatured 235U and 233U convert e r s outside center; Pu/Th converters and Pu-U/Th breeders i n s i d e center.

Option 7.

L i g h t transmutation option w i t h denatured breeder; LEU converters, denatured 23sU converters, and denatured 233U converters and breeders outside center; Pu/Th converters and Pu-U/Th breeders i n s i d e center.

Option 8.

Heavy transmutation option; same as Option 7 except i n s i d e breeder i s a Pu-Th/Th breeder.

*Four cases considered under each option, i d e n t i f i e d by l e t t e r s L, S, H, and G t o denote type o f converter employed i n a d d i t i o n t o LEU-LWRs (L = LWR, S = SSCR, H = HWR, G = HTGR).

6:

L

The e f f e c t o f varying the f u e l cycle system can be seen by reading across Table 7.4-2 and the e f f e c t o f changing the converter reactor option can be deduced by 'reading down a column. An i n s t a l l e d nuclear canacity o f 1100 GWe i n year 2050 indicates t h a t the projected

I:

energy demand is f u l l y met by tne reactors i n a given nuclear f u e l cycle system.

c


7-33 x

ibd

Table 7.4-2. (Note:

Maximum Nuclear Capacity o f Various Nuclear Power Options and Year i n Which Maximum Occurs

A capacity of 1100 GWe i n year 2049 meets demand.)

Maximum I n s t a l l e d Nuclear Capacity (GWe)/Year rnmimwn occurs Converter Reactor Opt ion

1

2

3

4

5u

5T

6

7

8

With High-Cost U308 Supply 433

611

1100

585

716

637

1100

1100

1087

2009

2022

2049

2029

2027

2022

2049

2049

2049

440

661

1100

660

820

764

1100

1100

1084

2009

2023

2049

2023

2053

2029

2049

2049

2049

444

630

1100

756

915

856

1100

1100

1100

2021

2022

PO49

2031

2041

2035

2049

2049

2049

437

818

1100

545

671

638

1091

1100

958

2009

2033

2049

2019

2023

2021

2049

2049

2041

1

With Intermediate-Cost U308 Supply

..

I

i,

L

HTGRs (GI

729

968

1100

1002

1062

1012

1100

1100

1097

2027

2041

2049

2047

2049

2047

2049

2049

2049

763

1078

1100

1084

1100

1100

1100

1100

1100

2029

2049

2049

2049

2049

2049

2049

2049

2049

852

1062

1100

1084

1100

1100

1100

1100

1100

2035

2049

2049

2049

2049

2049

2049

2049

2049

783

1100

1100

971

1065

996

1100

1100

1100

2031

2049

2049

2041

2049

2045

2049

2049

2049

-Non-FBR Systems, Options 1, 2, 4, and 5 '

For t h e high-cost U308 supply case ( 3 m i l l i o n ST U308 below $160/lb),

i t i s evident

t h a t i n t r o d u c i n g advanced converters on t h e throwawayfstowaway f u e l c y c l e (Option 1) has l i t t l e e f f e c t on t h e maximum a t t a i n a b l e nuclear capacity. This i s d i r e c t l y due t o t h e i

t

i n t r o d u c t i o n dates assumed f o r t h e advanced converter reactors.

By t h e time t h e converters


,

c

7-34

dominate the new capacity being b u i l t , a very s i g n i f i c a n t f r a c t i o n of the U30, supply has already been committed t o the standard LWR. I t follows t h a t i f the intermediate-cost U308 were used (6 m i l l i o n ST U30a below $160/lb),

together w i t h the same nuclear growth rate, the a d d i t i o n o f an advanced converter would have a much l a r g e r impact. For example, i n t h i s case the system including HWRs has a maximum a t t a i n a b l e i n s t a l l e d nuclear capacity f o r the throwaway/

c

stowaway option t h a t i s approximately 17% greater than the i n s t a l l e d capacity o f the system comprised o f LWRs alone, while f o r the high-cost supply case i t i s only 3% greater. I n Option 2 converter reactors are operated on the LEU f u e l cycle outside the energy center and Pu/U converters and 235U(HE)/Th, 23%/Th, and Pu/Th HTGRs are operated i n s i d e

As expected, the thermal recycle systems a l l support nuclear power growth w i t h l e s s U308 consumption than the once-through systems o f Option 1, and, i n general, the options including advanced converter reactors (SSCRs, HWRs , and HTGRs) provide f o r increased 'maximum i n s t a l l e d capacity r e l a t i v e t o the LWR option for both the high-cost and the intermediate-cost U p , supply assumptions. The HTGR option (26) provides f o r the greatest l e v e l o f i n s t a l l e d nuclear capacity f o r both U308 supplies. The resource e f f i c i e n c y o f these scenarios i s l a r g e l y due t o the f a c t t h a t they include the nondenatured 23%/Th f u e l cycle which i s used only by HTGRs i n t h i s study. the center.

c L

Option 4 u t i l i z e s only denatured 235U and 233U f u e l s and LEU fuel, a l l outside the energy center, and none o f the plutonium produced i s recycled.

Here i t i s i n t e r e s t i n g t o

observe t h a t f o r both uranium supply assumptions the HWR converter option (4H) has i n s t a l l e d capacity l e v e l s t h a t are greater than o r equal t o those o f any other converter reactor option, while the HTGR option (46) has the lowest i n s t a l l e d capacities. I t appears t h a t the HTGRs used i n t h i s study do n o t operate e f f i c i e n t l y on denatured f u e l cycles r e l a t i v e t o the other converters available (see also Options 5UG and 5TG). This can be p a r t i a l l y a t t r i b u t e d t o the fact t h a t the reactors used i n these evaluations were n o t optimized f o r the r o l e s i n which they were employed, and for the HTGR t h i s has a greater impact than f o r the other reactor types. Option 5 uses denatured and LEU-fueled reactors outside the center and Pu/Th-fueled converters w i t h i n the center.

This option i s divided i n t o two suboptions: Option 5U, i n which both denatured 235U and denatured 2 3 % u n i t s are used; and Option 5T, i n which the

Li

denatured 235U u n i t s are excluded.

c

I n both cases, 2 3 % i s produced i n the Pu/Th converters. I n these cases the HWR options produce the greatest maximum i n s t a l l e d nuclear capacity

w i t h the high-cost ore supply, and both the HWR options and SSCR options meet the power demand w i t h the intermediate-cost ore supply.

Again, the HTGRs do not appear t o operate as

e f f i c i e n t l y as the other converters for the reasons c i t e d above. I n summary, non-FBR power systems using denatured f u e l but discarding plutonium require about the same amount of U3O8 as thermal systems on the c l a s s i c a l Pu/U cycle and o f f e r p o t e n t i a l nuclear growth rates t h a t are roughly the same.

I f the plutonium i s re-


7-35

cycled i n Pu/Th converters, t h e systems have p o t e n t i a l nuclear growth r a t e s t h a t exceed those o f analogous r e a c t o r s operating on the Pu/U f u e l cycle. I f the intermediate-cost U308 supply assumption proves t o be correct, advanced converters i n t h e r e c y c l e mode can s a t i s f y t h e postulated nuclear energy demand through year 2050 a t competitive costs. This a n a l y s i s therefore i n d i c a t e s that, a t l e a s t under o p t i m i s t i c resource conditions, advanced converters using denatured fuels can d e f e r t h e need f o r commercial use o f an "inexhaustible" energy source (such as FBRs) beyond t h e year 2050. FBR Systems, Options 3 , 6, 7, and 8 Table 7.4-2 shows t h a t almost a l l o f .the nuclear system options using FBR f u e l cycles (Options 3, 6, 7, and 8) are able t o meet t h e p r o j e c t e d nuclear energy demand w i t h o u t mining U308 c o s t i n g more than $160/lb.

The o n l y exception i s Option 8 f o r t h e case o f t h e high-cost o r e supply, and even t h i s option, which includes t h e Pu-Th/Th breeder and t h e denatured

233U

breeder, would s a t i s f y t h e demand i f s l i g h t l y improved FBR

r e a c t o r design parameters were used, Thus, as was expected, t h i s a n a l y s i s i n d i c a t e s t h a t FBR-containing systems w i l l p o t e n t i a l l y support much l a r g e r nuclear c a p a c i t i e s than thermal r e c y c l e systems and/or w i l l r e q u i r e l e s s mining.

The Th-containing FBR cycles

supporting dispersed denatured converters perform as w e l l as t h e analogous Pu/U cycles w i t h i n t h e framework o f t h i s analysis.

O f t h e Th-containing cycles, t h e FBR with a Pu/U core and Th blanket i s p a r t i c u l a r l y r e s o u r c e - e f f i c i e n t .

7.4.4.

Results f o r Unconstrained Resource A v a i l a b i l i t y

The preceding r e s u l t s represent a somewhat a r t i f i c i a l s i t u a t i o n because o f t h e $160/lb l i m i t a t i o n on t h e U308 a v a i l a b i l i t y .

That i s , t h e f a i l u r e t o meet t h e p r o j e c t e d

power demand i n many o f t h e scenarios i n v e s t i g a t e d i s a d i r e c t r e s u l t o f t h e system's i n a b i l i t y t o u t i l i z e U308c o s t i n g more than $160/lb.

I n order t o address t h e p o t e n t i a l

o f t h e various f u e l c y c l e / r e a c t o r options under t h e c o n d i t i o n t h a t t h e p r o j e c t e d demand for nuclear power must be s a t i s f i e d , t h e $160/lb c o n s t r a i n t was removed. The cumulative q u a n t i t y o f U308 r e q u i r e d t o completely s a t i s f y t h e demand f o r nuclear generating c a p a c i t y was then estimated f o r each o f t h e nuclear power options; these r e s u l t s are presented i n Table 7.4-3. The r a t e a t which U308 i s r e q u i r e d t o support t h e p r o j e c t e d nuclear capacity represents an important a d d i t i o n a l c o n s t r a i n t on a system.

An o v e r a l l maximum U308

production r a t e i s d i f f i c u l t t o s p e c i f y because o f t h e p o s s i b i l i t y o f importing U308 and because any p r e d i c t i o n o f t h e production o f U308 from u n c e r t a i n resources i n t h e next century i s h i g h l y speculative.

Recognizing t h i s , and a l s o recognizing t h a t t h e r e q u i r e d

U308 production r a t e i s s t i l l an important variable, t h e maximum r e q u i r e d U308 production r a t e s f o r each scenario were estimated and a r e tabulated i n Table 7.4-4.

As a p o i n t o f

reference, note that DOE has estimated t h a t domestic mining and m i l l i n g could s u s t a i n a production r a t e o f 60,000 ST o f U308 per year i n t h e 1990s by developing U308 reserves and p o t e n t i a l resources a t forward costs

*

*

o f l e s s than $30 per pound.

Forward costs do n o t i n c l u d e t h e c a p i t a l costs o f f a c i l i t i e s o r i n d u s t r y p r o f i t s , which a r e included i n t h e l o n g r u n marginal costs used i n t h i s study.


7-36

Table 7.4-3. Cumulative U308 Consumption of Various Nuclear Policy Options -

Converter Reactor Option

Cumulative U308 Consumption (mi I1 ions o f tons) Through year 2025/Through year 2049

1

2

3

4

5u

5T

6

7

8

L

With High-Cost U308 Supply ~~

3.41

2.39

2.14

2.87

2.36

2.36

2.18

2.14

2.29

5.41

4.83

4.94

2.82

2.83

2.86

7.05

5.23

2.73

3.26

2.23

1.99

2.70

2.35

2.14

1.93

1.93

2.07

6.52

4.35

2.70

4.65

3.86

3.86

2.69

2.69

2.83

3.10

2.72

2.29

2.50

2.16

2.14

2.25

2.21

2.29

5.58

4.64

2.70

4.36

3.27

3.77

2.61

2.55

2.87

3.23

2.19

1.97

2.58

2.32

2.34

2.15

2.12

2.32

6.26

4.04

2.75

5.13

4.43

4.94

2.70

2.68

3.18

L

With Intermediate-Cost U3O8 Supply 3.41

2.39

2.28

2.87

2.36

2.36

2.37

2.37

2.37

4.38

4.38

4.48

7.05

5.23

4.40

5.41

4.91

4.94

3.26

2.23

2.20

2.70

2\14

2.14

2.14

2.14

2.14

6.52

4.35

4.14

4.65

3.86

3.86

3.86

3.86

3.86

3.10

2.72

2.31

2.94

2.52

2.51

2.32

2.30

2.38

5.58

4.64

2.71

5.40

4.32

4.37

3.66

2.70

3.37

3.23

2.32

2.30

2.58

3.32

2.34

2.23

2.23

2.26

6.26

4.23

4.22

5.13

4.43

4.94

4.19

4.19

4.24

The results presented in Tables 7.4-3 and 7.4-4 iqdicate the relative resource efficiencies of the various nuclear power systems since the energy produced was held constant. It should be noted that although the U308 cost limitation of $160/lb was removed, the uranium requirements were estimated for both the intermediate- and high-cost U308 supplies. Hence, the differences in the cumulative U308 requirements and annual U308 production rates for similar fuel cycle/reactor combinations are due to different reactor mixes associated with each uranium price structure.

I] f -â&#x20AC;&#x2122;

b

bâ&#x20AC;&#x2122;

t

c


7-37

Table 7.4-4. Maximum u308 Requirements o f Various Nuclear Pol i c y Options

L

Maximum u308 Consumption (thousand2 o f tons per year) Converter Reactor Option

1

2

3

4

5u

5T

6

7

~~

8 ~~

With High-Cost U308 Supply LWRs

183

120

60

111

115

115

62

60

68

SSCRs

160

115

52

83

83

83

50

50

55

HWRs

120

83

66

78

62

69

64

63

65

HTGRs

140

82

53

105

96

115

61

60

65

1'

b I_-

W th

1termed.ate-Cost U308 Supp Y

LWRs

183

120

92

111

117

115

86

86

92

SSCRs

160

115

93

83

83

83

83

83

83

HWRs

120

83

66

110

89

90

66

66

66

HTGRs

140

86

86

105

96

115

87

87

87

,-

i

b

--

I '

b

L

S a t i s f y i n g the demand f o r 1100 GWe i n year 2050 w i t h the standard LWR once-through cycle (Option 1L) would r e q u i r e t h a t about 183,000 ST U308 be produced i n year 2049, w i t h a cumulative consumption o f 7.1 m i l l i o n ST through t h a t date. Introducing advanced converters (Options l S , lH, and 1G) would reduce both the cumulative U308 consumption and the maximum production r a t e requirements on the once-through cycle HWR as low as 5.6 m i l l i o n

- in

the case o f the

ST and 120,000 STlyr, respectively.

Thermal recycle modes (Options 2, 4, 5U, and 5T) would reduce U308 consumption through year 2049 t o w i t h i n the range o f 3.3 t o 5.4 m i l l i o n ST U308, depending on the p o l i c y option chosen and t o a lesser extent on the uranium cost l e v e l . consumption would vary from 62,000 t o 120,000 ST/yr.

The maximum U308

The resource consumption i s sensi-

t i v e t o the uranium p r i c e l e v e l t o the extent t h a t high-cost uranium favors the choice o f e f f i c i e n t high-capital-cost systems such as the HWR, whereas lower-cost uranium favors continued use o f LWRs even i f other reactors are available. I t should be noted t h a t when plutonium i s recycled i n thermal power systems includ-

i n g denatured reactors (Options 5U and 5T) the t o t a l resource requirements ( i n c l u d i n g Pu)


7-38

are generally less than those for thermal systems in the Pu-U recycle mode (Option 2). Discarding Pu from the recycle of denatured thermal systems (Option 4) reduces the efficiency of the denatured cycle. The nuclear power systems that include fast breeders (Options 3, 6, 7, and 8) have cumulative U308 requirements through year 2049 within the range of 2.71 to 4.41 million ST U308 in the case of the intermediate-cost u& supply and within 2.6 to 3.2 million ST U& in the case of the high-cost supply. The maximum U308 consumption varies from 66,000 to 93,000 ST/yr for the intermediate-cost supply and from 52,000 to 68,000 STlyr for the high-cost supply. The breeder-containing options are able to .adjust the reactor mix effectively to reduce Ug8 consumption in the event u& costs are high. The larger the fraction of breeders in the reactor mix, the lower the U308 requirements. It should be noted that the U & requirements for the systems containing breeders with Pu/U cores and Th blankets (Options 6 and 7) are similar to the U308 requirements for the system containing the classical Pu/U breeder (Option 3). The systems containing breeders with Pu/Th cores and Th blankets require somewhat more U& on an integrated basis.

The U308 requirements presented in Table 7.4-4 qualitatively support the ranking of cycles in the cost-constrained runs. Specifically, the power systems operating on oncethrough cycles require 5.6 to 7.1 million ST U308 to satisfy the demand for nuclear power through 2050, the thermal-recycle systems require 3.3 to 5.4 million ST U308,and the breeder-containing systems require 2.6 to 4.4 million ST U308. The systems including denatured 23 5 reactors require approximately the same cumulative amount of U308 as their Pu/U counterparts. The results presented in Table 7.4-5 also suppovt these statements: the required production rates are highest for the once-through systems; they are reduced somewhat for the thermal recycle cases; and they are lowest for the breeder-containing scenarios. 7.4.5.

Systems Employinq Improved LWRs and Enrichment Technology

While not considered in the analysis summarized above, it is possible to optimize LWR designs to greatly enhance their utilization o f U308 per unit energy produced. These optimized designs may result in reduced U308 requirements o f u p to 30% relative to more conventional LWR designs. The 30% improvement in LWR U308 requirements assumes no spent fuel reprocessing, the improvements being the result o f increased discharge exposure fuels and/or reconfigured reactor cores. The effect of developing these LWR cores optimized for throwaway/stowaway operation was examined by assuming that the U308 utilization would be improved in sequential increments U308 requirements equal to 90% of the standard LWR. It was also assumed that this improvement would be retrofitted into existing reactors.t Similarly, reactors starting up tNeither the down time required for retrofitting nor the associated costs were addressed in this analysis.


7-39

between 1991 and 2001 were assumed t o have U&,

requirements equal t o 80% o f t h e standard

LWR, w i t h t h e improvements r e t r o f i t t e d t o a l l e x i s t i n g reactors a t t h a t time.

Finally,

those p l a n t s beginning operation a f t e r 2001 were assumed t o have U308 requirements equal t o 70% o f t h e standard LWR, again w i t h t h e improvements r e t r o f i t t e d t o e x i s t i n g plants. I n a d d i t i o n , t h e e f f e c t o f a lower enrichment t a i l s assay was examined f o r both t h e

standard and t h e optimized LWR designs.

The standard enrichment t a i l s schedule assumed

t h a t the assay f r a c t i o n was a constant 0.0020.

The reduced t a i l s schedule began a t 0.0020

b u t decreased t o 0.0005 between 1980 and 2010 and remained constant t h e r e a f t e r . . . a

The l a t t e r t a i l s schedule was assumed t o represent a changeover t o an improved enrichment technology.

4 The e f f e c t s o f considering both t h e improved LWR design and t h e improved t a i l s technology a r e sumnarized i n Table 7.4-5.

The r e s u l t s show t h a t w i t h t a i l s improvements

alone the U308 requirements may be reduced by 16% by year 2029. This reduced l e v e l o f U308 consumption t r a n s l a t e s t o an increase i n the maximum i n s t a l l e d capacity o f approximately 60 GWe f o r standard LWRs on t h e throwaway/stowaway f u e l cycle. Comparison o f U308 U t i l i z a t i o n o f Standard and Improved Table 7.4-5. LWRs Operating on Throwaway/Stowaway Option With and Without Improved Tai 1s ST U308/GWe Standard LWR Technology Year

Normal Tai 1s

Improved Tai 1s

Improved LWR Technolow Normal Improved Tails Tai 1s

I -

1 '

u

c1 Y

u L

1989

5236

4759

4649

4224

2009

6236

2029

5236

4508 4398

4079 3923

3560 3346

*Normal t a i l s assume 0.2 w/o 235U i n 238U; improved t a i l s assumed 0.05 w/o 235U i n 238U; 75% capacity f a c t o r .

With improved LWR technologies (no t a i l s improvements) t h e U308 consumption l e v e l s could be reduced *25% i n year 2029. This t r a n s l a t e s t o an increase o f 100 GWe i n t h e maximum i n s t a l l e d capacity f o r optimized LWRs. I f both reduced t a i l s and advanced LWR technologies were used, t h e maximum achievable i n s t a l l e d nuclear capacity would increase by about 144 GWe. I t i s important t o place these r e s u l t s w i t h i n t h e perspective o f t h e r e s u l t s r e ported i n Table 7.4-2.

The maximum i n s t a l l e d nuclear c a p a c i t i e s obtained w i t h these

improvements a r e comparable t o those f o r standard LWRs operating on t h e c l a s s i c a l P u / ~ ~ * U r e c y c l e mode o r on t h e denatured r e c y c l e were a v a i l a b l e ,

23%

cycle.

Obviously, ifboth improved LWRs and Pu

the nuclear capacity could be even greater.


7-40

7.4.6

Conclusions

From t h e preceding discussion and t h e r e s u l t s presented i n Chapter 6 and Appendix C, t h e f o l l o w i n g conclusions may be drawn concerning t h e r e a c t o r options, t h e fuel c y c l e options, and t h e U308 supply cases analyzed f o r t h i s study.

It should be emphasized t h a t

the conclusions are t e n t a t i v e and may be changed as a r e s u l t o f d i f f e r e n t demand growth p r o j e c t i o n s o r more accurate o r improved r e a c t o r characterizations.

I f nuclear power systems were l i m i t e d t o t h e once-through cycle, i t would be necessary t o u t i l i z e U308 sources a t above $160/lb sometime between y e a r 2009 and year 2035 i n order t o s a t i s f y t h e p r o j e c t e d nuclear power capacity demand.

I f nuclear power systems were l i m i t e d t o t h e once-through c y c l e and t o U& supplies below $160/lb,

t h e U.S. nuclear power capacity would peak some time

between 2009 and 2035.

Nuclear power would f a i l t o s a t i s f y t h e p r o j e c t e d

nuclear demand d u r i n g t h e 10-year p e r i o d preceding the peak.

I f improved

LWR

designs and improved t a i l s s t r i p p i n g techniques were implemented, t h e peaks would occur 10 t o 15 years l a t e r . I f t h e high-cost U308 supply i s assumed (3 m i l l i o n ST below $160/lb),

all

once-through systems, regardless o f t h e converter type employed, r e s u l t i n approximately t h e same maximum i n s t a l l e d nuclear capacity. U

For l e s s - r e s t r i c t i v e

supply assumptions, advanced converters have time t o increase t h e t o t a l

nuclear power supply on t h e once-through cycle. Thermal r e c y c l e systems have t h e c a p a b i l i t y of s u b s t a n t i a l l y reducing requirements f o r U308 o r o f increasing t h e maximum i n s t a l l e d capacity over t h e capacity o f t h e once-through cycle.

The best thermal r e c y c l e systems can support over t w i c e t h e maximum i n s t a l l e d capacity o f t h e once-through cycle, and, under t h e intermediatec o s t U308 supply assumption ( 6 m i l l i o n ST below $160/lb),

they can f u l l y support

t h e assumed nuclear power growth through year 2050. The systems i n c l u d i n g breeders have t h e c a p a b i l i t y o f s u b s t a n t i a l l y reducing t h e mining requirements and/or increasing t h e maximum i n s t a l l e d capacity beyond thermal systems w i t h recycle.

This c a p a b i l i t y i s needed t o s a t i s f y t h e nuclear capacity

demand through year 2050 under t h e high-cost uranium supply assumption ( 3 m i l l i o n ST below $160/lb). Thermal r e c y c l e systems i n c l u d i n g denatured 233U r e a c t o r s have the c a p a b i l i t y o f supporting more nuclear capacity than t h e thermal P U / ~ ~r e~c yUc l e systems. However, achieving t h i s c a p a b i l i t y would u s u a l l y r e q u i r e Pu u t i l i z a t i o n . From a resource u t i l i z a t i o n p o i n t o f view, nuclear power systems u t i l i z i n g denatured 2330 r e a c t o r s can be s t a r t e d e q u a l l y w e l l w i t h MEU(235)/Th o r Pu/Th fuels, p r o v i d i n g t h e eventual use o f t h e plutonium generated i n t h e MEU(235)/Th c y c l e i s assumed.


7-41

Systems t h a t use breeders (i.e.,

L

f a s t transmuters) t o produce 233U f o r LWRs o r

advanced converters operating on denatured 233U f u e l have a c a p a b i l i t y comparable t o systems employing t h e c l a s s i c a l Pu/U breeder c y c l e t o s a t i s f y t h e assumed demand through 2050 w i t h t h e U308 resource base assumed i n t h i s study. Section 7.4.

1.

e, I

I '

lb

t

References

John Klemenic, Director, and David Blanchfield, Mineral Economist, Supply Analysis D i v i s i o n , Grand Junction Office, DOE Uranium and Enrichment Division, i n paper e n t i t l e d "Production C a p a b i l i t y and Supply," paper presented a t Uranium I n d u s t r y Seminar, October 26-27, 1977, Grand Junction, Colorado; proceedings pub1 ished as GJO-l08( 77).


7-42

7.5.

TRADE-OFF ANALYSIS AND OVERALL STRATEGY CONSIDERATIONS T. J. Burns and I.Spiewak Oak Ridge National Laboratory

One o f the p r i n c i p a l concerns about c i v i l i a n nuclear power centers on the possible diversion o f recycled f i s s i l e material t o weapons fabrication, i n p a r t i c u l a r , the diversion o f plutonium. Depending on the degree t o which t h i s concern i s addressed, various nuclear power strategies can be developed between the current no-reprocessing option (and hence no recycle) and options t h a t would permit the unconstrained recycle o f plutonium. The denatured 233U f u e l cycle t h a t i s the subject o f t h i s r e p o r t provides a middle ground w i t h i n which nuclear power strategies may be developed. Although the denatured cycle does employ recycled f i s s i l e material, i t can be structured so t h a t i t has more p r o l i f e r a t i o n r e s i s t a n t c h a r a c t e r i s t i c s than the plutonium cycle. Before any proposed new f u e l cycle can be implemented, however, i t must be addressed i n the l i g h t o f p r a c t i c a l considerations such as the supply o f U308 available, the projected nuclear power demand, the reactors and f u e l cycles available, and the technological and implementation constraints imposed on the nuclear power system. These various aspects o f nuclear power systems u t i l i z i n g denatured 233U f u e l have been discussed a t length throughout t h i s report. I t i s the purpose o f t h i s f i n a l section o f the r e p o r t t o r e s t a t e the most important conclusions o f the study and t o address trade-offs inherent i n developing nuclear p o l i c y s t r a t e g i e s t h a t include t h e denatured 2 3 3 U f u e l cycle as opposed io s t r a t e g i e s t h a t do not. The nuclear power systems t h a t have been examined can be c l a s s i f i e d as (a) norecycle options, (b) c l a s s i c a l reference recycle options, and (c) denatured recycle options. An integrated assessment o f options i n these three categories i s presented i n matrix form i n Table 7.5-1, which also serves as a basis f o r the discussion t h a t follows. I n evaluating the systems, each option was characterized on the basis o f the f o l l o w i n g criteria: (1) Nuclear p r o l i f e r a t i o n resistance r e l a t i v e t o other nuclear power systems. I

(2)

I] ij

il

6

c

Potential f o r comnercial i z a t i o n o f the reactor/fuel cycle components.

(3) Technical f e a s i b i l i t y on a reasonable schedule (and cost) f o r research, development and demonstration o f the reactor/fuel cycle components. (4)

Capability o f the system f o r meeting long-term nuclear energy demands.

(5)

Economic f e a s i b i l i t y .

As has been pointed out i n e a r l i e r sections o f t h i s report, throughout t h i s study the United States has been used as a base case since the available i n p u t data ( t h a t is, reactor design data, nuclear growth projections, etc. ) required f o r the a n a l y t i c a l model are a l l o f U.S. o r i g i n . However, w i t h appropriate data bases, the same model could apply t o other i n d i v i d u a l nations. Moreover, i t could apply t o cooperating nations, i n which case the nuclear strategy would include a mutual nuclear interdependence o f the p a r t i c i p a t i n g nations.

L f

b:

I L


7-43 \

7.5.1. e-.

!

No-Recycle Options

.

hr

Since commercial-scale reprocessing i s n o t envisioned f o r some time, t h e c u r r e n t l y employed once-through low-enriched uranium c y c l e (LEU) represents the o n l y s i g n i f i c a n t commercial p o s s i b i l i t y i n t h e near term. A t c u r r e n t ore and separative work prices, power generated v i a t h e once-through LEU c y c l e i n LWRs i s economically competitive w i t h The once-through f u e l c y c l e a l s o has favorable p r o l i f e r a t i o n -

o t h e r energy sources.

L3

z

ib: i-

b

L

c

resistant characteristics:

i t s f r e s h f u e l contains an i n h e r e n t i s o t o p i c b a r r i e r ; and

w h i l e i t s spent fuel contains plutonium, t h e fuel i s contaminated w i t h h i g h l y r a d i o a c t i v e f i s s i o n products and thus has a r a d i a t i o n b a r r i e r . advantages (see Case

A

i n Table 7.5-l),

On t h e basis o f these and o t h e r t h e continued near-term use o f the once-through

LEU f u e l c y c l e f o r nuclear-based e l e c t r i c a l generation i s desirable. The p r i n c i p a l drawback o f t h e once-through f u e l c y c l e l i e s i n t h e f a c t t h a t i t i s t i e d t o resources t h a t w i l l become i n c r e a s i n g l y more expensive.

S a t i s f y i n g t h e nuclear demand postulated i n t h i s study t o year 2050 would r e q u i r e t h e consumption o f 5.6 t o 7.1 m i l l i o n tons u@8.

An e q u a l l y important consideration i s t h a t i t would a l s o r e q u i r e an

annual U & production capacity o f 90,000 t o 130,000 tons o f U$I8 by t h e year 2030. As t h e p r i c e o f uranium increases, t h e r e w i l l be incentives t o reduce both these requirements by using uranium more e f f i c i e n t l y . For example, improved LWR technology could p o t e n t i a l l y reduce U308 consumption l e v e l s up t o about 25% i n t h e year 2030. A reduction i n enrichment t a i l s assay c o u l d r e s u l t i n an a d d i t i o n a l reduction i n the uranium requirements o f about 16%; however, t h i s would r e q u i r e about 80% a d d i t i o n a l SWU c a p a c i t y t o maintain a constant production l e v e l of enriched uranium.

But even w i t h these gains t h e v i a b i l i t y o f t h e once-

through c y c l e would be l i m i t e d by the a v a i l a b i l i t y and p r o d u c i b i l i t y o f U308 from u n c e r t a i n resources i n the next century.

A second once-through o p t i o n (Case B i n Table 7.5-1)

i li

would i n v o l v e the a d d i t i o n o f

advanced converters t o t h e power system e i t h e r on t h e LEU c y c l e o r on t h e MEU(235)/Th cycle. The implementation o f t h e MEU(235)/Th once-through c y c l e i n LWRs i s uneconomic r e l a t i v e t o the LEU c y c l e p r i m a r i l y because i t would r e q u i r e higher f i s s i l e loadings and hence higher U308 commitments. And even i f i n c e n t i v e s were provided, t h e use o f thorium-based f u e l s i n LWRs would necessitate a d d i t i o n a l f u e l R,D&D. To use e i t h e r t h e LEU c y c l e o r t h e MEU/Th c y c l e i n o t h e r r e a c t o r types would e n t a i l s i g n i f i c a n t expenditures t o commercialize t h e r e a c t o r s i n t h e U.S. Moreover, t h e generic drawback o f once-through cycles - t h a t i s , t h e u n c e r t a i n t y i n t h e s i z e o f t h e economically recoverable resource base woula remain. On t h e o t h e r hand, as costs f o r e x t r a c t i n g the resource base increase ( t o above $100/lb u & y

-

f o r example), comnercialization o f t h e a l t e r n a t e r e a c t o r s w i l l become more a t t r a c t i v e .

i]:

I f continued r e l i a n c e on once-through f u e l cycles i s t o be a long-term p o l i c y , i t would be d e s i r a b l e t o make p r o v i s i o n s f o r r e s t r i c t i n g t h e spread o f enrichment f a c i l i t i e s . Also, safeguarding the spent f u e l elements i s necessary since the plutonium bred i n the

II'

spent fuel represents a p o t e n t i a l source of weapons-usable m a t e r i a l which becomes increasi n g l y accessible as i t s r a d i o a c t i v i t y decays w i t h time.

Near-term r e s o l u t i o n o f the storage


7-44

question could be accomplished v i a i n t e r n a t i o n a l f a c i l i t i e s chartered f o r j u s t such a purSuch centers (and the i n s t i t u t i o n a l arrangements attendant t o them) could a l s o serve

pose.

as forerunners o f the f u l l - s c a l e f u e l c y c l e service/energy c e n t e r concept considered f o r the r e c y c l e-based options. I

7.5.2.

Recycle Options

The i n h e r e n t l i m i t a t i o n s o f t h e resource base would r e q u i r e t h e use o f recycled m a t e r i a l t o supplement t h e LEU c y c l e i f t h e growth o f a nuclear-based e l e c t r i c a l generation capacity were t o be sustained. Table 7.5-1 compares t h r e e r e c y c l e options u t i l i z i n g denatured f u e l (Cases E-G) w i t h two reference r e c y c l e options u t i l i z i n g t h e c l a s s i c a l Pu/U c y c l e (Cases C and D). The two reference cycles d i f f e r i n t h a t Case D employs FBRs w h i l e Case C does not.

The denatured cases d i f f e r i n t h a t Cases E and F a r e a l l - t h e r m a l systems

and Case G employs FBRs i n a d d i t i o n t o thermal reactors. Case E uses o n l y LWRs as d i s persed r e a c t o r s w h i l e Case F uses both LWRs and advanced converters (HWRs, HTGRs, o r SSCRs). I t has been assumed t h a t , given a strong government mandate and f i n a n c i a l support,

a l l t h e f u e l cycles and r e a c t o r types t h a t have been considered i n t h i s r e p o r t could be developed by t h e time they would be needed - by t h e year 2000 o r l a t e r . However, t h e Pu/U c y c l e i s much c l o s e r t o being commercialized than the Th-based cycles, and, as noted i n Chapter 5, t h e research, development, and demonstration costs f o r implementing t h e denatured 233U f u e l c y c l e i n LWRs would be between $0.5 and $2 b i l l i o n higher than t h e costs f o r implementing t h e reference Pu/U c y c l e i n LWRs. I f t h e HWR o r HTGR were t h e r e a c t o r o f choice, an a d d i t i o n a l $2 b i l l i o n would be r e q u i r e d f o r r e a c t o r research, development, and demonstration.

A system i n which r e a c t o r s consuming Pu and producing 233U (transmuters) are combined with reactors operating on denatured 2 3 3 U f u e l appears t o have somewhat b e t t e r p r o l i f e r a t i o n - r e s i s t a n t c h a r a c t e r i s t i c s than a system based s o l e l y on t h e Pu/U cycle. The "fresh" 233U f u e l i s denatured w i t h 238U, and thus some o f t h e p r o l i f e r a t i o n - r e s i s t a n t features o f t h e f r o n t end o f t h e LEU c y c l e would be extended t o t h e r e c y c l e mode. That i s , both chemical and i s o t o p i c separation o f t h e f r e s h f u e l would be necessary t o o b t a i n weapons-usable m a t e r i a l .

A d d i t i o n a l l y , t h e f r e s h denatured f u e l i s contaminated w i t h

r a d i o a c t i v i t y due t o t h e decay daughters of a 232U i m p u r i t y t h a t i s unavoidably produced 33U, and t h e associated complications introduced i n t o t h e isotope separation

along w i t h t h e

procedure would be severe.

By contrast, weapons m a t e r i a l could be obtained from Pu/U o r

23%/Th f u e l through chemical separation alone, although t h e 2 3 % obtained would a l s o be (The Pu/U f u e l would a l s o be r a d i o a c t i v e b u t much

r a d i o a c t i v e due t o t h e 232U daughters. l e s s so.)

The spent denatured f u e l represents a somewhat lower p r o l i f e r a t i o n r i s k than t h e spent f u e l from o t h e r options would. 238U

The recovery o f a given q u a n t i t y o f Pu bred i n t h e

denaturant would r e q u i r e t h e processing o f more m a t e r i a l than would be necessary i n


j

7-45

I

Table 7.5-1.

Integrated Assessment o f Various Nuklear Policy Options for Meeting U.S. Nuclear Power Gr I

ReactorlFuel Cycle Cmbi nati on

Pro1 iferation Resistance

R,D&D Cost and Time of Comnercial Introduction

Imp1ementati on/Comnerci a1 i zation

Ability to Meet P

Economics

No-Recycl e Options

LWRs on LEU cycle

0

0 0

LEU-LWRs fol lewd by advanced converters on LEU (SEU) cycle o r on MEU(235)/Th cycle

Probably best to the extent that non-nuclear weapons states continue t o forego national fuel recycle Fresh fuel has isotopic barrier; spent fuel contains radioactive fission products Spent fuel stockpile containing Pu is a risk; requires institutional barriers Similar t o above HTGRs on MEU/Th cycle would reduce Pu production by factor of 5 over LEU-LWRs b u t fresh fuel would have higher 235U content (20%)

HWRs on SEU cycle about equal to LWRs on LEU cycle i n Pu production

In wide comnercial use Concern exists about fuel supply Emphasis on improved LWRs and U308 resource development needed

0

0

L i t t l e comnercial incentive to introduce advanced converter Known t o be technically feasible Concern exists about long-term fuel supply

Low cost Gradual improvements introduced from year 1980 to year 2000

0 0

Economics closely linked to U308 price Very favorable a t current U308 prices

Uncertain capital costs cloud near-term interest Advanced converters favored a t high U308 prices .( >$lOO/l b)

Up to $2 billion for advanced converter R,D&D Advanced converters introduced i n 1990's

C1assi cal Reference Recycle Options Once-through LEU-LWRs followed by LWRs w i t h Pu recycle 0 0

Once-thwugh LEU-LWRs followed by LWRs and FBRs w i t h Pu recycle

Recycled Pu i n fresh fuel chemically separable; probably acceptable i f Pu can be limited t o nuclear weapons states and to secure international fuel service centers Option requires technical and institutional barriers for Pu-fueled reactors (*30%) Spent fuel contains radioactive fission products Increased rfsk over Case C because system tends to become Pu dominated Leads to significant Pu invertories and requires extensive Pu transportation for dispersed reactors Requires technical and institutional barriers

0

0

Preferred (>$100/1b) over Case A a t high U308

R&D 0

Introduction i n l a t e 1980's

0

FBR R,RltD UP to $10 billion Fuel cycle R,D&D $1.6 to $3 billion FBRs not available before 2000

'

Preferred by private sector

0

About $1 billion, mainly for fuel cycle

Acceptable t o private sector Requires completion of Generic Environmental Impact Statement on Mixed Oxide Fuel

FBR licensing and comnercial- 1

ization may be d i f f i c u l t Uncertain public acceptance

0

Superior a b i l i t y t o growth greater than this study Divorce from mining

Economics uncertain because of FBR costs, b u t probably acceptable

Up t o $0.5 billion, PWRs and BWRs Fuel cycle R,D&D $1.8 to $3.3 billion Introduction i n 1990's

Somewhat better than superiority of 233U fuel

Close to Case C

Up to $2.5 billion for advanced converters Fuel cycle same as i n Case E Introduction i n l a t e 1990's

Can fully satisfy as med demand through year 2050 for plenti$l U308 supply; especially true i f HWR converters used

Possibly lowest cost for U308 price range of $100-$200/lb, especially f o r HTGR converter

Up to $10 billion for FBRs Converter R,D&D as i n Cases E and F Fuel cycle $2 to $3.6 billion Introduction a f t e r year 2000

As good as Case D ab power demand Divorce from U m i n i n for Case D above

Denatured Recycle Options Dispersed LWRs operating on LEU and denatured 233U fuel w i t h U recycle; energycenter thermal transmuters (LWRs) w i t h Pu recycle -

* 0

0

Dispersed LWRs and advanced converters operating on LEU and denatured 233U fuel w i t h U recycle; energy-center thermal transmuters (LWRs and advanced converters) w i t h Pu recycle Dlspersed LWRs and advanced converters operating on LEU and denatured 233U fuel U recycle; energy-center f a s t transmuters w i t h Pu recycle

0 0

0

0

"Fresh" denatured fuel has isotopic and radioactive barriers; spent fuel contains radioactive fission products Spent denatured fuel contains less Pu than spent LEU fuel (factor of 2.5 less) Requi res technical and institutional barriers to limit Pu to secure energy centers Reduces Pu-fueled reactors by factor of 2 compared with Case C Fresh and spent denatured fuel advantages same as for Case E Requires technical and institutional barriers Use bf HWRs o r HTGRs substantially reduces Pu production relative to Cases C and E Pu produced i n denatured HWRs and HTGRs may be discarded w i t h minor loss of fuel efficiency Very similar t o Case E except that 15 to 50% of reactors may be Pu-fueled FBRs, depending on choice of cycles

0

0

0

Fuel cycle somewhat more complex than Pu/U cycle, b u t func: ti onall y equivalent Requires government incentive

0

0 0

Same as Case E Advanced converters 1i kely t o t o be attractive i f FBRs are unavai lab1 e

0

Same as Case E Private sector 1i kely to accept government mandate Should be structured for maximum thermal -to-fast reactor r a t i o to allow siting f l e x i b i l i t y

0

0 0

0 0

0

0

0 0

I

l

Economics similar to Case D above If FER costs are high, can compensate by reducing the fraction of FBRs i n the mix and increasing the mining rate


7-46

e i t h e r t h e Pu/U c y c l e o r t h e LEU c y c l e (about 2.5 times more than t h e LEU cycle). I t must be noted, however, t h a t t h e presence o f chemically separable f i s s i l e m a t e r i a l a t any p o i n t i n a fuel c y c l e represents a p r o l i f e r a t i o n r i s k , and thus these p o i n t s must be s u b j e c t t o s t r i n g e n t safeguards.

Also, t h e p o t e n t i a l spread o f enrichment f a c i l i t i e s and improve-

ments i n enrichment technology (and hence g r e a t e r ease i n o b t a i n i n g f i s s i l e m a t e r i a l ) may make such d i f f e r e n c e s between t h e various f u e l cycles l e s s important. As i s evident from Table 7.5-1,

t h e p r i v a t e sector p r e f e r s t h e Pu/U c y c l e t o t h e

denatured f u e l cycle, and a government mandate would probably be r e q u i r e d t o induce commercialization of denatured r e c y c l e i n preference t o Pu/U recycle.

Private investors have developed r e c y c l e technology f o r mixed-oxide Pu f u e l s extensively, w h i l e p u t t i n g l i t t l e e f f o r t i n t o r e c y c l e technology f o r thorium-based f u e l s . Because reprocessing i s i n h e r e n t i n t h e denatured 233U cycle, implementation o f t h e c y c l e i s l i k e l y t o r e q u i r e the development o f " f u e l s e r v i c e centers,"

safeguarded f a c i l i t i e s

whose purpose would be t o p r o t e c t s e n s i t i v e fuel c y c l e a c t i v i t i e s .

Such centers could evolve from t h e safeguarded spent f u e l storage f a c i l i t i e s r e q u i r e d f o r t h e once-through f u e l cycles.

For t h e r e c y c l e scenarios, t h e center would f i r s t c o n t a i n s e n s i t i v e f u e l c y c l e f a c i l i t i e s t o produce denatured 2 3 3 U f u e l s from stored 233U-containing spent f u e l ; l a t e r i t would include those r e a c t o r s t h a t operate on f u e l from which t h e f i s s i l e component could be chemically separated. Under the assumption t h a t no weapons-usable f u e l t h a t i s chemically separable can be used i n dispersed reactors, a power system u t i l i z i n g denatured Z 3 3 U f u e l has a s i g n i f i c a n t advantage over one based on t h e Pu/U c y c l e alone. The Pu/U c y c l e would necessitate t h a t a l l r e a c t o r s be constrained t o the energy center, which w i l l r e s u l t i n a penalty f o r e l e c t r i c power transmission since energy centers could n o t be s i t e d as conveniently as d i s p e r i e d reactors. With a denatured system, a s i g n i f i c a n t f r a c t i o n (up t o 85%) o f t h e power could be dispersed since o n l y the Pu-fueled transmuters would be operated i n such centers and thus t h e system could maintain a r e l a t i v e l y h i g h energy-support r a t i o ( r a t i o o f nuclear capacity i n s t a l l e d outside c e n t e r t o nuclear c a p a c i t y i n s t a l l e d i n s i d e center). Evaluation of t h e denatured 2 3 % fuel c y c l e on t h e basis o f economics and/or energy supply i s d i f f i c u l t due t o t h e u n c e r t a i n t i e s i n u n i t c o s t f a c t o r s and p o t e n t i a l energy demand.

With the economic and energy demand assumptions employed i n t h e a n a l y s i s pre-

sented i n Chapter 6, however, t h e economics o f t h e denatured c y c l e appear t o be e q u i v a l e n t to, o r s l i g h t l y b e t t e r than, t h e economics o f t h e c l a s s i c a l Pu/U c y c l e f o r moderate growth-rate scenarios ( i .e.

, those

employing combinations o f f a s t and thermal systems).

Although the fuel c y c l e u n i t costs o f t h e denatured c y c l e were assumed t o be higher than those of t h e Pu/U cycle, power systems u t i l i z i n g denatured 233U f u e l t y p i c a l l y a l l o w a l a r g e r f r a c t i o n o f t h e r e a c t o r s constructed t o be thermal reactors, which have lower c a p i t a l costs. This i s p o s s i b l e because t h e nuclear p r o p e r t i e s o f * 3 3 U are such t h a t i t can be used i n thermal reactors more e f f i c i e n t l y than i n f a s t reactors.


7-47

Although the s t r a t e g y analyses presented i n Chapter 6 considered various advanced

r-

isi.

c

converters as p o t e n t i a l dispersed denatured reactors, t h e s e l e c t i o n o f an optimum advanced converter i s precluded a t t h i s time due t o c o s t and performance u n c e r t a i n t i e s and the f a i l u r e o f t h i s study t o i d e n t i f y a s i n g l e advanced converter f o r f u r t h e r development on t h e basis o f comnonly accepted s e l e c t i o n c r i t e r i a . For example, a t high U308 prices, t h e HTGR appears t o generate t h e lowest-cost power of t h e thermal reactors, w h i l e an HWR appears t o be t h e most r e s o u r c e - e f f i c i e n t and t o have t h e best energy-support r a t i o on t h e denatured cycle. The SSCR might be developed most q u i c k l y and cheaply. A l l t h e advanced converters, b u t p a r t i c u l a r l y the HWR and the HTGR, appear t o have c e r t a i n superior fuel u t i l i z a t i o n c h a r a c t e r i s t i c s r e l a t i v e t o standard LWRs due t o t h e i r higher conversion r a t i o s (i .e., lower Pu production.

â&#x20AC;&#x2DC;Id

lower

233U

makeup requirements), lower f i s s i l e inventories, and

Denatured advanced converters a l s o can be sustained a t higher support

r a t i o s than can denatured LWRs.

[Cycles w i t h p o t e n t i a l l y higher thermal e f f i c i e n c i e s (such

as t h e d i r e c t c y c l e ) and p o t e n t i a l s i t i n g advantages were n o t considered i n t h e comparisons o f t h e advanced converters. ] The i n t r o d u c t i o n o f denatured advanced converters, however, i s estimated t o r e q u i r e up t o $2 b i l l i o n more research, development, and demonstration expenditures than would t h e i n t r o d u c t i o n o f a denatured LWR.

Moreover, a denatured LWR could be commercialized

up t o 10 years sooner than a denatured advanced converter.

Developing a denatured LWR

would be l e s s d i f f i c u l t due t o t h e backlog o f LWR experience and t h e reduced r i s k associated w i t h a p r e v i o u s l y demonstrated r e a c t o r system. The c a p i t a l c o s t o f an advanced converter, although g e n e r a l l y lower than t h e c o s t o f a f a s t reactor, i s estimated t o be somewhat higher than t h a t o f an LWR.

L i

Thus, t h e improved performance must be weighed

a g a i n s t t h e increased c a p i t a l costs, the delay i n i n t r o d u c t i o n , and the research and development costs i n any d e c i s i o n r e l a t i v e t o t h e use o f advanced converters i n conj u n c t i o n w i t h t h e denatured cycle. The a n a l y s i s o f Chapter 6 i n d i c a t e s t h a t , as 23%J producers, f a s t transmuters would have more favorable resource c h a r a c t e r i s t i c s than thermal transmuters. For t h e energy demand assumed i n t h i s study, t h e most s a t i s f a c t o r y denatured power system would c o n s i s t o f denatured thermal r e a c t o r s coupled t o f a s t transmuters i n a symbiotic r e l a t i o n s h i p , t h e l o g i c a l transmuter candidate being a f a s t r e a c t o r w i t h (Pu-U)02 d r i v e r s and Tho2 blankets. I t should be noted, however, t h a t a more r a p i d growth i n energy demand could d i c t a t e t h a t

Pu/U breeders be constructed t o meet t h e demand o r t h a t some combination o f Pu c y c l e breeders containing thorium and dispersed denatured breeders be used.

I n these cases t h e

nuclear power capacity could grow independent of t h e resource base.

c

Although t h e denatured c y c l e appears t o possess advantages r e l a t i v e t o the Pu/U cycle, several important areas r e q u i r e f u r t h e r study.

I n p a r t i c u l a r , t h e refinement o f t h e denatured advanced converter c h a r a c t e r i z a t i o n i s o f prime importance, both t o evaluate

various r e a c t o r options and t o study t h e o v e r a l l use o f advanced converters as opposed t o

LWRs.

As t h e p o t e n t i a l f o r improving t h e performance o f LWRs, both on t h e once-through


7-48

and r e c y c l e modes, i s b e t t e r defined and as advanced converter designs a r e optimized f o r denatured systems, the analysis w i l l become more u s e f u l f o r R,D&D planning.

Also, system

i n t e r a c t i o n studies f o r t h e dispersed denatured reactors and c e n t r a l i z e d transmuters r e q u i r e refinement based on improved r e a c t o r designs and updated mass balances. F i n a l l y , t h e question o f implementing t h e energy-center concept, together w i t h t h e use o f s p e c i a l l y designed transmuters as a source o f denatured f u e l , deserves more d e t a i l e d study.

The

L

Monprol i f e r a t i o n A l t e r n a t i v e Systems Assessment Program (NASAP) i s c u r r e n t l y developing c h a r a c t e r i z a t i o n s o f improved f a s t transmuters, improved LWRs, and reoptimized advanced converters and LMFBRs. L i g h t Water Breeder Reactors (LWBRs) w i l l a l s o be included i n these c h a r a c t e r i z a t i o n studies.

7.5.3.

Overall Conclusions and Recommendations

I

The denatured 233U c y c l e emerges from t h i s assessment as a p o t e n t i a l a l t e r n a t i v e t o t h e conventional Pu/U cycle. 0

I t s advantages may be characterized as f o l l o w s :

L L

The denatured 233U c y c l e o f f e r s p r o l i f e r a t i o n - r e s i s t a n c e advantages r e l a t i v e t o the Pu/U c y c l e i n t h a t the " f r e s h " denatured f u e l has an i s o t o p i c b a r r i e r ; t h a t i s , i t does n o t contain chemically separable Pu o r h i g h l y enriched uranium. By contrast, t h e Pu/U c y c l e together w i t h f a s t breeder reactors tends toward

an e q u i l i b r i u m w i t h a l l r e a c t o r s using Pu f u e l s .

Also, f r e s h denatured f u e l

has a much more intense r a d i o a c t i v e b a r r i e r than does t h e f r e s h f u e l o f t h e c l a s s i c a l Pu/U cycle. For moderate growth r a t e scenarios, deployment o f power systems t h a t i n c l u d e r e a c t o r s operating on denatured 233U f u e l would a l l o w a l a r g e r f r a c t i o n o f t h e reactors i n a power system t o be thermal reactors.

This would tend t o

minimize t h e o v e r a l l c a p i t a l costs o f t h e system compared t o fast/thermal power systems based on t h e Pu/U cycle. 0

I f i n a d d i t i o n t o LWRs, t h e denatured thermal r e a c t o r s o f t h e power system were t o i n c l u d e denatured advanced converters, t h e dependence o f t h e power system on a f a s t r e a c t o r component (i.e.,

f a s t transmuters) could be f u r t h e r

minimized due t o the improved resource u t i l i z a t i o n o f denatured advanced converters compared t o denatured LWRs. A1 though the advanced converters would have higher c a p i t a l costs than the LWRs, t h i s might be o f f s e t by reduced requirements f o r FBRs. The disadvantages o f t h e c y c l e are t h e f o l l o w i n g : 0

The denatured 2 3 % f u e l c y c l e i s more complex than t h e Pu/U cycle, and since

z3%

must be produced i n transmuter reactors, t h e r a t e a t which denatured 233U

r e a c t o r s can be introduced w i l l be i n h e r e n t l y l i m i t e d .

Because t h e Pu/U c y c l e

L L L


7-49

technology i s c l o s e r t o commercialization, U.S.

r e

t h e r e i s a reluctance both by

i n d u s t r y and by f o r e i g n governments t o embrace an a l t e r n a t i v e which

i s l e s s developed and which i s considered p r i m a r i l y on t h e basis o f i t s n o n p r o l i f e r a t i o n advantages, and t h i s would have t o be overcome. The R,D&D costs f o r developing t h e denatured 233U f u e l c y c l e are s i g n i f i c a n t l y higher than those f o r t h e Pu/U cycle.

Ifadvanced converters must a l s o be

developed, s i g n i f i c a n t a d d i t i o n a l costs would be incurred. Other important conclusions from t h i s study are as follows: e

The once-through c y c l e based on LWRs i s l i k e l y t o dominate nuclear power production through t h e year 2000. This provides time t o develop e i t h e r t h e denatured c y c l e o r t h e Pu/U c y c l e f o r the r e c y c l e mode.

e

The denatured 233U f u e l c y c l e can be used i n LWRs, SSCRs, HWRs, HTGRs, and FBRs w i t h o u t major changes from t h e present conceptual r e a c t o r designs based on t h e i r reference f u e l s .

e

A f t e r t h e necessary R,D&D i s completed, the denatured 2 3 % f u e l c y c l e appears t o be economically competitive w i t h the Pu/U f u e l c y c l e i n LWRs, advanced converters, and i n symbiotic fast-thermal r e c y c l e systems. With the f u e l resources assumed, the nuclear power demand postulated i n t h i s study (350 GWe i n the year 2000 and a n e t increase o f 15 GWe/yr t h e r e a f t e r ) can be met as w e l l by the denatured f u e l c y c l e as i t can by the Pu/U cycle. However, the Pu/U-FBR c y c l e has an inherent a b i l i t y t o grow a t a f a s t e r r a t e than the other cycles.

On the basis o f t h i s study, i t i s recommended t h a t : e

Optimized designs o f a1 t e r n a t e breeders , improved LWRs , HWRs , SSCRs , and HTGRs be examined t o r e f i n e t h e c h a r a c t e r i s t i c s o f t h e denatured c y c l e r e l a t i v e t o f u e l u t i l i z a t i o n , economics and energy-support r a t i o .

The

study should a l s o be expanded t o i n c l u d e LWBRs and the f a s t breeder designs developed by DOE i n t h e P r o l i f e r a t i o n Resistant Large Core Design Study (PRLCDS).

More d e t a i l e d assessments o f t h e p r o l i f e r a t i o n

r i s k s and the economics o f the denatured cycles compared t o other r e c y c l e options (Pu/U and HEU/Th) should a l s o be pursued.


7-50

These studies could provide guidance f o r the f o l l o w i n g R&D programs:

0

0

0

0

Thorium f u e l cycle R&D t o investigate the use o f MEU(235)/Th, MEU( 233)/Th (denatured 233U), and Pu/Th f u e l s i n LWRs and HWRs (the l a t t e r i n cooperation k i t h Canada). This program might also include the LWBR f u e l cycle. Studies t o consider denatured HTGR reference f u e l cycle.

233U

o r 235U f u e l s as candidates f o r the

Thorium technology studies, p a r t i c u l a r l y f o r blanket assemblies, as an i n t e g r a l p a r t o f the FBR programs (LMFBRs and GCFBRs).

L

Exploratory work w i t h u t i l i t i e s and PWR and BWR vendors f o r q u a l i f i c a t i o n and use o f MEU/Th and Th f u e l rods i n commercial reactors. An example o f the b e n e f i c i a l use o f Th would be i n corner rods o f the BWR f u e l assembly.

c


t

a

APPENDICES


E

c


A-3

Appendix A.

ISOTOPE SEPARATION TECHNOLOGIES

E. H. G i f t Oak Ridge Gaseous D i f f u s i o n P l a n t A.l.

Current Separation C a p a b i l i t y

Three enrichment technologies e x i s t t h a t are s u f f i c i e n t l y advanced t o be c l a s s i f i e d as c u r r e n t separation technology. a. b. c.

These are:

The Gaseous D i f f u s i o n process. The Gas Centrifuge process. The Becker Separation Nozzle process (and i t s variant, t h e South A f r i c a n Helikon process). Both t h e c e n t r i f u g e and t h e Becker processes are expected t o provide enrichment

services t h a t are competitive w i t h gaseous d i f f u s i o n .

The c e n t r i f u g e process, i n p a r t i -

c u l a r , i s p r o j e c t e d t o provide a 30%l saving i n separative work c o s t when f u l l y implemented i n a l a r g e scale p l a n t .

A b r i e f d e s c r i p t i o n o f each o f these processes and t h e i r c u r r e n t productive capacity follows. The Gaseous D i f f u s i o n Process2 The gaseous d i f f u s i o n process i s based upon t h e physical f a c t t h a t i n a gas made up o f molecules o f d i f f e r e n t masses, molecules containing the l i g h t e r mass isotopes w i l l , as a r e s u l t o f t h e d i s t r i b u t i o n o f k i n e t i c energies, have average v e l o c i t i e s s l i g h t l y As a r e s u l t , these l i g h t e r isotopes

f a s t e r than those which contain the heavier isotopes.

w i l l reach t h e w a l l s o r pores i n t h e w a l l s o f a containment vessel more f r e q u e n t l y and a t higher v e l o c i t i e s .

I

I n the gaseous d i f f u s i o n process, the container w a l l i s a porous tube

( b a r r i e r ) through which d i f f u s i o n i s accomplished. The maximum t h e o r e t i c a l separation t h a t can be achieved i s a f u n c t i o n o f t h e square r o o t of t h e r a t i o o f t h e masses o f t h e gas molecules.

I n t h e d i f f u s i o n process,

u t i l i z i n g uranium hexafluoride, the square r o o t o f t h e r a t i o i s 1.00429.

Because t h i s

number i s so c l o s e t o u n i t y , the degree o f enrichment which can be achieved i n a s i n g l e d i f f u s i o n stage i s very small, b u t t h e e f f e c t can be m u l t i p l i e d by making use o f a cascade c o n s i s t i n g o f a number o f stages. Production o f 90 weight percent 235U from 0.711 weight percent 235U material, as found i n n a t u r a l ore, requires about 3,000 A p l a n t constructed f o r t h e purpose o f producing m a t e r i a l

d i f f u s i o n stages i n series.

o f up t o 4.0 weight percent 235U, as might be r e q u i r e d f o r t y p i c a l l i g h t water power reactors, would c o n t a i n about 1200 stages.


A-4

To take advantage o f the small separation f a c t o r discussed above, d i f f u s i v e must be ensured, not j u s t simple gas flow. D i f f u s i v e flow requires not only small i.e., less than two-millionths o f an i n c h i n diameter, but also u n i f o r m i t y o f pore Because o f the small pore size, l i t e r a l l y acres o f b a r r i e r surface are required i n production plant.

flow pores, size. a large

Complexity o f p l a n t design i s increased by the d i f f i c u l t i e s a r i s i n g from the nature o f the d i f f u s i n g gas i t s e l f . A v o l a t i l e compound o f uranium must be used, and the hexafluoride (UF6) i s the only known s u i t a b l e compound. It i s a s o l i d a t room temperature; consequently, the d i f f u s i o n plants must be operated a t temperatures and pressures necessary t o maintain the UF6 i n gaseous form. Although i t i s a stable compound, UF6 i s extremely r e a c t i v e w i t h water, very corrosive t o most comnon metals, and n o t compatible w i t h organics such as l u b r i c a t i n g o i l s . This chemical a c t i v i t y d i c t a t e s the use o f metals such as n i c k e l and aluminum and means t h a t the e n t i r e cascade must be l e a k - t i g h t and clean. The corrosiveness o f the process gas also imposes added d i f f i c u l t i e s i n the f a b r i c a t i o n o f a b a r r i e r which must maintain i t s separative q u a l i t y Over long periods o f time. The enrichment stage i s the basic u n i t o f the gaseous d i f f u s i o n process. I n a l l stages gas i s introduced as UF6 and made t o flow along the i n s i d e o f the b a r r i e r tube. I n the standard case about one-half the gas d i f f u s e s through the b a r r i e r and i s fed t o the next higher stage; the remaining undiffused p o r t i o n i s recycled t o the next lower stage. The d i f f u s e d stream i s s l i g h t l y enriched w i t h respect t o 235U, and the stream which has n o t been d i f f u s e d i s depleted t o the same degree.

L

c

'

The basic equipment components v i t a l t o the process are the a x i a l flow compressors, the converter s h e l l and the b a r r i e r tubes. Axial flow compressors are used t o compress the UF6 gas t o maintain the interstage flow, and e l e c t r i c motors are used t o d r i v e the compressors. A gas cooler i s provided i n the converter since gas compression unavoidably generates heat which must be removed a t each stage. The d i f f u s e r , o r converter, i s the l a r g e c y l i n d r i c a l vessel which contains the b a r r i e r material. It i s arranged i n such a fashion t h a t the d i f f u s e d stream and the stream t h a t has not d i f f u s e d are kept separate. Groups o f stages are coupled t o make up operating u n i t s and such groups, i n turn, make up the cascade. Gaseous d i f f u s i o n plants are i n operation i n the United States, England, France, and Russia.

L: L I'

Li


A- 5

The Gas Centrifuge Process The countercurrent gas c e n t r i f u g e separation o f uranium isotopes i s based on processes developed more o r l e s s independently i n t h e U.S. a t t h e U n i v e r s i t y o f Virginia,3 i n germ an^,^ and i n Russia5 d u r i n g World War 11. Much o f t h i s work was reported a t t h e

I]

1958 Geneva Conference. I n t h e U.S. t h i s work was continued a t t h e U n i v e r s i t y o f V i r g i n i a and reported i n 1960.6 The machine developed i s shown i n Fig. A-1. The theory4s7 f o r operation o f t h e gas centrifuge shows t h a t t h e maximum separative capacity o f a gas c e n t r i f u g e i s p r o p o r t i o n a l t o :

1$i

a.

The f o u r t h power o f t h e peripheral speed,

b.

t h e length, and

c.

t h e square o f t h e d i f f e r e n c e i n molecular weights.

Thus, i t i s evident t h a t one should make the peripheral speed and the l e n g t h o f the c e n t r i f u g e as l a r g e as possible. The peripheral speed i s l i m i t e d by t h e b u r s t i n g strength

A long r o t o r o f small diameter i s comparatively f l e x i b l e and w i l l pass through a series o f resonant mechanical v i b r a t i o n frequencies w h i l e being accelerated t o h i g h peripheral speed. Unless provided w i t h special damping bearings, a c e n t r i f u g e would destroy i t s e l f w h i l e passing through one o f these resonant speeds. Much

o f t h e m a t e r i a l o f the r o t o r w a l l .

o f t h e w o r l d ' s e f f o r t i n advanced c e n t r i f u g e development has been designed t o keep below the f i r s t resonant frequency.

As a r e s u l t , they are comparatively s h o r t and have r e l a t i v e l y

low separative capacity.

ill

Some o f the d i f f e r e n c e s between gas c e n t r i f u g e and gaseous d i f f u s i o n technologies Gaseous d i f f u s i o n requires f a b r i c a t i o n o f permeable b a r r i e r s

should perhaps be noted.

w i t h a very small pore size; the manufacture o f these b a r r i e r s i s a d i f f i c u l t process and a c l o s e l y guarded secret. Gas c e n t r i f u g a t i o n requires manufacture o f high-speed r o t a t i n g equipment. While such manufacture i s c e r t a i n l y n o t t r i v i a l , i t b a s i c a l l y

8 LI b'

requires a well-equipped p r e c i s i o n machine shop t h a t may w e l l be w i t h i n t h e t e c h n i c a l c a p a b i l i t i e s o f many nations.

The technology o f r o t a t i n g machinery i s widespread and

designs f o r gas c e n t r i f u g e s are i n t h e open l i t e r a t u r e . The power requirements f o r a c e n t r i f u g e f a c i l i t y are much l e s s than f o r a d i f f u s i o n f a c i l i t y o f t h e same size.

For U.S. p l a n t s o f economic scale and o f t h e same separative capacity, gas c e n t r i f u g a t i o n requires about 7% o f t h e power needed f o r gaseous

diffusion. Following t h e e a r l y work i n t h e

U.S., f u r t h e r research on t h e c e n t r i f u g e process

was undertaken f o r t h e USAEC by t h e U n i v e r s i t y o f V i r g i n i a , Union Carbide Corporation Nuclear D i v i s i o n and G a r r e t t Corporation-AiResearch Manufacturing Co., and D r . Lars Onsager.

The c u r r e n t s t a t u s o f t h e U.S.

program can best be i n d i c a t e d by a b r i e f

d e s c r i p t i o n o f t h e operating and planned f a c i 1 i t i e s : l


A-6

rv L i g h t Fraction Heavy Fraction

-'If

uF6 Feed Magnet

/ Stationary Suspension

To Vacuum Pump

-

Magnet Spinning Magnet Damper

Molecular Pump scoop

L

Rotor

Baffle P1a t e

Casing

c scoop Drive Motor

\ Fig. A-1.

Bearing

ZIPPE Centrifuge (Simplified).

The Equipment Test F a c i l i t y (ETF) was conceived t o provide f o r the r e l i a b i l i t y t e s t i n g o f "high capacity" centrifuges. This f a c i l i t y , which began operation i n 1971, has been the source o f r e l i a b i l i t y t e s t i n g f o r two generations o f machine designs. Many o f the f i r s t generation high capacity machines are s t i l l operating i n t h i s f a c i l i t y .

L L L


A- 7

The Component Preparation Laboratories (CPL) i n Oak Ridge, Tennessee and Torrance, C a l i f o r n i a , were b u i l t t o evaluate, improve and demonstrate techniques amenable t o t h e

id

mass production f o r manufacturing centrifuges.

This f a c i l i t y became operational i n

e a r l y 1974. The Component Test F a c i l i t y (CTF) was designed t o demonstrate t h e machine r e l i a b i l i t y and o p e r a b i l i t y t e s t i n g o f s u b s t a n t i a l numbers o f c e n t r i f u g e s i n a cascade operation.

Construction was begun i n 1972 and t h e f i r s t phase o f s t a r t u p o f t h e f a c i l i t y

was completed i n January 1977 w i t h cascade operation o f about one-half o f t h e machines operating. The remaining machines were operable w i t h i n a few weeks l a t e r . The capacity o f t h e CTF i s s i g n i f i c a n t , about 50,000 SWU/yr, o r about t h e annual e n r i c h i n g requirement f o r a 500 MW power r e a c t o r . The Advanced Equipment Test F a c i l i t y (AETF), i n a d d i t i o n t o being a r e l i a b i l i t y t e s t f a c i l i t y w i l l a l s o t e s t the p l a n t subsystems which support t h e machines. The machines t o be i n s t a l l e d i n t h i s f a c i l i t y w i l l have s i g n i f i c a n t l y greater separative work c a p a b i l i t y than those i n t h e CTF.

The AETF i s expected t o be operable i n the s p r i n g

o f 1978.

I n Europe, t h e URENCO organization, c o n s i s t i n g o f p a r t i c i p a n t s from England, Germany, and Holland, has a program t h a t so f a r has been d i r e c t e d toward machine r e l i a b i l i t y and long l i f e t i m e . URENCO i s c u r r e n t l y producing about 200 MTSWU/yr from p l a n t s a t Almelo, Holland and Capenhurst, England. Expansion o f these f a c i l i t i e s i s planned by 1982.

The URENCO group expects t o have 2000 MTSWU/yr i n operation, 1300

MTSWU/yr a t Almelo, and t h e remaining 700 MTSWU/yr a t Capenhurst. The Becker Separation Nozzle The Becker process,9 being developed i n Germany by D r . E. W. Becker and h i s associates, u t i 1 i z e s t h e pressure g r a d i e n t developed i n a curved expanding supersonic j e t t o achieve separation i n a gas mixture. The separation nozzle stage i s shown schematically i n Fig. A-2. A l i g h t gas, helium o r hydrogen, i s added t o t h e UF6 i n order t o increase the v e l o c i t y o f the j e t .

As t h e expanding j e t traverses t h e curved path, t h e heavier component i s enriched i n t h e v i c i n i t y o f t h e wall. A k n i f e edge d i v i d e s t h e j e t i n t o two fractions--one enriched i n t h e l i g h t component, and t h e o t h e r enriched i n t h e heavy component--which are then pumped o f f separately from t h e stage. A1 though t h e separation obtained per stage i s r e l a t i v e l y h i g h ( ~ 1 . 0 2 5 ) ~many separation nozzle stages are needed t o o b t a i n an appreciable enrichment.

This process avoids t h e problems associated w i t h

t h e fine-pored membrane r e q u i r e d f o r gaseous d i f f u s i o n , and those associated w i t h t h e high-speed r o t a t i n g p a r t s o f t h e gas c e n t r i f u g e . It does s u f f e r , however, from t h e disadvantage o f a r e l a t i v e l y h i g h power requirement, p r i m a r i l y because a g r e a t deal o f l i g h t gas must be recompressed between stages along w i t h t h e UF6 process gas.


A-8

G

P = total pressure; N = mole fraction of UF in the U F d H e mixture. Subscripts 0, M, and K refer to feed gas, fight and heavy fractions, respectively.

Fig. A-2.

Cross Section of the Separation Nozzle System o f the Becker Process.

A small 10-stage p i l o t p l a n t was operated i n 1967 t o prove the technical feasib i l i t y o f the process. Following that, a s i n g l e l a r g e prototype stage s u i t a b l e for use i n a p r a c t i c a l cascade was fabricated.

A prototype separation stage contains 81 separatlâ&#x20AC;&#x2122;ng elements and i s reported t o have a separative capacity o f approximately 2000 kg U SW/yr. A p l a n t producing a product enriched t o 3% 235U and w i t h t a i l s a t 0.26% *35U i s expected t o require about 450 such stages

.

Figure A-3 shows the i n d i v i d u a l separating elements, each containing 10 separation nozzle s l i t s On i t s periphery. The f a b r i c a t i o n o f these u n i t s i s not as simple as one might a t f i r s t expect. I n order t o obtain the desired separation performance a t reasonable pressures, i t i s necessary t o employ very small geometries. The spacing between the k n i f e edge and the curved w a l l i n the prototype separating u n i t should be about 0.0005 of an inch. I n order t o obtain good performance, i t i s necessary t h a t t h i s spacing n o t deviate by more than +lo%over the 6-foot length o f s l i t . The power requirement f o r the Becker process i s c u r r e n t l y estimated t o be about one and one-third times as great as t h a t required f o r gaseous d i f f u s i o n . Dr. Becker believes t h a t f u r t h e r process improvement i s s t i l l possible and t h a t the power w q u i r e ment can be s u b s t a n t i a l l y reduced.

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A-9 DWG. NO. G-69643

light fraction

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Fig. A-3.

Becker Separating Element With Ten S l i t s

The South A f r i c a n Helikon Process The South A f r i c a n l o ( o r UCOR) process i s o f an aerodynamic type whose separating element i s described by t h e developers as a high-performance stationary-walled c e n t r i -

A l l process pressures throughout t h e system w i l l be above atmospheric and, depending on the type o f " c e n t r i f u g e " used, the maximum process pressure w i l l be i n a range o f up t o 6 bar. The UF6 p a r t i a l pressure w i l l , however, be s u f f i c i e n t l y low t o e l i m i n a t e t h e need f o r process heating d u r i n g p l a n t operation, and t h e maximum temperature a t the compressor d e l i v e r y w i l l n o t exceed 75째C. fuge using UF6 i n hydrogen as process f l u i d .

The process i s characterized by a h i g h separation f a c t o r over t h e element, namely from 1.025 t o 1.030,depending on economic considerations. Furthermore, i t has a h i g h degree o f asymmetry w i t h respect t o t h e UF, f l o w i n t h e enriched and depleted streams, which emerge a t d i f f e r e n t pressures. The feed-to-enriched streams pressure r a t i o i s t y p i c a l l y 1.5, whereas t h e feed-to-depleted streams pressure r a t i o i s t y p i c a l l y o n l y 1.12. To deal w i t h t h e small UF6 cut, a new cascade technique was developed--the so-called "helikon" technique, based on t h e p r i n c i p l e t h a t an a x i a l f l o w compressor can simultaneously t r a n s m i t several streams o f d i f f e r e n t i s o t o p i c composition w i t h o u t t h e r e being s i g n i f i c a n t mixing between them.

The UCOR process must, therefore, be regarded as a

combination o f t h e separation element and t h i s technique, which makes i t p o s s i b l e t o achieve t h e desired enrichment w i t h a r e l a t i v e l y small number o f l a r g e separation u n i t s by f u l l y u t i l i z i n g t h e h i g h separation f a c t o r a v a i l a b l e .

A further feature o f the helikon


A-10

technique i s t h a t a module, de'fined as a separation u n i t consisting o f one s e t o f compressors and one s e t o f separation elements, does not as i n the c l a s s i c case, produce only one separation factor o f enrichment i n one pass b u t can produce f o r a constant separative work capacity various degrees of enrichment up t o a maximum o f several times the separation f a c t o r over the element. F u l l scale modules of t h i s type are nearing the prototype stage. Recent design improvements are expected t o r e s u l t i n a nominal capacity o f 80 t o 90 kg SWU/yrll per separation module.

A valuable feature of a p l a n t based on t h i s process i s i t s very low uranium inventory, which r e s u l t s i n a short cascade e q u i l i b r i u m time, o f the order of 16 hours f o r a commercial p l a n t enriching uranium t o 3% 235U.

e

The t h e o r e t i c a l lower l i m i t t o the s p e c i f i c energy consumption o f the separation element can be shown t o be about 0.30 MW.h/kg SW. The minimum f i g u r e observed by the developers w i t h laboratory separating elements i s about 1.80 MW.h/kg SW, based on adiabatic compression and ignoring a l l system i n e f f i c i e n c i e s . This d i f f e r e n c e i s a measure o f the improvement p o t e n t i a l expected by the South Africans.

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Current and Projected Enrichment Capacity Most o f the known i n s t a l l e d enrichment capacity i s based upon gaseous d i f f u s i o n technology. Only small increments o f centrifuge technology are i n operation (1.e. , URENCO, Japan and U.S.), and one p l a n t u t i l i z i n g modified nozzle technology ( t h e South A f r i c a n Helikon p l a n t ) may be operating. I n d i c a t i v e o f the status o f other isotope separation methods, a l l planned additions t o the world enrichment capacity are based on e i t h e r d i f f u s i o n , centrifuge o r nozzle technology.

E. P

The e x i s t i n g worldwide capacity and planned additions t o capacity are shown i n Table A-1 by colcntry and technology type. I n t h e t a b l e t h e groups i d e n t i f i e d as Eurodif and Coredif are multinational organizations b u i l d i n g gaseous d i f f u s i o n plants i n France.

A.2,

New Separation Technologies

I n a d d i t i o n t o the more developed technologies (gaseous d i f f u s i o n , gas c e n t r i fuge, and the Becker nozzle), there are several other separation methods t h a t e i t h e r have been u t i l i z e d i n the past o r are c u r r e n t l y being developed. These technologies

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are l i s t e d i n Table A-2.

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A-1 1

Table A-1.

Year 1977 i--

L

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Nation o r Group U.S.’ UK-France

Diffusion Diffusion

Russiac

Diffusion

China URENCO U.S. S. A f r i c a

O i ff us ion Centrifuge Centrifuge Helikon-Fixed w a l l centrifuge

800 Unknown 200 50 Unknown

Existing Existing, b u t dedicated t o m i l i t a r y use Existing. actual t o t a l capacity unknown Existing, mostly m i l i t a r y Existing Existing Existing p i l o t plant o r i n process o f coming on-line From CIP/CUP olus added power purchase F a c i l i t i e s a t Almelo & Capenhurst now i n construction Currently under construction

15.400 16,400 17.200 17,400 17,450

200

Japan RussiaC

Centrifuge Diffusion

20 200

us.’

Diffusion Diffusion Centrifuge Diffusion

2.200 500 400 2,600

Under construction

us.’

Diffusion Centrifuge Diffusion Centrifuge Diffusion

1,600 400 3.700 30 500

From CIP/CUP Planned Under construction Under construction

us.’

Diffusion Centrifuge Diffusion Diffusion

700 400 2,100 500

From CIP/CUP Planned Under construction

U.S.’ IJRENCO

Eurodif RussiaC Brazi 1

Diffusion Centrifuge Diffusion Diffusion Becker nozzle

300 400 2.400 500 180

Incr. Power Implementing CUP Planned Under construction Planned

37,000 37.500 39.900 40.400 40,580

1983

UREtlCO Coredif

Centrifuge Diffusion

1.300 1,800

Planned Planned

41.880 43,680

1984

us.’

Diffusion Centrifuge Fixed wall centrifuge Diffusion

2,000 1.300 1,600

Incr. Power Implementing CUP Planned Planned

45,680 46.980 48,580

1,800

Planned

50.380

2,000 1.400

Incr. Power Implementing CUP Planned Planned

52,380 53.780 55.380

1,800 6,000

Planned Planned, b u t should be considered conditional

57,180 63,180

550 1.800

Planned Planned

63,730 65.530

1980

URENCO Eurodif Japan RussiaC

LIREMCO Eurodif Russiac

1982

uT-

URENCO S. A f r i c a

Coredi f 1985

US.’ URENCD S. A f r i c a Coredif Japan

Diffusion Centrifuge Fixed w a l l centrifuge Diffusion Centrifuge

1986

US.’ S. A f r i c a URENCO

Centrifuge Fixed w a l l centrifuge Centrifuge

1987

U.S.’ URENCO

Centrifuge Centrifuge

1988

US.’

1989

U.S.’ Coredif

‘Infornation

D

15.400 800-1000

Present Status of Increment

Diffusion

J !

f

(MT SWU)

World’s Cumulative Capacity (MT SWU)

Centrifuge

1981

G;

Caoacit v Increment

US.’

RussiaC URENCO Eurodif

i

Technology type

URENCO

1978

1979

c

Approximate Schedule of World Enrichment Capacitya

3,300

1.600

From CIP/CUP

20.750 20.950 20.970 21,170 23.370 23.870 24,270 26.870 28.470 28,870 32.570 32,600 33.100

36 ;SO0

2.000

Planned

67,530

2.750

2.000

Planned Planned

70,280 72,280

Centrifuge

3,300

Planned

75,580

Centrifuge Diffusion

2,200 5.400

Planned Planned, b u t should be considered conditional

77.780 83,180

from references 12 and 13.

’Not included i n t h i s schedule are possible additions t o the U.S. enrichment capacity by p r i v a t e corporations, such as Exxon Nuclear, G a r r e t t and Centar; these may amunt t o as much as 10.000 MT SWU by 1990. =For Russia, t h i s i s a schedule o f growth i n enrichment sales a v a f l a b i l i t y and not necessarily o f capacity expansion.


A 12

Other Isotope Separation Technologies

Table A-2. ~~

A.

~~

~

Discarded Technologies Thermal D i f f u s i o n Electromagnetic (the Calutron Process)

B.

Developing Technologies Photo-Exci t a t i o n Methods (Laser) Chemical Exchange Methods Aerodynamic Methods (Other Than the Becker Nozzle and the Fixed Wall Centrifuge) P1asma Based Processes

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The discarded technologies l i s t e d i n Table A-2 have been used t o produce enriched uranium.

A large-scale, liquid-phase, thermal-diffusion p l a n t was constructed i n 1945 This p l a n t produced very s l i g h t l y enriched uranium by the Manhattan Project.14 (0.86%). Thermal d i f f u s i o n i s impractical f o r commercial enrichment o f uranium isotopes because o f i t s very high energy requirements. Compared t o gaseous d i f f u s i o n , the energy requirement i s over 200 times greater. The electromagnetic o r Calutron methods were used during the Manhattan P r o j e c t t o produce h i g h l y enriched uranium. 14 The process was discarded s h o r t l y a f t e r the more economical gaseous d i f f u s i o n p l a n t began ,operation. A b r i e f d e s c r i p t i o n o f t h e process follows. The Calutron Process involved the vaporization o f a s a l t feed material, t y p i c a l l y UC14, from an e l e c t r i c a l l y heated charge b o t t l e through s l o t s i n t o an arc chamber where the s a l t was ionized by an e l e c t r o n beam which t r a v e l s along the l i n e s o f f l u x o f t h e magnet. The ionized uranium, as the U+ i o n f o r the most part, passed through another s l o t where i t was accelerated by other s l o t t e d electrodes i n t o the vacuum tank which f i l l e d the pole area o f a l a r g e electromagnet. The ions from the accelerating electrodes diverged several degrees from the s l o t s and a t the 90" p o i n t passed by some b a f f l e s as a r a t h e r t h i c k beam. This beam was brought t o a focus a t the s l o t s o f a receiver system as curved l i n e s by the shimmed magnetic f i e l d . I n the large units, 96-in. beam diameter, there were up t o f o u r o f these beams i n a given tank. The divergent t r a j e c t o r i e s o f the ions from the f o u r sources intersected some few degrees from the accelerating electrodes and separated as d i s t i n c t beams, again a U++, and s i m i l a r distance from the receivers. There were various side beams o f UCl', other ions which h i t the b a f f l e s and the w a l l s o f the tank a t a series o f locations. The uranium content o f these beams condensed as various compounds o f uranium. The product was, f o r the most part, converted t o UC by i n t e r a c t i o n o f the very high voltage uranium ions w i t h the graphite o f the receivers. Since, i n even the most

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A-13

e f f i c i e n t o f t h e u n i t s developed, o n l y about 22% o f t h e feed was c o l l e c t e d as product i n a v a p o r i z a t i o n c y c l e o f t h e feed, t h e r e were l a r g e amounts o f uranium compounds t o be recovered and r e c y c l e d through t h e system.

The chemical operations r e q u i r e d were

complex, b u t t h e amount o f space and t h e number o f workers r e q u i r e d i n the chemical f u n c t i o n were always small compared t o t h e requirements o f the r e s t o f t h e process. The processing o f t h e r e c e i v e r s t o recover t h e product uranium was a small scale b u t very demanding s e r i e s o f chemical procedures, The developing technologies l i s t e d i n Table A-2 o f f e r no c u r r e n t c a p a b i l i t y f w p r o d u c i n g kilogram q u a n t i t i e s o f enriched uranium. I f any o f them approaches commercial f e a s i b i l i t y , they may provide enhanced o p p o r t u n i t i e s f o r a clandestine enrichment operation.

A b r i e f d e s c r i p t i o n o f each o f these processes follows.

Photoexci t a t i o n (Laser) Methods The development o f h i g h i n t e n s i t y narrow-frequency tunable l a s e r s has r a i s e d t h e p o s s i b i l i t y of n e a r l y complete i s o t o p i c separation i n a s i n g l e step. Thus, r e a c t o r grade and perhaps even weapons grade uranium could be produced i n one pass through t h e apparatus.

Such a single-stage process would a l l o w f o r a much more compact

enrichment p l a n t , saving l a n d area, c a p i t a l investment and power consumption.. These hopes have l e d t o a c t i v e research and development programs i n t h e United States, t h e Soviet Union, I s r a e l , France and p o s s i b l y o t h e r countries. I n t h e U.S. d i s t i n c t lines.

t h e development o f l a s e r enrichment i s being pursued along two One l i n e o f development uses atomic uranium vapor as the source

m a t e r i a l f o r t h e l a s e r e x c i t a t i o n whereas the o t h e r l i n e o f development i s pursuing e x c i t a t i o n o f molecular uranium hexafluoride.

Each method has i t s v i r t u e s and

defects . Laser Enrichment w i t h Atoms.15

I n t h e atomic enrichment process most o f t e n

discussed, molten uranium i s heated i n an oven t o about 2500째K. The atomic vapor emerges i n t h e fohn o f a long, t h i n ribbon i n t o a h i g h l y evacuated r e g i o n where i t i s i l l u m i n a t e d by two v i s i b l e o r n e a r - u l t r a v i o l e t lasers. One l a s e r i s tuned t o a t r a n s i t i o n from t h e ground s t a t e of uranium t o an e x c i t e d s t a t e roughly halfway up t h e ladder t o i o n i z a t i o n .

This i s t h e i s o t o p i c a l l y s e l e c t i v e step, and i t i s hoped

t h a t very h i g h s e l e c t i v i t i e s w i l l be achieved here. The purpose o f t h e second l a s e r i s t o boost the e x c i t e d 235U atoms t o a l e v e l j u s t below t h e i o n i z a t i o n l i m i t .

This step need n o t be i s o t o p i c a l l y s e l e c t i v e , and

i n p r i n c i p l e t h e second l a s e r could be used t o i o n i z e t h e atom d i r e c t l y . But i o n i z a t i o n cross sections a r e g e n e r a l l y about 1000 times smaller than resonant e x c i t a t i o n cross sections, and so i t i s f a r more e f f i c i e n t t o use a resonant t r a n s i t i o n t o e x c i t e

L"

t h e atom t o a s t a t e j u s t below t h e i o n i z a t i o n l e v e l and then t o use e i t h e r a s t a t i c


A-14

e l e c t r i c f i e l d or an infrared l a s e r pulse t o p u l l the electrons off the atoms. Once the atoms are ionized, they can be separated from the neutral atoms i n the beam by the use of e l e c t r i c or magnetic f i e l d s , o r both. The major l i m i t i n g factor i n the above process is the density of atoms i n the uranium "ribbon." There is an upper limit on the density and therefore on the r a t e of production of enriched uranium, because both excitation energy and ionic charge are very easily transferred t o other atoms i n collisions. Such collisions must be kept t o a minimum i f a high s e l e c t i v i t y is t o be obtained.

Other technical d i f f i c u l t i e s i n the 'development of the process are: a. The corrosiveness of the uranium vapor. b. The presence of thermally excited o r ionized atoms of 235U i n the uranium vapor ( a t 2500째K, 4 5 % of 2 3 % atoms a r e not i n the ground s t a t e ) . c. The potential f o r self lasing of the uranium vapor. d. Thermal ionization of 238U will seriously degrade the selectivity and t h u s limit the enrichment. e. Lasers combining h i g h energy density, rapid pulse repetition r a t e , h i g h t u n i n g precision, and long-term s t a b i l i t y and r e l i a b i l i t y must be developed.

Laser Enrichment w i t h molecule^.^^ Gaseous UF6 is used i n a l l proposed schemes f o r molecular enrichment, since this is the only compound of uranium w i t h a sizable vapor pressure a t reasonable temperatures. Because the molecule contains zeven atoms and exhibits a high degree of symmetry, i t produces a complicated spectrum of vibrational and rotational excitations. The most interesting vibrational modes from the point of view of l a s e r excitations a r e those which involve motion of the uranium atom and which therefore produce an oscillating e l e c t r i c dipole moment. Only these modes a r e ' l i k e l y t o produce transitions from the ground s t a t e when excited by electromagnet ic energy.

'

The low energies associated w i t h these transiFions lead t o two serious problems f o r l a s e r enrichment i n UF6. The first problem i s the creation of an infrared l a s e r with the correct frequency. The second problem i s related t o the high occupation numbers of the low-energy vibrational s t a t e s a t temperatures where UF6 has a h i g h vapor pressure. Because so many low-lying s t a t e s a r e occupied, i t is impossible t o f i n d a single excitation frequency that will be absorbed by most of the molecules. The presence of these so-called "hot bands" reduces the efficiency of the process very drastically. The second problem i s easily solved, a t l e a s t i n principle, if warm UF6 gas is passed through a supersonic nozzle. The e f f e c t o f the expansion is t o convert most of the kinetic energy of random motion o f the gas i n the reservoir i n t o kinetic energy of translational motion of the gas i n the nozzle. As the gas accelerates

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A-15

through the nozzle, i t becomes c o l d e r and t h e energy stored i n t h e v i b r a t i o n a l

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and r o t a t i o n a l degrees o f freedom o f t h e molecules i s reduced by intermolecular c o l l i s i o n s i n t h e narrow region j u s t downstream o f t h e s l i t .

The molecules can now be i l l u m i n a t e d by a l a s e r beam which has been tuned t o e x c i t e s e l e c t i v e l y molecules c o n t a i n i n g 23511. This technique y i e l d s t h e f i r s t step i n t h e molecular isotope separation

L

process; however, t h i s s e l e c t i v e e x c i t a t i o n does n o t provide a way o f segregating the e x c i t e d molecules. To do t h i s , considerably more l a s e r energy must be absorbed by t h e molecules t o g e t them t o d i s s o c i a t e t o 235UF5 and f l u o r i n e .

L t

I n theory, t h i s

energy can be provided by e i t h e r an i n f r a r e d o r an u l t r a v i o l e t laser. Since i t i s n o t necessary f o r e i t h e r o f these secondary processes t o be i s o t o p i c a l l y s e l e c t i v e , t h e primary demands on t h e u l t r a v i o l e t o r i n f r a r e d l a s e r s are r e l a t e d t o t h e i r energy output and pulse r e p e t i t i o n r a t e s . I n both cases considerably higher powers are r e q u i r e d f o r t h e molecular than f o r t h e atomic processes because

I

h

much l a r g e r numbers o f molecules can be processed i n t h e same p e r i o d o f time.

This

h i g h power requirement f o l l o w s because t h e d e n s i t y r e s t r i c t i o n s apparently are l e s s severe f o r molecules than f o r atoms.

h The d i s s o c i a t e d product must s t i l l be p h y s i c a l l y separated from t h e undissociated m a t e r i a l and s u b s t a n t i a l recombination could occur i f t h e recombination p r o b a b i l i t i e s f o r UF5 and F are high. As w i t h t h e atomic process, t h e molecular process must a l s o overcome formidable t e c h n i c a l d i f f i c u l t i e s before i t becomes a f e a s i b l e production process.

Some

o f these obstacles are: a.

The h i g h p r o b a b i l i t y o f resonant v i b r a t i o n a l energy exchange between the 235UF6 and t h e 238UF6.

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b. c.

The recombination o f d i s s o c i a t e d molecules. An i n f r a r e d high-powered l a s e r tunable t o t h e r e q u i r e d wave l e n g t h f o r t h e primary e x c i t a t i o n must be invented.

d.

The secondary l a s e r must s a t i s f y t h e combined demand o f h i g h pulse energy, r a p i d

e.

r e p e t i t i o n r a t e and h i g h e f f i c i e n c y . The r a p i d and e f f i c i e n t separation o f t h e d i s s o c i a t e d product from t h e depleted tails.

_-

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is

Chemical Exchange Methods The use o f a chemical exchange system t o separate metal isotopes has been under i n v e s t i g a t i o n i n t h e U.S. f o r several years. I n a d d i t i o n t o work i n t h e U.S., t h e French r e c e n t l y have made a l l u s i o n s t o s i m i l a r research. I t has been shown t h a t calcium


A-16

- isotope enrichment can be accomplished using a simple e x t r a c t i o n process i n v o l v i n g the r e l a t i v e l y new class o f compounds known as polyethers. Work i s underway t o determine whether a s i m i l a r process could be used f o r uranium isotope enrichment. The electron exchange e q u i l i b r i u m between U(1V) and U(V1) may r e s u l t i n a s i g n i f i c a n t isotope enrichment. The e x t r a c t i o n o f a s i n g l e uranium c a t i o n without a valence change y i e l d s a small isotope e f f e c t which by i t s e l f would have no p r a c t i c a l use. Combining the two processes leads t o a p o t e n t i a l l y economic process f o r uranium isotope enrichment.

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The electron exchange reaction which occurs i n the aqueous phase can be a

described by Equation 1: 235u4+ + 238Uo 2 2+

+ 238u4+ + 2351)022+

(11

This reaction was reported t o have an a = 1.0014 w i t h 238U concentrating on the U(1V) ion. The solvent e x t r a c t i o n exchange reaction o f the U(V1) i o n can be described by Equation 2:

Although the a f o r Equation 2 i s unknown, theory and experience p r e d i c t t h a t 238U w i l l concentrate i n the aqueous phase. The constructive nature o f the two processes might, therefore, be expected t o r e s u l t i n an a s u i t a b l y l a r g e t o be the basis o f a uranium isotope enrichment process.

ones.

G

From a chemical standpoint, several problems immediately appear as c r i t i c a l It Obviously, one needs an extractant which w i l l separate U(1V) and U(V1).

must operate under some very s p e c i f i c conditions s e t by other portions o f the system. In order t o form the basis o f a useful process, the electron exchange reaction i n Equation 1 must have a half-time, tzi,on the order o f a few seconds. Also, the exchange reaction shown i n Equation 2 must be rapid. Both these reactions must, therefore, be w e l l understood. F i n a l l y , i t must be demonstrated t h a t a s u f f i c i e n t l y l a r g e a e x i s t s under these conditions

.

Based on these exchange reactions and based on a reasonable value o f a (between 1.0014 and 1.002), countercurrent l i q u i d extractors can be s e t up i n t o a cascade arrangement. Further assuming t h a t the exchange reactions and the a are independent o f the r e l a t i v e concentrations o f 235U and 238U, estimates o f the e q u i l i b r i u m time t o achieve 3%enrichment range from approximately three months t o one year. To achieve

90% enrichment, t h e e q u i l i b r i u m time may range from 3 t o 30 years.

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A-1 7

Aerodynamic Methods Both the separation nozzle and the stationary-walled c e n t r i f u g e can be classed as aerodynamic processes.

These are considered t o be competitive processes

by t h e i r proponents and plans f o r t h e i r implementation are w e l l advanced.

Research

e f f o r t s have been d i r e c t e d a t several other aerodynamic methods such as t h e vortex tube, t h e separation probe, crossed beams, v e l o c i t y s l i p and t h e j e t membrane.

None

o f these appear a t t h e present time t o o f f e r t h e promise of t h e two aforementioned aerodynamic processes, although an expanded e f f o r t i s proceeding on the j e t membrane process.

Commonly known as t h e Muntz-Hamel process, i t involves t h e p e n e t r a t i o n o f

a stream o f UF6 gas i n t o an expanding j e t o f e a s i l y condensible c a r r i e r gas.

The

l i g h t e r 235UF6 molecules penetrate t h e j e t more e a s i l y than t h e heavier 238UF6

A tube placed on t h e a x i s o f the j e t c o l l e c t s t h e enriched UF6. The depleted UF6 flows o u t o f t h e o t h e r end o f the s c a t t e r i n g chamber, a f t e r the c a r r i e r

molecules.

gas i s separated from i t by condensation. Plasma-Based Processes Since a plasma can be made t o r o t a t e a t speeds greater than t h a t o f an u l t r a centrifuge,

i t occurred t o various i n v e s t i g a t o r s t h a t such h i g h speed gas r o t a t i o n

w i t h o u t t h e use o f r e v o l v i n g equipment might p o s s i b l y be developed i n t o a more e f f i c i e n t isotope separation process than t h a t based on a mechanical c e n t r i f u g e . Five papers on t h i s t o p i c were presented a t the I n t e r n a t i o n a l Conference on Uranium Isotope Separation i n London i n March 1975. The authorsâ&#x20AC;&#x2122; assessment o f the prospects f o r such a process r a n t h e gamut from h i g h l y optimistic--technology i s simple and w e l l known so t h a t minimal development w i l l be required--to pessimistic--a r o t a t i n g plasma process cannot p o s s i b l y be economically competitive. To our knowledge, no one has separated uranium isotopes by means o f t h e plasma centrifuge. Since t h a t time, several o t h e r plasma-based processes have been proposed.

Of

a l l these processes, t h e c u r r e n t l y most feasible seems t o be t h e Plasma I o n Enrichment process ( t h e Dawson separation process). I n t h i s process a plasma o f uF6 ( o r o f uranium atoms) within a strong uniform magnetic f i e l d i s exposed t o a low energy radio-frequency wave resonant w i t h t h e c y c l o t r o n frequency o f t h e 235UF6 ions. The r o t a t i o n thereby imparted p r e f e r e n t i a l l y t o t h e 235UF6 i o n s enables t h e 235U t o be separated from t h e 238U by p r o p e r l y placed c o l l e c t i o n plates. This method has been used successfully t o e n r i c h macroscopic samples of potassium.

The c o l l e c t o r was a cooled tungsten r i b b o n having a voltage b i a s t o c o l l e c t s e l e c t i v e l y t h e e x c i t e d ions. The potassium vapor was contact i o n i z e d a t

t h e entrance t o t h e mass spectrometer.

To e l i m i n a t e spurious e f f e c t s , samples were

c o l l e c t e d under t h r e e conditions of rf e x c i t a t i o n :

(1) no rf; (2) e x c i t a t i o n a t t h e

39K c y c l o t r o n frequency; and ( 3 ) e x c i t a t i o n a t the 41K c y c l o t r o n frequency.

The


i

A-18

r e s u l t i n g r a t i o s o f 41K/39K abundance as measured by the mass spectrometer were, respectively, 0.07 (the natural abundance), 0.02 and 4. The abundance r a t i o o f 4 corresponds t o a more than t e n f o l d enrichment o f '+lK. I n a d d i t i o n t o potassium ions, work has been done on neon, argon, xenon and uranium toward resolving the i o n cyclotron resonances f o r i n d i v i d u a l p o s i t i v e ions. The work w i t h uranium i s proceeding toward estimates o f r e a l i s t i c operating parameters ( i o n densities, magnetic f i e l d strength, i s o t o p i c e x c i t a t i o n energies, device length, i o n temperatures, and c o l l e c t o r types).

A second process involves the achievement o f a UF6 plasma by chemi-ionization. UF6 molecules are accelerated by expansion w i t h an i n e r t c a r r i e r gas through a supersonic j e t . A cross beam o f a l k a l i metal molecules r e s u l t s i n the formation o f NA' o r Cs' and UF6'. A radio-frequency quadrupole mass f i l t e r d e f l e c t s the 238UF6 out o f the plasma beam, p e r m i t t i n g the separation o f the two isotopes by c o l l e c t i o n o f the two beams on separate b a f f l e s cooled by l i q u i d nitrogen. This process seems t o have l e s s p o t e n t i a l than the f i r s t .

r;

c I

L

Comparison o f Advanced Separation Processes The estimated costs o f the processes mentioned are compared i n Table A-3 w i t h t h a t o f gaseous d i f f u s i o n . With two exceptions, the t a b l e i s based on process evaluations made by the Nuclear D i v i s i o n o f the Union Carbide Corporation17 f o r ERDA. For the exceptions, which are the FRG's separation nozzle and South A f r i c a ' s stationary-walled centrifuge, the comparison i s based on published statements by the developers of the process. O f a l l the processes l i s t e d , only the costs f o r the centrifuge, and possibly f o r the separation nozzle, are known w i t h any degree o f certainty

.

L


A-19

Table A-3.

Comparison o f Process Economics

i

Specific Capital Investment

Power cost

Operating costs Other Than Power

i d

Centri f uge

>

<

>

.I .

Separation Nozzle'

<

>

c

Stationary-Wal l e d Centrifuge*

--

-

?

LIS-Atomi c

<

<

>

<

>

<

7

ii

LIS-Molecul a r Ch. Exchange:

hi

L i: i

U1"(aq)-UV1(org)

Other Aerodynamic Processes

>

>

Plasma:

>

<

>

<

<

>

Chemi-ionization

Plasma I o n Enrichment (Dawson Process) -

*Based on estimates made by t h e process developers.

DEFINITION

-

>,<

?

OF SYMBOLS: Approximately equal t o t h e d i f f u s i o n process. Greater than o r 'less than the d i f f u s i o n process, respectively. Unknown.

References f o r Appendix A 1.

E. 6. Kiser, Jr.,

"Review o f U.S. Kansas City, Mo. ( A p r i l 1977).

Gas Centrifuge Program,"

AIF Fuel Cycle Conf. '77,

2.

USAEC, "AEC Gaseous Diffusion P l a n t Operations,"

3.

H. D. Smythe, Atomic Energy f o r M i l i t a r y Purposes, Princeton U n i v e r s i t y Press, Princeton, N.J., Aug. 1945.

4.

W. E. Groth,

0R0-684 (Jan. 1972).

K. Beyerle, E. Nann, and K. H. Welge, "Enrichment o f Uranium Isotopes by t h e Gas Centrifuge Method," 2nd I n t ' l . Conf. on t h e Peaceful Uses o f Atomic Energy, Geneva , Switzerland (Sept 1958).

.

5.

.

t '

Li

1

i"

J. Los Kistemaker and E. J. Z. Veld Huyzen, "The Enrichment o f Uranium Isotopes w i t h Ultra-Centrifuges," 2nd I n t ' l . Conf. o f t h e Peaceful Uses o f Atomic Energy, Geneva , Switzerland (Sept 1958).

6.

G. Zippe, "The Development o f Short Bowl U l t r a Centrifuges," 1960).

7.

K. Cohen, The Theory of Isotope Separation, McGraw-Hill Book Company, New York, 1951.

ORO-315 (June 15,


1

A-20

8.

W. R. Voigt, "Enrichment Policies," AIF Fuel Cycle Conf. '77, Kansas City, Mo. (April 1977).

9.

E. Von Halle, "Summary Review of Uranium Isotope Separation Methods Other Than Gaseous Diffusion and Gas Centrifugation," ORO-690 (Feb. 23, 1972).

i:

10. A. J. A. Roux and W. L. Grant, "Uranium Enrichment i n South Africa," presented t o European Nuclear Conference, Paris, France (Apri 1 1975). 11.

Nuclear News, p. 80 (June 1977).

12. S. Blumkin, "Survey of Foreign Enrichment Capacity, Contracting and Technology: Jan. 1976-Dec. 1976," K/OA-2547, P t . 4 (April 18, 1977). 13.

Letter, R. J. Hart, ERDA-ORO, t o W. R. Voigt, ERDA-HQ, "Interim Uranium Enrichment Long-Range Operating Plan," (August 9, 1977).

14. M. Benedict, e t a l . , "Report of Uranium Isotope Separation Review Ad Hoc Committee," 0R0-694 (June 2, 1972). 1

i

15. Allas S. Krass, "Laser Enrichment of Uranium: The Proliferation Connection,' Science ' Vol. 196, No. 4291, p. 721-731 (May 13, 1977). 16. J. M. Dawson, e t a l . , "Isotope Separation i n Plasmas Using Ion Cyclotron Resonance," TRW Defense & Space Systems, Redondo Beach, Calif. 17.

I:

p. R. Vanstrum and S. A. Levin, "New Processes f o r Uranium Isotope Separation," IAEA-CN-36/12 (11.3), Vienna, Austria (1977).

I;

I'


B-1

ECONOMIC DATA BASE USED FOR EVALUATIONS OF NUCLEAR POWER SYSTEMS

Appendix B. la

M. R. Shay, D. R. Haffner, W. E. Black, T. M. Helm, W. G. J o l l y , R. W. Hardie, and R. P. Omberg

t

Hanford Engineering Development Laboratory The economic data base used i n .the assessment o f t h e impact o f denatured f u e l cycles i n t h e various nuclear systems options described i n Chapter 6 was j o i n t l y developed by Combustion Engineering, Oak Ridge National Laboratory, United Engineers and Constructors, Argonne National Laboratory, Resource Planning Associates, Hanford Engineering Development Laboratory, DOE D i v i s i o n o f Uranium Resources and Enrichment, and DOE D i v i s i o n o f Nuclear Research and

Applications.

L I

I,

The data base includes c a p i t a l costs, operation and maintenance costs, f u e l f a b r i c a t i o n and reprocessing costs, capacity factors, money costs, and u n c e r t a i n t i e s . The d e f l a t e d and present-valued c a p i t a l costs f o r LWRs, SSCRs. HTGRs, CANDUs, and FBRs,

excluding i n t e r e s t d u r i n g construction, a r e shown i n Table B-1. i n c l u d i n g i n t e r e s t d u r i n g c o n s t r u c t i o n are shown i n Table B-2.

The same c a p i t a l costs

I n e i t h e r case, t h e stream o f

expenses i n c u r r e d d u r i n g t h e c o n s t r u c t i o n o f the p l a n t i s discounted t o t h e date o f s t a r t u p

lj

and i s measured i n d o l l a r s o f constant purchasing power.

The u n c e r t a i n t y ranges included i n

Table 8 - 2 represent c u r r e n t best estimates of the most probable v a r i a t i o n s i n c a p i t a l costs. For f l e x i b i l i t y , t h e u n c e r t a i n t i e s are expressed r e l a t i v e t o the reference LWR c a p i t a l cost.

ir

Table B-1. C a p i t a l Costs o f Power Plants Excluding I n t e r e s t During Construction Power P l a n t Type

1 Li

c I L L'

Costs ($/kWe)*

LWR

500 520 + 39

SSCR HWR

( f o r D20) = 558

605 t 156 ( f o r D20) = 761 560 t o 580 625 t o 875

HTGR FBR

The operation and maintenance costs f o r the same power p l a n t s are shown i n Table 6 - 3 . The higher costs f o r the SSCR and the CANDU over t h e standard LWR a r e due t o the heavy water replacement requirement and the necessity f o r performing some maintenance i n atmospheres containing t r i t i u m . A d d i t i o n a l minor r e a c t o r costs are given i n Table 8-4.

*

Based on 7/1/76 d o l l a r s . Table B-2.

C a p t i a l Costs of Power Plants I n c l u d i n g I n t e r e s t During Construction

cost ($/kWe)*

Power P l a n t TY Pe LWR

Cost Re1a t i v e t o LWR Cost

625

cost Uncertainty 95% t o 105% reference c o s t

SSCR

650

+ 40 (heavy water) = 690

+lo%

105% t o 120% o f LWR c o s t

HWR

755

+ 160

+46%

120% t o 150% of LWR c o s t

(heavy water) = 915

HTGR

715

+14%

105% t o 125% o f LWR cost

FBR

800

+28%

125% t o 175% o f LWR c o s t

*Based on 1/1/77 d o l l a r s .


8-2

The f u e l f a b r i c a t i o n costs f o r the various r e a c t o r types are shown i n Table B-5 as a function o f time beginning w i t h the expected i n t r o d u c t i o n date f o r a p a r t i c u l a r reactor and fuel design. I f a p a r t i c u l a r reactor and f u e l design should prove successful, f a b r i c a t i o n costs should decrease as l a r g e r plants w i t h higher throughput r a t e s are constructed. The decrease i n f a b r i c a t i o n costs over the f i r s t decade a f t e r i n t r o d u c t i o n i s simply i n d i c a t i v e o f a t r a n s i t i o n from small f a b r i c a t i o n plants w i t h high u n i t costs t o l a r g e r f a b r i c a t i o n plants w i t h lower u n i t costs. These costs are a strong function o f the f i s s i l e isotope and a weak function o f the f e r t i l e isotope. The s e n s i t i v i t y t o the f i s s i l e isotope i s caused e i t h e r by the spontaneous f i s s i o n associated w i t h high-exposure f i s s i l e plutonium o r by the gamma a c t i v i t y associated w i t h high-exposure 233U. The costs are based on the assumption t h a t f u e l s containing 23% are fabricated on a l i n e w i t h contact operation and contact maintenance, f u e l s containing f i s s i l e plutonium are fabricated on a l i n e w i t h remote operation and contact maintenance, and f u e l s containing 233() are fabricated on a l i n e w i t h both remote operation and remote maintenance. The expected v a r i a t i o n s i n f u e l f a b r i c a t i o n costs (cost uncertainties given i n footnote b o f Table 8 - 5 ) represent the upper and lower cost boundaries anticipated f o r f a b r i c a t i o n costs and are expressed as percentages. For example, the expected f a b r i c a t i o n cost f o r plutonium-bearing LWR f u e l w i t h uncertainties applied ranges from $306 per kg HM (-10% o f reference) t o $510 per kg HM (+50% o f reference) f o r year 2001 and beyond.

Table 8-3. Power Plant Operation and Maintenance Costs {=[Fixed +(Variable x Capacity Factora)]xPower}

Power Plant Type

Fixed Cost ($/kWe-yr)b

Va r3ab 1e

LWR

3.6

1.9

SSCR

4.8

1.9

HWR

8.4

1.9

HTGR

3.6

1.4

FBR

4.1

2.3

a

See Table B-9 f o r capacity factors. bBased on 1/1/77 dollars.

Table 8-4.

Minor Reactor Costs

Property Insurance Rate

0.0025

Capital Replacement Rate

0.0035

Nuclear L i a b i l i t y

58 x lo4 $/yr

The expected reprocessing costs are shown i n Table 8-6. These costs were obtained by estimating the c a p i t a l and operating costs associated w i t h each o f f i v e stages o f the reprocessing process. The stages were: headend, solvent extraction, product conversion, off-gas treatment, and waste treatment. The costs are shown as a function o f time r e f l e c t i n g the t r a n s i t i o n from a new industry consisting o f small plants w i t h high u n i t costs t o a mature industry consisting o f l a r g e r plants w i t h lower u n i t costs. The expected costs f o r spent f u e l shipping, waste shipping, and waste storage are also included i n Table 8-6, as w e l l as the t o t a l costs f o r a l l these processes. The t p t a l cost uncert a i n t y f a c t o r f o r a l l f u e l types i s estimated t o be a 50% increase f o r the reference values. Thus, the t o t a l reprocessing cost f o r LWR f u e l w i t h the uncertainty included ranges from $220 t o $330 per kg HM f o r year 2001 and beyond. I t should be noted t h a t i t i s assumed here t h a t a p o l i c y decision w i l l have been made i n t i m e f o r the f i r s t reprocessing p l a n t t o be i n operation by 1991. A l l f u e l discharged from the reactor p r i o r t o t h i s date i s

u L

L L

c

L

L


B-3

assumed t o have been stored, w i t h the Table 8.5.

i;

Reactor Fuel Fabrication Costsa

Reactor Type LWR-U5(LE)/U LWR-U5( DE)/U/Th LWR-U3( DE)/U/Th LWR-PU/U

LWR-PU/Th SSCR-US( LE)/U SSCR-U3( DE)/U/Th SSCR-PU/Th

100 (1991 * 2089)c 880 (1991) + 550 (2001) 550 (1991) + 340 (2001)

HWR-U5( NAT)/U HWR-U5(SEU)/U HWR-U5( DE)/U/Th HWR-U3( DE)/U/Th

60 60 140 560 320 320

HWR-PU/U

1

t

Cost ($/kg HMlb Over F i r s t Introductlon 100 (1969 + 2089)= 230 (1987) + 140 (1997) 880 (1991) + 550 (2001) 550 (1991) + 340 (2001) 550 (1991) + 340 (2001)

Decade A f t e r

HWR-PU/Th

(1995 (1995 (1995) (1995) (1995) (1995)

+ +

+ + +

+

2089)c 2089)c 85 (2005) 350 (2005) 200 (2005) 200 (2005)

cessing.

A f t e r t h e spent f u e l s t o c k p i l e

has been reduced t o zero, the o u t - o f - r e a c t o r time r e q u i r e d f o r reprocessing and refabr i c a t i o n i s assumed t o be two years. The long-run marginal costs estimated f o r U308 o r e as a f u n c t i o n o f the cumulative supply a r e shown i n Table B-7.

As noted i n

Chapter 6, t h e U308 estimates have been provided by DOE'S D i v i s i o n o f Uranium Resources and Enrichment (URE), t h e highc o s t supply being based on t h e assumption t h a t approximately 2.5 m i l l i o n tons o f U308

w i l l be a v a i l a b l e from conventional uranium o r e resources and t h e intermediate-cost

HTGR-U5 (LE)/U HTGR-U5 (DE)/U/Th

HTGR-U5( HE)/Th C/Th + U = 150 C/Th + U = 238 C/Th + U = 335 C/Th + U = 400 C/Th + U = 650 HTGR-U3( DE)/U/Th HTGR-U3/Th C/Th + U = 150 C/Th + U = 238 C/Th + U = 335 C/Th + U = 400.. C/Th + U = 650

spent f u e l s t o c k p i l e being reduced i n an o r d e r l y manner a f t e r the advent o f repro-

supply being based on t h e assumption t h a t 340 500 660 760 1220

1995) 11995) (1995) (1995) (1995)

+

+

+ + +

210 300 400 470 770

(2005) (2005) (2005) (2005) (2005)

approximately 4.5 m i l l i o n tons o f U308 w i l l be a v a i l a b l e . I n e i t h e r case, i t i s assumed t h a t shales can be mined a f t e r t h e conventional resources a r e depleted.

The

c o s t o f e x t r a c t i n g t h e shales increases 860 1220 1640 2000 3200

(1995) (1995) (1995) (1995) (1995)

+

+ + + +

470 670 900 1100 1750

(2005) (2005) (2005) (2005) (2005)

from $125/lb t o $240/lb f o r t h e high-cost supply case and from $100/lb t o $180/lb f o r t h e intermediate-cost supply case.

It

i s important t o note t h a t t h e long-run HTGR-PWTh C/Th = 238 FBR-PU-U core

u 1 1

FBR-Pu-Th core FBR-U3-U core FBR-U axial blanket FBR-U radial blanket FBR-Th axial blanket FBR-Th radial blanket

1220 (1995)

+

670 (2005)

1750 1750 3000 35 250 35 250

+

950 950 1650 25 150 25 150

(2001) (2001) (2001) (2001) (2001) (2001) (2001)

-+ + + +

+ +

(2011) (2011) (2011) (2011) (2011) (2011) (2011)

'Fabrication costs based on the following: f o r LWR and SSCR, a 17 x 17 pin assembly (374-mil-OD pin); f o r the HWR, a 37-pin CANDU assembly -20 in. long (531-mil-OD pin); f o r the HTGR, standard carboncoated uranium carbide f i s s i l e microspheres formed i n t o cylindrical rods located i n a hexagonal graphite block; and f o r the FBR, a 217-pin assembly i n a hexagonal duct (310-mil-00 pin). bUncertainities on fabrication costs: 235U-bearing fuels, no uncertainty; Pu-bearing fuels, -10% t o 50% increase; 233U-bearing fuels, -10% t o 50% increase. CCosts assumed t o remain constant.

marginal costs shown i n Table B-7 a r e l a r g e r than the forward c o s t s shown i n Table 6.1-1 o f Chapter 6 because the long-run marginal c o s t s c o n t a i n the c a p i t a l c o s t o f f a c i l i t i e s c u r r e n t l y i n operation, p l u s a normal p r o f i t f o r the i n d u s t r y .

The long-run marginal

c o s t s are more appropriate f o r use i n a nuclear s t r a t e g y analysis. The enrichment costs and t a i l s compositions assuming e5 t h e r a c o n t i n u a t i o n of t h e gaseous d i f f u s i o n technology o r the deployment of an advanced enrichment technology a r e shown i n Table 8-8.

I t was

assumed t h a t i f the gaseous d i f f u s i o n technology i s continued the t a i l s composit i o n w i l l be s t a b i l i z e d a t 0.0020 and t h a t t h e c o s t o f enrichment w i l l increase t o $80/SWU i n 1987 and remain constant thereafter.

I f an advanced enrichment technology


8-4

Table B-6.

Reprocessing, Shipping, and Waste Storage Costs f o r Various Reactor Types Costs ($/kg HM)

Reactor Type

LWR SSCR HWR

HTGR FBR

Reprocessing Costs

Over First Decade9

Spent Fuel ShiPP’S9

Waste Shipping

Waste Storage

Total Costs

costs

costs

Over First Decade

costs 225 225 225 800 500

(1991)* (1991) * (1995) + (1995)+ (2001)+

150 150 150 400 200

(2001) (2001) (2005) (2005) (2011)

15 15 10 85 80

After IntroductionC 10 10 5 35 50

45 45 15 65 115

295 295 255 985 745

(1991) + (1991)+ (1995) + (1995)+ (2001) +

220 (2001) 220 (2001) 180 (2005) 585 (2005) 445 (2011)

aFissile storage costs after reprocessing = f2/g-yr for ‘’’0 and fissile plutonium. bTotal costs for throwaway cycle are spent fuel shipplng costs plus $100/kg HM. ‘50% uncertainty on total costs for all reactor types.

i s deployed, the t a i l s composition would decrease continuously from 0.0020 t o 0.0010 between the years 1980 and 2000 as the i n s t a l l e d capacity o f the advanced technology increased, and the cost o f a u n i t o f separative work would decrease t o approximately 60% o f t h a t o f the gaseous d i f f u s i o n process. I t was also assumed t h a t t h e t a i l s composition would f u r t h e r decrease from 0.0010 t o 0.0005 between the years 2001 and 2030 due t o improvements i n technology, while t h e cost o f a u n i t o f enrichment would remain constant during t h i s period. The t a i l s composition and enrichment cost were assumed t o remain constant thereafter.

L

c c

The capacity f a c t o r s o f a p l a n t throughout i t s 30-yr l i f e t i m e are shown i n Table B-9. The capacity f a c t o r increases from 60% t o 72% during the f i r s t 3 yr o f operation and remains a t 72% during the subsequent 14 yr. It then decreases continuously as the forced outage r a t e increases and as the p l a n t i s s h i f t e d from a base-load u n i t t o an intermediate-load u n i t . The long-term r e a l cost of money t o the e l e c t r i c u t i l i t y i n d u s t r y i s shown i n Table 6-10. These costs were developed by analyzing the deflated cost o f debt and e q u i t y t o the i n d u s t r y over the past 30 yr. The long-term deflated cost o f debt has been 2.5%/yr and the long-term deflated cost o f e q u i t y has been 7.0%/yr. Assuming the i n d u s t r y t o be funded a t approximately

55% debt and 45% equity, the long-term r e a l money cost i s approximately 4.5%/yr. The combined e f f e c t s o f c a p i t a l , f u e l fabrication, and reprocessing ( o r permanent disposal) c o s t uncertainties on the l e v e l i z e d t o t a l power costs f o r i n d i v i d u a l reactor and f u e l cycle options are shown i n Fig. B-1. These costs represent t y p i c a l nonfuel components whose uncertainties are e a s i l y quantified. Figures 6-2a and B-2b show the r e l a t i o n s h i p of t o t a l power costs t o t h e U & p r i c e f o r f o u r reactors on the throwaway f u e l cycle. The s e n s i t i v i t y o f the t o t a l power costs t o the U308 p r i c e was analyzed f i r s t by assuming t h a t the p r i c e remained constant over the 30-yr l i f e o f the reactor, and second by assuming t h a t the p r i c e increases i n r e l a t i o n t o the r a t e o f consumption (see Fig. 8-3). Thus, the t o t a l power costs i n Fig. B-2b are given f o r a reactor s t a r t i n g up w i t h the U308 p r i c e shown on

L L I]

c L


8-5

Table 6-7. Marginal Costs o f U308 as a Function o f Cumulative Supplyalb Long-Run Marginal Cost Q u a n t i t y of U308 (106 tons) ($/1 b )

Table 8-8. T a i l s Composition and Enrichment Costs Tails Composit ion (235U F r a c t i o n )

Time

Cost($/SWU)

Intermediate-Cost U308 Supply

0.0 0.25 0.75 1.25 1.75 2.5 3.5 4.25

0.25

14

0.75

23

1.25

33

1977 t o 1986

0.0020

75

44

0.0020

80

-

1.75 2.5

1987 t o 2089

-

3.5 4.25

61

- 4.75

80 107

5.75

6.0

165

6.00

- 8.5

165

5.25

5.25

128

5.75

143

above 8.5

1969 t o 1976

53

-

4.75

Gaseous D i f f u s i o n Technology

-

180

0.0020

50

Advanced Technology

1981 t o 2000

50

0.0020

1969 t o 1976 1977 t o 1980

0.0020

75

0.0020 t o 0.0010

2001 t o 2030 0.0010 t o 0.0005 2031 t o 2089 0.0005

Table B-9.

75 t o 55

.

55 55

P l a n t Capacity Factors

High-Cost U308 Supply

0.0 0.25 0.75 1.25 1.75 2.25 2.75 3.00 3.25 3.75 4.25 4.75

- 0.25 -

-

-

14

1.75

54

1 2 3

2.25

84

4

2.75

128

0.75

24

1.25

35

3.00 3.25 3.75

- 4.25 - 4.75 - 6.5

above 6.5

60.0

20

65.7

66.0

21

64.1

72.0

22

62.6

72.0

23 24

61 .0

25 26

57.9 56.3

27 28 29

54.7 53.1

158 158 173 180 180

15

72.0

16

72.0 70.4

17 18 19

68.9 67.3

30

59.4

51.6 50.0

210 240 Table B-10.

aFor those cases i n which p l a n t s e l e c t i o n was determined by uranium u t i l i z a t i o n a l i m i t o f 3 m i l l i o n tons of ore are assumed a t below $150/lb U308 f o r the high-cost U308 supply and 6 m i l l i o n tons for t h e i n t e r medi ate-cost supply. bCost o f converting U308 t o UF6 = $3.50/kg o f u.

Long-Term Real Costs o f Money

Debt I n t e r e s t

2.5%

Equity Interest

7.0%

F r a c t i o n Debt

0.55

Fraction Equity

0.45

E f f e c t i v e I n t e r e s t Rate

4 525%


L 8-6

''111

-

5 17 E 16 15 4

I

REACTOR

OPTIONS: 6

f

l

I

LWR

SSCR

HWR

HTGR

A

h

h

h

I

I

7,

I

I

I'C

I

I

,'[I

I

I

FER

I'

ff7

HEDL 78M-WO.W

Fig. B-1. Sensitivity of Total Levelized Power Cost t o Capital, Fabrication, and Reprocessing Cost Uncertainties.

II

the abscissa. The major difference between the two methods of analysis i s the u308 price a t Which reactor options incur the same total power cost. For example, whereas a t a constant U & price the PWR and HWR options have the same power generation cost a t * $160/lb U308 f o r an increasing U J O ~price they have the same cost a t * $130/lb U&.

From the data shown i n Fig. B-1 i t i s clear that the total power cost for each reactor and fuel cycle option i s dominated by uncertainties. The uncertainty e f f e c t produces a significant overlap between reactor power costs. In addition, i t is evident from Fig. B-2 t h a t fuel costs, viz., U308 prices, also significantly a f f e c t n o t only the levelized power costs b u t also the competitive relationship between reactor options. Therefore, i t is d i f f i c u l t t o classify reactors as e i t h e r more economical or less economical based solely on power generation cost estimates.

'I


8-7

LJ

I

I

I

I

I

I

(a) U308 p r i c e constant over 30-yr reactor l i f e .

L

U308 PRICE, $/lb HEDL 7805-090.16

I

I

I

I

I

I

I

I

120

140

(b) U308 p r i c e increases w i t h r a t e o f consumption.

HWR-US(SE)/II

\

L1

4 '

a

.E P

18

Y W

5

2 16 ::

0

m 4

0 0

I

I

40

60

I

I

80 100 U308 PRICE AT STARTUP, $/lb

160

HEDL 78Q545'0.17

Fig. 8-2. Effect o f Throwaway Cycle.

U308

Price on Total Power Cost for Reactors Operating on


8-8

YEAR

Fig. 8-3.

Time-Dependent Behavior o f U308 Price.

c


DETAILED RESULTS FROM EVALUATIONS OF VARIOUS NUCLEAR POWER SYSTEMS UTILIZING DENATURED FUEL

Appendix C.

M. R. Shay, D. R. Haffner, W. E. Black, T. M. Helm, G. J o l l y , R. W. Hardie, and R. P. Omberg Hanford Engineering Development Laboratory

W.

This appendix presents d e t a i l e d r e s u l t s from t h e c a l c u l a t i o n s performed f o r t h e economic/resource evaluation o f denatured nuclear reactors operated i n concert w i t h o t h e r r e a c t o r s t o form nuclear-based power generation systems. For purposes o f comparison, i t a l s o presents r e s u l t s f o r s i m i l a r systems t h a t do n o t u t i l i z e denatured f u e l . As pointed o u t i n Chapter 6, n i n e d i f f e r e n t nuclear p o l i c y options were examined w i t h four cases under each option. The r e s u l t i n g cases can be c l a s s i f i e d as shown i n Table C-1, where t h e l e t t e r s L, S, G, and H i n d i c a t e t h e thermal converter o p t i o n employed i n each case. For a l l cases i d e n t i f i e d w i t h an L, t h e o n l y converters used are LWRs. For cases i d e n t i f i e d w i t h an S, SSCR converters are used in addition to LwRs. S i m i l a r l y , f o r cases i d e n t i f i e d w i t h H and G, t h e converters used are HWRs and HTGRs respectively, both again i n combination w i t h LWRs. Under Options 3, 6, 7, and 8, FBRs are a l s o included i n t h e nuclear systems. I n a d d i t i o n t o these 36 cases, Case 1L was r e c a l c u l a t e d f o r a standard LWR alone; t h a t i s , t h e LWR with an extended discharge exposure, which i s included i n Case l L , was e l i m i n a t e d from t h e system.

This case i s i d e n t i f i e d i n t h i s appendix as Case 1E.

Table C-1. Options Throwaway Option (1 )

Nuclear P o l i c y Options* LWR

SSCR

HTGR

HWR

1L

1s

1G

1H

2L 3L

2s

2H

3s

26 36

Pu/U Options With Converters Only (2) With Converters and Breeders (3)

3H

Denatured Uranium Options w i t h Converters Only Plutonium Throwaway (4)

4L

4s

46

4H

Plutonium Miminization (5U)

5UL

5us

5UG

5UH

Plutonium "Transmutation" (5T)

5TL

5TS

5TG

5TH

6L

6s

66

6H

7L

7s

76

7H

8L

8s

86

8H

Denatured Uranium Options w i t h Converters and Breeders

'

L i g h t "Transmutation" Rate (6) L i g h t "Transmutation" Rate, Denatured Breeder (7) Heavy "Transmutation" Rate, Denatured Breeder (8)

*See Table 6.1-5 i n Chapter 6 and Tables C-2 and C-4 I n t h i s appendix f o r i d e n t i f i c a t i o n of s p e c i f i c r e a c t o r types i n each case.


c c-2

I n a l l cases the reactors operating on plutonium o r on h i g h l y enriched uranium were assumed t o be r e s t r i c t e d t o secure energy centers, while those operating on low-enriched,

I]

s l i g h t l y enriched, natural, o r denatured uranium were permitted t o operate outside the

c

centers.

The s p e c i f i c reactors used f o r each case, and t h e i r locations, are given i n

Table 6.1-5 o f Chapter 6.

L

A l l cases were run assuming 350 GWe o f i n s t a l l e d nuclear capacity i n the year 2000 and a net increase i n i n s t a l l e d capacity o f 15 GWe per year thereafter. assumed t o have a 30-yr l i f e t i m e . energy demand

--

Each new p l a n t was

For Option 1, some a d d i t i o n a l cases were run f o r a lower

200 GWe i n the year 2000 and a net increase o f 10 GWe per year thereafter.

These latter,cases a r e - i d e n t i f i e d w i t h a C following the case number (i.e.,

cases lLEC, lLC,

etc.). I n the r e s u l t s presented here, p a r t i c u l a r emphasis i s given t o uranium u t i l i z a t i o n , separative work u t i l i z a t i o n , and energy-support r a t i o s .

Two important c r i t e r i a are t o be considered when analyzing uranium u t i l i z a t i o n of reactor systems. The f i r s t i s the a b i l i t y o f the system t o meet the s p e c i f i e d nuclear energy demand w i t h the a v a i l a b l e U308 supply. For these calculations two d i f f e r e n t supplies were assumed: 3 m i l l i o n and 6 m i l l i o n ST below

L

$160/1 b U308, corresponding t o a high-cost and an intermediate-cost supply, respectively.

c

(As shown i n Appendix D, nuclear power plants do not compete well a t higher U308 costs.) The second c r i t e r i o n i s the c a p a b i l i t y o f the uranium industry t o discover, mine and m i l l the ore a t a r a t e adequate t o s a t i s f y the demand for uranium.

The s p e c i f i c a t i o n o f the o v e r a l l

maximum production r a t e i s d i f f i c u l t t o postulate because o f the p o s s i b i l i t y o f importing U3Oi

and because o f the d i f f i c u l t i e s t h a t might be encountered i n developing uncertain

resources. As pointed out i n Section 7.4.4 o f Chapter 7, the DOE Uranium and Enrichment D i v i s i o n has estimated t h a t by developing known and p o t e n t i a l reserves domestic mining and m i l l i n g could sustain 60,000 ST o f U30s per year. When analyzing enrichment u t i l i z a t i o n , the same two c r i t e r i a ment capacity

- were also used,

-

t o t a l amount and enrich-

1:

the more meaningful being the capacity since enrichment i s

not a l i m i t e d natural resource l i k e uranium.

c

For the cases i n which 3 m i l l i o n ST o f uranium below $160/lb U308 was assumed, the lack o f low-cost U308 dominates the p l a n t selection because the amount o f ore a v a i l a b l e i s inadequate f o r meeting the projected nuclear energy demand. As a r e s u l t , resource-efficient reactors are constructed regardless o f t h e i r cost.

c

With a U308 supply below $160/lb as

l a r g e as 6 m i l l i o n ST, however, most systems are no longer dominated by the l a c k o f U308,

L

and the r e l a t i v e t o t a l power costs o f the i n d i v i d u a l reactors p l a y a more important r o l e . I n fact, i f the system i s n o t l i m i t e d i n any way by the supply o f U308, then the s o l u t i o n i s determined s o l e l y by economics.

The r e s u l t s i n t h i s case become more tenuous because o f the

uncertainty i n c a p i t a l costs, f a b r i c a t i o n costs, reprocessing costs, etc. The cumulative nuclear capacities t h a t could be constructed through the year 2050 f o r the various cases are shown i n Table C-2.

Only those cases t o t a l i n g 1959 GWe w i l l have

.

d

LJ

L


1

i; c-3

Cumulative Nuclear Capacity B u i l t Through Year 2050 w i t h Various Nuclear P o l i c y Options

Table C-2.

Li

(Adequate Capacity = 1959 GWe) Advanced Converter option

Option Capacity (GWe)

1E*

1

2

3

4

b 572

lkd

t L L

6

7

8

594

953

1959

945

120s

1027

1959

1959

1547

607

1043

1959

1071

1423

1275

1959

1959

1943

667

987

1959

1334

1747

1505

1959

1959

1959

603

1417

1959

855

1064

1004

1950

1959

1591

Intermediate-Cost U308 Supply

1135

*System with standard

I

5T

High-Cost U30R Supply

L

hd

5u

1193

1783

1959

1852

1921

1864

1959

1959

1956

1271

1937

1959

1943

1959

1959

1959

1959

1959

1497

1921

1959

1943

1959

1959

1959

1959

105;

1320

1959

1959

1794

1924

1844

1959

1959

1559

LWR only.

met t h e p r o j e c t e d nuclear demand under t h e c r i t e r i a o f an i n s t a l l e d capacity o f 350 GWe i n y e a r 2000 and an increase o f 15 GWe per year t h e r e a f t e r . *

With t h e high-cost U308 supply

I '

some o f t h e systems f a l l f a r s h o r t of s a t i s f y i n g t h e demand; i n f a c t , the o n l y nuclear systems

L;

t h a t f u l l y meet the demand are those i n c l u d i n g FBRs (Options 3, 6, 7, and 8). The throwaway option, i n p a r t i c u l a r , b u i l d s l e s s than a t h i r d o f the desired nuclear p l a n t s . O f t h e cases t h a t do n o t i n c l u d e FBRs, those employing HWRs come c l o s e s t t o meeting the demand. One HTGR case (26) i s a l s o c l e a r l y s u p e r i o r t o most of the o t h e r cases. This i s t o be expected since Case 26 includes t r a d i t i o n a l HTGRs t h a t are fueled w i t h h i g h l y enriched 235U and a l s o w i t h 3U/Th.

A doubling o f t h e economic U308 supply t o 6 m i l l i o n tons allows many more nuclear system options t o meet t h e p r o j e c t e d nuclear energy demand. I n fact, o n l y t h e throwaway optSon has cases t h a t d o n ' t even come c l o s e t o s a t i s f y i n g t h e demand. None o f t h e Option 4 cases meet t h e demand e i t h e r ; however, Cases 4s and 4H a r e w i t h i n 16 GWe o f t h e demand.

A l l o t h e r systems have a t l e a s t one advanced converter o p t i o n t h a t b u i l d s t h e desired 1959 GWe o f energy. It should be emphasized t h a t f o r t h e systems where t h e demand was met

i

w i t h t h e high-cost U308 (i.e.,'the

systems w i t h FBRs), a doubling o f t h e ore supply means

t h a t t h e o r e supply i s no longer t h e s o l e c o n s t r a i n t and p l a n t s e l e c t i o n i s based on economics. *NOTE: Since t h i s i s a 50-year span, some o f the r e a c t o r s b u i l t i n t h e f i r s t few years w i l l have been decommissioned a f t e r having operated 30 years.


c-4

Table C-3. Advanced Converter Option

1E*

LWR's (L)

5236 3.08

Utilization o f U308 Ore and Enrichment Through Year 2050 w i t h Various Nuclear Policy Options

f'

&

UsOa U t i l i z a t i o n (tons U30a/GWe)/Enricllment U t C Z i a a t i a (milZion SWU/GWel

HWR's

(HI

1

4

5T

6

7

a

2480

2908

1512

1514

1525

2.12

2.08

1.03

1.03

1.17

1487

1487

1528

0.85

0.85

1.01

1345

1314

1520

0.86

0. 84

1.00

2974

1503

1496

1666

2.10

1.02

1.01

1.20

2

3

5042

3138

1497

3165

3.08

2.03

0.82

2.70

4931

2864

1492

2793

2098

2340

2.83

1.76

0.87

2.38

1.78

1.58

4489

3027

1391

2243

1707

1983

2.18

1.37

0.88

1.78

1.33

0.80

4963

2105

1505

3497

2807

3.10

1.71

1.15

2.75

2.22

5u

L

c

Intermediate-Cost U30R Supply

LHR's (L)

5236

4973

3188

2758

3103

2957

3037

2733

2733

2798

2.85

2.82

1.76

1.45

2.46

1.86

1.77

1.68

1.68

3.61

4657

2820

2711

2844

2511

2511

2511

2511

2.43

1.38

1.27

2.03

1.34

1.34

1.34

1.34

2511 1.84

2431

2475

2195

1392

1924

1.56

1.58

1.32

0.88

1.23

HTGR'S (6) *System with standard

3916

2894

1398

1.40

1.22

1.00

3030 2.10

4478

2683

2680

3172

2865

3055

2683

2682

2698

2.88

1.61

1.60

2.21

1.77

1.77

1.58

1.58

1.82

I

LHR only.

Uranium and enrichment utilization f o r the various cases are shown i n Table C-3. The uranium utilization values are the total amount of uranium consumed plus the forward commitment per GWe o f nuclear power constructed through the year 2050. The enrichment utilization values are the total amount of separative work u n i t s required through the year 2050. As pointed out above, for the cases f o r which only 3 be available below $160/lb, the ore is the limiting factor. gives the savings i n ore on the throwaway cycle as a result exposure LWR less than 4% i n ore and none i n enrichment.

million ST U308 was assumed t o Comparing Case 1LE w i t h Case 1L of introducing the extended Cases l L , l S , l H , and 16 -compare the relative ore and enrichment uti1 ization of the various advanced converter options on the throwaway cycle. The HWRs clearly offer the greatest savings i n both ore and enrichment. Compared w i t h LWRs, the HWRs reduce ore requirements by over 10%and SWU requirements by almost 30%. In contrast, the SSCRs only offer a 2% ore savings and an 8% enrichment savings. The HTGRs reduce the ore usage by l e s s than 2%, with about the same enrichment requirements. The impact on ore utilization of the SSCR, HWR, and HTGR advanced converters on the throwaway cycle i s less than might k expected. The reason f o r the minimal effect is because most of the 3 million ST of U308 has already been comnitted t o LWRs before enough advanced converters can be b u i l t t o have much influence. Allowing the recycle of fuel i n thermal reactors (Option 2 ) results i n significant savings i n ore compared t o the throwaway cycle almost 60% f o r the HTGRs and from 30 t o

--

40% f o r the other converters.

For this nuclear policy option and the high-cost U308 supply, the HTGR clearly has the best ore utilization, although the HWRs have better enrichment u t i 1i zation.

I: L


c-5

The i n t r o d u c t i o n of t h e c l a s s i c a l Pu-U/U FBR i n Option 3 r e s u l t s i n an a d d i t i o n a l ore and enrichment savings o f about a f a c t o r o f two from t h a t i n Option 2 except f o r t h e HTGRs. Note, however, t h a t i n Option 2 the HTGRs already had a low o r e and enrichment usage.

In

Option 3 a l l t h e advanced converter cases have about t h e same usage. Recycling uranium i n denatured reactors and throwing t h e p l u t o n um away (Option 4) requires enrichment about halfway between Options 1 and 2.

Compared w t h the c l a s s i c a l

r e c y c l e o f plutonium i n thermal reactors (Option 2), Option 4 consumes roughly t h e same q u a n t i t y o f uranium w i t h LWRs and SSCRs.

That i s , t h e increased worth o f 233U i n LWRs and

SSCRs i s n e a r l y balanced by throwing away t h e plutonium.

The requirements f o r HWRs, however,

are considerably reduced over those of Option 2 when 233U i s recycled compared t o r e c y c l i n g p l u tonium. The very low f i s s i l e requirements f o r t h e denatured 233U HWRs i s responsible f o r t h e more favorable U308 u t i l i z a t i o n i n Option 4 compared t o Option 2.

I n contrast, t h e HTGRs i n

Option 4 look much worse than i n Option 2. This i s because t h e HTGRs were already operating on t h e 233UITh c y c l e i n Option 2. However, i n Option 2 t h e uranium-fueled r e a c t o r s a l l use h i g h l y enriched fuel w h i l e i n Option 4 they use denatured f u e l . Options 5U and 5T a l l o w t h e r e c y c l e o f plutonium i n plutonium/thorium transmuters, t h e d i f f e r e n c e between t h e two being t h a t denatured 235U reactors are a v a i l a b l e i n 5U whereas they a r e n o t i n 5T. f o r 233U.

This forces t h e 5T system t o i n i t i a l l y r e l y on the Pu/Th-fueled reactors

Compared t o Option 4, Option 5U r e s u l t s i n 20 t o 25% savings i n ore usage and

Option 5T i n 10 t o 15% savings.

The HWRs are t h e most e f f i c i e n t advanced converters f o r

uranium and enrichment u t i l i z a t i o n f o r Options 5U and 5T. Option 6 introduces FBRs w i t h thorium blankets, a1though these FBRs have uranium as f e r t i l e m a t e r i a l i n t h e core. Comparing Option 6 w i t h Option 3 reveals t h a t both systems have approximately t h e same resource u t i l i z a t i o n . t h e denatured 233U FBR i s included.

Option 7 i s i d e n t i c a l t o Option 6 except

The impact o f t h i s r e a c t o r

on resource u t i l i z a t i o n f o r

these cases i s small. I n Option 8 t h e Pu-U-fueled FBRs o f Option 7 a r e replaced w i t h Pu-Th-fueled FBRs.

The

longer doubling time o f t h i s r e a c t o r type r e s u l t s i n somewhat increased uranium and enrichment

A key p o i n t f o r a l l o f t h e systems c o n t a i n i n g FBRs (Options 3, 6, 7, and 8) i s t h a t t h e o r e and enrichment usage i s r e l a t i v e l y independent o f the advanced converter option. This i s i n c o n t r a s t t o t h e nonbreeder systems where the type o f advanced converter a v a i l a b l e (LWR, SSCR, HWR, o r HTGR) much more s t r o n g l y a f f e c t s t h e resource u t i l i z a t i o n . requirements.

L

Another very important p o i n t t h a t needs emphasis i s t h a t t h e s u p e r i o r ore u t i l i z a t i o n o f t h e HWRs r e l a t i v e t o t h e o t h e r advanced converters f o r the a l t e r n a t e f u e l e d systems (Options 4

- 8)

i s d i r e c t l y dependent on the denatured 233U-fueled HWR.

O f a l l the reactor

designs, t h e design of a l t e r n a t e f u e l e d HWRs have probably received t h e l e a s t amount o f analys i s and therefore have t h e l a r g e s t uncertainty. Thus, before i t can be concluded t h a t t h e HWRs o f f e r s i g n i f i c a n t resource savings, more work needs t o be performed t o v e r i f y t h e o p t i m i s t i c performance c h a r a c t e r i s t i c s o f t h e denatured 233U-fueled HWR.


C-6

Since 6 m i l l i o n ST o f U308 below $160/lb i s adequate, o r nearly adequate, t o s a t i s f y the projected nuclear energy demand f o r most cases i n the various nuclear options, the power growth patterns f o r these cases-are strongly influenced by economics a s - w e l l as resource u t i l i z a t i o n . Thus, as mentioned e a r l i e r i n t h i s appendix, the r e s u l t s f o r the cases based on the intermediate-cost U308 supply are subject t o much l a r g e r e r r o r s because o f l a r g e cost uncertainties. Table C-3 shows t h a t the advanced converters f o r the throwaway cycle r e f l e c t a l a r g e r U308 savings when 6 m i l l i o n ST i s used as a base r a t h e r than 3 m i l l i o n ST. This i s because many more nuclear plants are b u i l t w i t h the l a r g e r supply and therefore more advanced converters can be b u i l t , r e s u l t i n g i n a l a r g e r impact. For the high-cost U308 case, most o f the economic U308 was already committed t o the LWR before the advanced converters could have an e f f e c t . For Option 2, the r e s u l t s are about the same f o r both U308 supplies except f o r the case w i t h HTGRs (Case 26). Ore requirements per GWe are 27% higher f o r t h i s case w i t h t h e intermediate-cost U308 assumed t o be available. This i s because 6 m i l l i o n ST o f economic U308 i s an adequate amount o f ore f o r the system o f reactors i n Case 26 t o s a t i s f y the nuclear energy demand and economic considerations are also a f f e c t i n g t h e mix o f reactors t h a t are b u i l t . Thus, the f r a c t i o n o f low-enriched LWRs constructed i s l a r g e r because t h i s reactor i s less expensive than the HTGRs, even though the HTGRs use l e s s uranium.

i: L L L1

c

The p l a n t s e l e c t i o n f o r the cases t h a t include FBRs (Options 3, 6, 7, and 8) i s also determined by economics when 6 m i l l i o n ST o f U308 below $160/lb i s assumed t o be available. Therefore, the uranium u t i l i z a t i o n f o r these cases has l e s s meaning. Similarly, some o f the advanced converter options f o r the denatured cases (Options 4, 5U, and 5T) are resource l i m i t e d and some are not, so i t i s d i f f i c u l t t o draw conclusions regarding r e l a t i v e uranium and enrichment u t i 1 i z a t i o n . To summarize, there are two important and competing e f f e c t s when comparing t h e cases f o r the two uranium supplies: (1) For systems t h a t f a l l f a r short o f meeting the demand w i t h the high-cost U308 supply, the l a r g e r supply allows the advanced converters t o have a greater impact and therefore b e t t e r ore u t i l i z a t i o n ; and (2) systems t h a t have almost enough ore w i t h the high-cost U308 supply have p l e n t y o f ore w i t h the intermediate-cost supply, and therefore p l a n t selection w i t h the l a r g e r supply i s based on cost and ore u t i l i z a t i o n i s lower. The maximum annual U308 requirements and the maximum annual enrichment requirements through the year 2050 are shown i n Table C-4, The number i n parentheses next t o each maximum indicates the year the maximum occurs. As was mentioned above, i t has been estimated t h a t the maximum domestic mining and m i l l i n g r a t e may be approximately 60,000 ST/yr. Table C-4

L

L I]

indicates t h a t i f the high-cost U308 supply i s assumed, t h e annual U308 requirements vary from 50,000 ST/yr (Case 7s) t o 80,000 ST/yr (Case 4L). For most o f the cases, the maximum occurs during the f i r s t decade o f tile next century. Thus, most o f the cases require annual ore usage w i t h i n the next 25

- 30 years

t h a t exceeds the 60,00O/yr c r i t e r i o n .

L


c-7

Table C-4.

i ' ibd

Advanced Converter option

Maximum Annual U308 and Enrichment Requirements Through Year 2050 f o r Various Nuclear P o l i c y Options U308 Requirements (thousands tons/yr)/Enrichment Requirements (million swv/yr)

1E*

1

2

1

5u

5T

6

75(2009)

65(2011)

62(2009)

60(2009) 68(2005)

SS(2011)

45(2011)

44120091

42(2009) 55(2005)

50(2005) 35120051

SO(2005) SS(2009) 35(2005) 3812009)

7

8

.High-Cost U308 Supply

b i

4

3

80(2005) 69(20091

LWR's

73(2007)

72(2007)

67(2009)

60(2009)

(L)

44(2007)

45(2007)

46(2009)

41 (2009)

-

72(2007) 42(2007)

62(2011)

SZ(2009)

79(2009)

69(2011)

58(2017)

4012011)

3412009)

68(2009)

60(2011)

39(2010)

-

L

SSCR's

L

HWR's (HI

-

68(2009)

58(2011)

66(2009)

71(2009)

SS(2003)

53(2019)

64(2009)

63(2009) 65(2009)

36(20051

36120031

46(20091

58(2011)

4612023)

35(2003)

4612009)

4412009) 4612009)

HTGR' s (GI

-

72(2007)

57(2019)

53(2003)

65(2009)

57(2011)

64(2011)

61(2009)

60(2009) 65(2009)

45120091

51(20191

39(2005)

5212011)

4912017)

45(20111

44(20091

42120091 46(20091

86(2033)

86(2033) 92(2043)

61 (2033)

61 (2033) 6512043)

-

6)

i

Intermediate-Cost U308 Supply LWR's (L) SSCR's (S)

124(2025) 7412025)

-

HWR's (HI HTGR's (GI

L I

-

120(2025) 7712025)

114(2027)

llO(2039)

92(2037) lOS(2037)

72(2039)

60(20371 lOO(20371

llS(2039) 109(2039) 9012039)

77(2039)

96(2043)

93(2047)

82(2049)

83(2049)

83(2049)

83(2049)

83(2049) 83(2049)

6312029)

57(2045)

5312047)

7312039)

5512049)

55(20491

5512049)

55120491 5512049)

98(2031)

81(2023)

66(2009) 117(2031)

89(2029)

90(2029)

66(2009)

66(2009) 66(2009)

42(2009)

5312011)

47(20091 960033)

64(20291

64120311

47(2009)

47(20091 4612009)

llO(2029)

86(2049)

86(2049)

94(2043) 108(2041)

87(2047)

87(2047) 87(2047)

84(2029)

70(2049)

7012049) 90120391

7412047)

74l.20471 7512047)

96(2039)

86(20471

76(2041)

*System with standard LWR only.

The maximum annual separative work requirements based on t h e high-cost U308 supply This means t h a t t h e c u r r e n t separa-

Li

v a r i e s from 34 m i l l i o n SWU/yr t o 69 m i l l i o n SWU/yr.

.. .

t i o n s c a p a c i t y would have t o be doubled o r quadrupled t o meet t h e demand.

b

year when t h e U308 demand i s greatest.

As expected,

t h e year i n which t h e maximum separative work capacity occurs i s n e a r l y t h e same as t h e

Assuming t h e intermediate-cost U308 supply, t h e maximum annual ore requirements are g r e a t e r than 60,000 ST f o r a l l cases. For most o f t h e options, t h e year t h e maximum occurs i s 40 yr l a t e r than f o r t h e high-cost cases. This i s because, w i t h 6 m i l l i o n ST o f economic

U308, t h e nuclear i n d u s t r y continues t o expand. The breeder r e a c t o r systems t h a t i n c l u d e HWRs (Cases 3H, 6H, 7H, and 8H) a r e t h e o n l y cases t h a t have ore requirements t h a t a r e c l o s e t o being as low as 60,000 ST/yr. The maximum separative work requirements are a l s o very h i g h f o r t h i s uranium supply from 42 t o 100 m i l l i o n SWU/yr.

--

b

Table C-5 shows t h e energy support r a t i o s c a l c u l a t e d i n t h i s study f o r t h e year 2025, the energy support r a t i o being t h e r a t i o o f i n s t a l l e d nuclear capacity outside the energy centers t o t h e i n s t a l l e d nuclear capacity i n s i d e the centers. A l l t h e r e a c t o r types t h a t are a v a i l a b l e i n Options 1 and 4 could be constructed outside the centers; therefore, the energy support r a t i o f o r each case i n these options i s a. However, i t has already been shown t h a t these systems o f f e r t h e lowest uranium u t i l i z a t i o n and t h e r e f o r e t h e lowest nuclear growth p o t e n t i a l , even i f i t i s assumed t h a t 6 m i l l i o n ST o f U308 i s a v a i l a b l e a t below $160/lb.


c

C-8

Energy Support Ratios i n Year 2050 f o r Various Nuclear P o l i c y Options Table C-5. (Support Ratio = I n s t a l l e d Nuclear Capacity Outside Ener y Center/Installed Nuclear Capacity Inside Energy Center

B

Advanced Converter Ootion

Support Ratlo 1 E*

1

2

3

5u

4

High-Cost

U308

5T

6

7

8

Supply

m

1.54

0.72

m

5.69

3.74

1.27

1.46

3.09

m

1.47

0.76

m

6.33

3.86

2.13

2.13

3.27

m

0.49

0.92

m

5.79

3.07

1.07

1.06

2.89

m

0.24

0.24

*

4.02

2.50

1.26

1.28

3.11

L L

L

Intermediate-Cost U308 Supply m

2.42

1.65

m

5.06

5.05

5.37

5.37

5.49

m

2.10

1.65

m

4.78

4.78

4.78

4.78

4.78

m

1.85

0.94

m

4.03

3.84

1.03

1.04

3.07

m

1.77

1.82

m

3.30

3.20

2.74

2.74

3.62

%

L

*System with standard LWR only.

As pointed out previously, w i t h only 3 m i l l i o n ST o f U308 a v a i l a b l e below $160/lb, the only systems t h a t s a t i s f y the energy demand o f 350 GWe i n t h e year 2000 and 15 GWe/yr t h e r e a f t e r are those w i t h breeders. The disadvantage o f the c l a s s i c a l Pu-U breeder cycle (Option 3), o f course, i s the low energy support r a t i o since the pl'utonium t h a t i s produced must be used i n the energy centers. One technique f o r increasing the energy support r a t i o i s t o load thorium i n the blanket o f these breeders, while r e t a i n i n g plutonium and uranium i n the cores. The 233U t h a t i s produced i n the blankets i s then burned i n denatured LWRs located outside the centers (Option 6). The r e s u l t i n g energy support r a t i o s f o r Option 6 vary from 1 t o 2, depending upon the advanced converter option. Option 7 introduces a denatured FBR which would provide 233U t o the system and therefore should increase i t s nuclear growth p o t e n t i a l . However, since Option 6 can meet the projected nuclear growth demand i t s e l f , the a d d i t i o n o f the denatured breeder i n Option 7 a c t u a l l y had a minimal impact. The energy support r a t i o s o f Options 6 and 7 could be f u r t h e r increased by replacing the uranium i n the core o f the Pu-U breeder w i t h thorium (Option 8). With the high-cost U308 supply, energy support r a t i o s o f about 3 are obtained f o r t h i s system. The i n t r o duction o f thorium i n the core o f a breeder lowers the breeding r a t i o t o the p o i n t that, i n contrast t o Option 7, s i g n i f i c a n t q u a n t i t i e s o f FBRs operating on denatured fuel must be b u i l t t o meet the projected nuclear growth demand.

c c G

c

c I'

L

c


c-9

I n general, t h e energy support r a t i o trends f o r t h e various options a r e t h e same i f 6 m i l l i o n tons o f U308 i s a v a i l a b l e below $160/lb; however, they a r e s i g n i f i c a n t l y higher, l a r g e l y because more low-enriched LWRs can be b u i l t . E

,

1 '

b _ . I

'

L 1

b

_-

i

G i'

b

Selected d e t a i l e d r e s u l t s f o r a l l t h e cases c a l c u l a t e d are presented i n Table C-6, C-7, and C-8. While many o f t h e numbers i n these t a b l e s appear elsewhere i n t h i s report, many numbers a r e a l s o shown f o r t h e f i r s t time. For example, t h e p l a n t mix i n year 2025 and t h e l e v e l i z e d power cost f o r each p l a n t s t a r t i n g up i n t h e year 2025 are shown. The purpose o f these t a b l e s i s t o group a l l t h e data together and a l s o t o provide s u f f i c i e n t data t o h e l p e x p l a i n t h e behavior o f t h e various r e a c t o r systems. (Note: Cases 1LT and 1LTM i n Table C-6 are f o r changing enrichment compositions; see Section 6.2-1

i n Chapter 6.)


c-10

Table C-6. Summary o f Results f o r Cases Assuming High-Cost U308 Supply, 350 GWe Installed Capacity i n Year 2000, and 15 GWe Installed Capacity Each Subsequent Year 11

a

a

579

I84 OU

1029

651 945

1005

916

1959

1205

I027

1959

440

47s 221 135

480 220

U9

594

&

n-nsz 1029 1959

1029 1947

509

207

110 222

PI

u( 229

IO29

c c

wtem Costa ($8) 1977 thm& 2050 dilcMted

8t

4.5% 7.5% 10.0%

L a v e l i d !System R*rer b t a (UllJIWlr) in

m

mi5 2025 2035 Oulative U 0 Cmsmptlm million t m h 8 t h m g h 2025 2049

159 165

362 11

209

119

119

129

507 220 1S2

16.2 20.1 20.9 21.6

16.0 19.6 20.1 21.1

16.1 17.7 16.4

1t.2 20.0 20.6 21.0

16.2

18.6

15.5 16.1 17.2 17.8

19.0 19.6

15.7 l7.S 16.1 19.0

15.7 16.0 16.6 M.0

15.7 16.2 16.6 17.9

16.0 17.0 16.1 20.0

2.57 2.97

2.55 2.96

2.S 2.95

2.14 2.75

2.6S 2.S

2.50 2.W

2.W 2.94

2.18 2.62

2.14 2.6s

2.29 2.M

2.1 2.99

2.92 2.99

2.65 2.99

2.49 2.93

2.90 2.99

2.66 2.99

2.6.7 2.99 45

2.49 2.96

2.54 2.97 42

2-59 2.97

44

133

18.0

128

IS2

1S2

1s

Total U 0 Eolrmitted Btillim

Tons) d & h 2025 2049

kina knual EnrWnent Pawin-

mt t h w h 2050 (Milllm SiU/Yr) Ormilath Fnridnent (sillim 9u)

th%25

2049

U ~ Utiliratim O ~ oms u 2025 2049

hri-t

~ o , / ~inn) )

1.U 1.76

1.58

1.60 1.95

1.47 1.79

2.20 2.15

2.08 2.S

1.61 2.11

1.U 2.01

1.51 2.02

1.62

1.63

52%

SMS

2420 1497

242s 1512

2469

25lS

a65

2867 2480

SO86

5042

S22t SlP

uo1

52%

1114

1525

2.74 Loll

2.72 3.1

1.61 2.03

1.4s

2.0 2.70

2.07 2.12

1.75

1.49 1.03

1.46

1.76 1.17

2908

2.29

h3

u t i l i u t i m ( r ~ ~ i sarny c w ~ ) ( ~ )

in 2025 2049

llmlative W l e a r W i t y hilt (-1 thrmgh 2025 2049

System G n t a

(SB) 1977

.92

2.06

1.O.l

1029 1959

1029 1943

500

51s 222

591 bo7

1043

94b

1029 1959

944 1071

1029 142.3

1029 I275

1029 1959

169

451

502

495

490

I8H

211 129

219 131

226

221

470 213

218

I20

116

133

129

IS1

I18 IS1

18.1 19.7 20.4 21.0

16.0 17.1 17.6 17.9

15.4 15.9 16.6 17.0

16.0

19.5 20.1 20.5

15.9 17.2 18.1 19.0

15.5 16.6 17.4 18.2

15.4 15.9 15.9 14.4

15.4 15.9 15.9 14.4

15.6 16.S 17.0 17.6

2.54 2.96

2.27 2.92

1.99 2.70

2.62 2.98

2.S5 2.%

2.14 2.91

1.93

1.93 2.69

2.07 2.83

2.92 2.99

2.81

2.43 2.92

2.89

2.81

2.99

2.199

2.77 2.98

2.36 2.91

2.%

2.99

2.58 2.97

40

34 (2009)

6a

35

(2009)

bo (2011)

s9

(2flll)

(2010)

(2005)

1.47 1.84

1.32 1.70

2.19 2.54

1.94 2.54

1.45 2.02

1.33 1.86

.

thm@

2050 d i v a n t c d at

Levelired Spta hucr Costs pIills/Kvdir) in -WO

2015 2025 2035

Torn1 II 0 Committed Fcillion Tons) l h 8 q h 20%

2049

42 .

(2Ilrn)(3)

thrcugh 2025 2049

1.48 1.72 4939 4931

2.50 2.83

*System with standard LWR only.

2975

2864

1.55 1.76

2362 1492

1.28

.87

5066

2793

2.32 2.P

nso 2091)

1.66 1.79

26117

2340

1.41 1.59

500

2.69

2297

1487

1.29 .%

2.91

IS2

35 YI (ZIIIIS) 12mw

1.35 1.66 2297 1487

1.29 .95

1.42 1.97 2506 1528

l.P 1.01

.

f1 â&#x20AC;&#x2DC;


c-11

b-,

Table C-6 (cont.)

ii

-latin k l c a r Capacity hilt (Ore) thmugh 2025 2049

I ;

System Costs ($I)) 1977 through 2050 discanted a t 4.5% 7.5%

b

10.01

i-

Levelired -tea bier Costs W I l d M r ) in 2000 2015 2025 2015

Total II 0 b i t t e d (Uillion Tons) d & h 2025 2049 Cbxirn kvual Enrichcnt Rguiremt thmugh 2050 (Million WJ/yr)

641 667

1029 1959

lo03

907

1134

1747

1029 1505

Y7 192 121

494 222

524 225

134

134

551 214 137

566 2% 158

549 a2 1%

134

225 113

227 I34

17.9 20.3 21.3 21.6

17.6 20.0 20.8 21.1

17.2 15.7

17.9 20.8 22.4 21.1

16.1 19.1 20.7 21.7

15.6 16.9 20.7 22.4

16.9 17.6 17.6 17.3

16.7 17.1 17.1 17.6

15.6 17.1 18.5 20.6

2.47 2.95

2.24 2.91

2.29 2.70

2.44 2.97

2.16 2.92

2.14 2.90

2.25 2.61

2.21 2.55

2.29 2.67

2.90 2.99

2.81 2.99

2.63 2.72

2.90 2.99

2.70 2.98

2.79 2.90

2.55 2.63

2.50 2.57

2.69 2.98

16 (2001)

4b (2009)

58 (2011)

46 (2021)

35 (2001)

46 (2009)

44 (2009)

(2009)

1.24 1.35

1.61 1.94

1.90 2.37

1.61 2.31

1.2s

1.45

1.35

1.58 1.8.3

1.55 1.84

1.61 1.95

4524 4489

son

3095

25% 1391

2890 2241

2620 17W

2707 IS81

2482 1345

2426 1114

1520

2.02 2.18

1.17 1.37

1.57 .99

1.76

1.90

1.58 1.33

1.19 .W

1.54

1.51 .91

1.00

908

(ZOO$)

17.4 10.0

~

1029

1029 1959

I029 1959

1029

1959

us

52s

529 224

46

(Lndstive E n r i c h r ~ ~millim t SW)

ii

1.10

th?!25 2049 u308 Utilization

2025 2049

oms U ~ O ~ C iWn’) ~ )

thriciment u t i l i z a t i m w i l l i o n S W / W ) ( ~ ) in 2025 2049

.%

1.57

L

b Cutlstlw Hrlear Capacity h i l t (Ore) thmigh 2025 2049

4

568

60s

System Costs ($6) 1977 thmugh 2050 discomted a t

36.9 I68

120

i

Levelired Systea b i e r Costs (\(ills/Mr) i n

-.

439 209

I29

917 iool

451 210

209

1029 19y1

1029 1959

1029

1791

678 703

705

505 220 132

516 224 15O

187 197 124

417

221 112

15.6 16.0

15.3 16.6

16.1

16.8

17.4 19.1 19.7

17.6 19.0 19.6

16.4

20.6

20.3

20.2

442

I29

129

15.9 17.1 17.7 18.1

15.6 17.3 16.6

15.6 16.0 16.1 16.4

xu

201)

IS2

16.0 16.4 16.6

15.7 16.0

20.5

21.1

,17.3

14.2

17.7 18.6 16.9 18.9

2.55 2.%

2.19 2.92

1.97 2.75

2.35 2.91

2.n 2.92

2.31 2.91

2.15 2.70

2.12. 2.60

2.32 2.91

2.41 2.95

2.35 2.94

2.92 2.99

2.n 2.98

1.41 2.95

2.99

2.85

Z.M 2.99

2.63 2.99

2.42 2.93

2 3 2.93

2.77 2.90

2.89 2.99

2.67 2.99

52 (2011)

49

(ZOOS)

(2017)

45 (2011)

44 (2009)

(20091

46 (2009)

92 (2011)

95 (2011)

1.60

15.6

39

(20&

16.1

42

mt t h m g h 2050 (UillimWJ/yr)

(20%’

Qrrulative Enrichmt (Billim SWI) thmgh 2025 2049

1.59 1.67

1.71 2.42

1.49 2.35

2.B

1.69 2.36

1.62 2.11

1.53 2.00

1.50 1.98

1.64 2.15

2.69 3.42

3.25 4.06

4973 4%3

no5

2700

2502 1505

3557 3497

2920 2807

ma2 2914

us2

1503

2316 14%

2 ~ 2 4268 1666 4258

4078 4074

b r i c t n c n t ~ t i ~ ~ z a t iwillim on SU/W)(~) in 2.70 2025 3.10 2049

1.66 1.71

1.45 1.15

2.24 2.75

1.76 2.22

1.77 2.10

1.48 1.02

1.45 1.01

1.60 1.20

3.97 4.86

4.60 5.53

2025 2049

(2) (3)

502 219 131

958 io64

16.2 19.9

U ~ Utilization O ~

(1)

484 217 131

155

2015

Fhrinm krual Enrichcnt wire-

i‘ W

803

202s

Total II 0 C a a i t t e d (Uillim Tons) ti&h 2025 2049

L

1029 1959

ZWO

2015

L

1029 1417

oms U ~ O ~ Win(’))

Cumulative U,Os consumed through year 2050 (including forward comnitments) per cumulative nuclear capacity b u i l t through 2050. Cumulative enrichment requirements through 2050 per cumulative nuclear capacity b u i l t through 2050. Year i n which maximum enrichment requirements occur.


c-12

Table C-6 (cont.)

Reactor

t#R-US(tE)h)

W-US(LE)/U-EE

1E*. -

&

269/22.3

30/22.3 259/21.4

-

Installed Cnpacity (lXe)/Levelized Power Cost (Mills/Kuhr) in year 2025 2L 3L 6L 4L SIZ 7L

-

-

0L - - - - - -

360h9.6

310/10.0

52/21.5

49/19.7

412/19.0

2W19.0

72/19.3

292/23.0 220120.4

2%/21.4 264120.0

02/20.?

0/19.0 0~17.6

io?/ia.?

132119.6

-

-

--

-

-

357/20.6

-

327/1?.5

342/10.0

110/17.9

aona.4

0/19.5

107/10.0 9120.7

9120.9

21124.1

9/2b.o

316/19.0

200ll0.0

-

W19.5

2s ss 4s sus - 1s

101/22.2

SIS

03h9.7

03/10.0

-

266h7.0

123h7.3

-

200/21.0

307/10.6

237h7.2

0/20.7 300/20.5

-

FBR-h-Ufll FBR-h-U/lh FBR-h-ThDll FBR-u3-u/Th

0/24.9 222/22.0

-

8s

4/1?.9 303/19.? 101/19.0

3?2/1?.6 135/19.9 152/19.1

257/16.4 l66/15.5 40/14.9

257/16.4 310/16.6 166/lS.S 42/1?.2 4W14.9 25122.9

-

iami.7

188/11.7

-

00h7.4 0/10.L

00h7.4 0/10.2

-

355/19.9

151/21.3

15?/10.0

O/25.6 0/22.0

0/27.0 0/23.1 222/24.2 339/24.O

0/22.0 0/20.3 170/22.0 2%/22.4 109/21.?

0/26.0 45/22.9

-

415/21.1

0/20.S 384h4.6

-

217/21.4 20/20.0

323/10.7

2G

E -

142/19.4

142n7.5

1?2/21.2

163/24.3

-

0/20.0 32/19.? 12/21.6 O/22.9

102/22.7

2/20.9

11/20.0

0/26.2

-

0/19.8

0/10.5

0/20.4

I -

-

305h7.2

-

195/15.8

17G17.9 12G14.2 117/15.0 79D6.5

-

195/11.4

4G SIK; -

1Lrn

W21.5 358/20.?

30/21.5 305/20.7

0/19.9

29V16.9

7G

0G -

295/1?. 2 347/17.0 0/10.4

0/19.4

l4A7.4

V17.4

so/10.1

4W9.0

104/15.4

91/16.0

0/20.2

133n0.3 179110.5

15/22.3

30/21.0

0/27.5

313117.a

294/17. S

2W19.0 305/10.5

-

56ll0.5

-

6G -

-

190/22.3 176/21.1

0/10.4

87/10.l

.

1LT

STG -

142h9.1 404/19.2 0/19.0

329/10.3

0/23.3 0/20.7 Ol26.5 0/19.4

57/1?.2 1G -

-

150/19.0 126/19.3

0/23.9 0/21.1 0/20.9 45h7.3

-

172/22.3

?9/1?.9 l/l0.0

arr m BH 150/10.4 337/19.0

3W17.4 340117.4

125/20.7

*System with standard LWR only.

6s

00/10.0

3H 4H SUI -

-

FBR-h-U/IJ FBR-h-U/Th FBR-h-nJTh FBR-U3-U/Th

UV-R-US(LE)/U

-

?S - -

45119.2 207/19.6

0/1?.2

m - -

tIwR-h/Ill

-

297h7.0

12W22.1 158/21.1

Hrn-h/U

-

49/21.5 209/22.3

-

1?2/21.7 24V21.7

-

29/15.9

I]

180j25.7 162123.8

I’

c


C-13

Table C-7. Summary o f Results f o r Cases Assuming High-Cost U308 Supply, 200 GWe I n s t a l l e d Capacity i n Year 2000, and 10 GWe I n s t a l l e d Capacity Each Subsequent Year _ -

I

ii

2U25

2049

I

E

1.w

1IC

lIX

533

554

570

(100

579 63.9

619 727

589 654

269

269

279

128

128 60

130

302 135 83

281 131

16.5 19.3 20.5

16.5 18.5 19.3

IlK* &latire Hrlenr Capacity hilt We) thrwgh

S y s t a Casts (IS)

zusn discourtal a t

1977 through

4.5%

7.5%

81

10.0%

81

81

16.8 19.2 20.1

16.5

16.5

18.6

18.5

19.5

19.4

2.08

2.90

2.02 2.89

2.87

1.88 2.82

1.94 2.86

2.79 2.98

2.76 2.90

2.72 2.90

2.62 2.97

2.71 2.98

__

I

L

1.94

Total II 0 C m m i l t d (Million lams) t & g h 2025 _ . . .

2049 bhxinm kuual rnrichnent Rcquiremt thnnNh 2050 (Million .pnvYr)

I td

&latIve Fnriclmcnt (Billion SHI) tllrlmgh 2025 2019

uo

Utilizatini

2025

mons

u ~ o ~ in(') / ~ )

2049

C i

b

F n r i c k n t Utilization (Hillion ill

41 ( 2 0 E ~ (2021)

35

24

45

(2021)

(2011)

(2n23)

1.23

1.26

1.73

1.81

1.11 1.62

.94 1.20

1.99

5236 5236

4979 4974

694 4669

4222

4603

4090

4554

2.31 3.03

2-20

1.Yl 2.Y

1.52 1.66

2.18

1.211

.SUJfcWc)(2)

2025 2049

5-92

3.01

Installed Capacity (cHe)/Levelized Power Cost @Iills/Mr) in Year 2025

L

i,

Reactor

LWR-U5(LE)/U LWR-US(LE)/U-EE M-US@E)/U/lh m - U 3 @E)/u/lh

lW*

1Lc

363/21.7

11/21.6 374/20.0

1SC 44/21.4

1Hc

la:

144/21.2

114/21.4

m-pu/u LWWU/Th

L

=:ux'/u/Th

BJ

Hm-us(sar)/u Mn-US@E)/U/lh H m 4 3 @E)/u/Th WIm-PuhJ

S c R - U S (MI hJ

c

Mm-us(NAT)/U

36u20.4

0/24.2 305/2l. 5

tun-hl/lh

304/20.1

L

1 ) Cumulative UsOB consumed through year 2050 (including forward comnitments) per cumulative nuclear capacity b u i l t through 2050. 2 ) Cumulative enrichment requirerents through 2050 per cumulative nuclear capacity b u i l t through 2050. 3) Year i n which maximum enrichment requirements occur.

*System with standard

LWR

only.


-.

C-14

Table C-8. Summary o f Results for Cases Assuming Intermediate-Cost U308 Supply, 350 GWe Installed Capacity in Year 2000, and 15 GWe Installed Capacity Each Subsequent Year 2L -

1L -

4L -

1029

I029 1959

1029 18U

1783

System Costs

Sn

m -

u

n -

U. -

1029 1959

lo29 1959

1029 19%

486

1029 1921

I029

409 214 129

485 2x3 129

485

415

I29

129

14.7 15.1 15.7 15.8

14.7.

14.7 15.2 15.4 14.1

14.7

1864

($8) 1977 thmugh

2050 dixorntal

at

542 231

473 212

470 211

485

129

485 213 129

16.6 18.5 19.5

15.0 15.5 16.1 16.5

14.8 15.0 15.3 14.9

16.6 17.6

20.1

16.4 17.9 18.7 19.3

18.2

14.8 15.5 16.0 16.1

2025

1.53

2.0 4.81

4.70

2.17 4.38

2.17

4 -40

2.87 5.11

2.%

2049

2.B 4.16

2.28

5.61

3.41

4.P

2.37 4.48

5.06

5.91

1.50 5.68

3.28 5.40

1.66 5.74

1.17 5.68

1.37 5.66

3.39

2049

5.20 5.94

1.39 5.35

1.40 5.47

n

72

4.51 7.5: 10.0:

la

l m l i r e d System lbwcr bt. F(ills//Mr) in 2Wo

mi5 2025 2035

Qmrlative U 0 Conarptim (Viilim Tci~h~thmugh

k i m a AMP1 Buicbent kpllim.ent thmph 2050 (Millim

74 (2025jJ)

In

5.56

n4

(2025)

IS7

18.0

77

ns

15.2 15.4

14.1

5.1s 61

n3

t:

213 129

15.2

15.5 15.0

c L;

4 65 (20s) (2045)

(ma)

100

90

(2037)

(20m

(mw

(20JJ)

1.I7 2.84

2.41 4.55

1.64 1.51

1.64

1.64 3.09

1.64

3.21

3552 SIOS

1272

3270 3037

3296

2%?

27SS

nu

2798

1.59 1.1

1.59 1.1

1.59 1.61

60

L

OrnilatIra h r i h t (Billim SU) -25 2049

2025 2049

(Lulative ICulear -icy

J(!wth-@

1.64

2.09 1.35

2.12 3.49

1.56

a%

4 s 4973

3386

3l82

Su6

SI88

I1y

2.11 2.95

2.09 2.92

1.51 1.75

1.43 1.45

2.34

2.46

1.59 1.86

1.59 1.77

IS -

2s -

P -

rs

PIS -

m -

1029

1029 1917

I029

1029 1941

I029 1959

1029 1959

1029

1029

1959

1029 1959

481 212 128

5% 210 1%

485

485 214 129

485 214

485 214

485

214

129

214 129

1.12

3.09

3296

1.16

110s

as -

Lilt

2025 2049

1271

1959

1959

S y s t a Costs ($8) 1977 t h w h

2050 d i x a n t a l at

lavelid System Rmcr Costs (Nilis/*rhr) in 2OOo 2015

2025 2035

129

129

16.4 17.9 18.7 19.2

14.9 15.2 15.5 15.6

14.7 14.7 14.9 14.6

16.1 17.0 17.1 17.0

14.7 14.9 15.2 15.1

14.7 14.9 15.2

1s.s

14.7 14.9 15.2 15.3

14.7 14.9 15.7. 15.1

14.7 14.9 15.2 15.1

3.26

2.21

2.14

4.w

2.20 4.14

2.m

5.46

4.61

3.86

2.14 1.86

2.14 1.86

2.14 1.86

2.14 1.86

4.85 5.92

3.18 5.46

5.10

3.% 5.52

2.90

5.n

4.92

2.94 4.92

2.94 4.92

2.91 4.92

2.91 4.92

(2047)

73 (2039)

(2049)

55 (2049)

(2049)

55 (2049)

55 (2049)

I I

~

I

th%Z5 Z049

53

*System w i t h standard LWR only.

55

1.06

3.09

1.40 2.63

1.a 2.49

2.26 1.94

1.46 2.62

1.46 2.62

1.46 2.62

1.46 2.62

1.46 2.62

4714

sow

1010

3262

2858 2511

205a

2511

2858 2511

2858

a44

as

2711

2511

2511

I.%

1.Y

2.19 2.0s

1.42 1.54

1.42 1.34

1.42

1.42 1.54

1.42 1.34

4657

2025 2049

55

1.81 2.41

am 1.36

1.27

1.34

L L I;:


C-15

Table C-8 (cont.)

Qulatiw k l u r W i t y hilt (*I thrmgh 2025 2049

IH -

m -

a

1029 1497

1029 1921

I029 1959

1029 1945

1029 1959

I029 1959

1029 I959

1029 1959

1029 1959

519 221 I30

544 228

512 222

552 229

523 224 153

523 224 153

514 222 132

512 222 152

514 222 132

16.3 18.5 19.6 20.1

16.3 17.4 18.3

18.8 19.5

16.0 16.7 17.0 17.1

16.0 16.7 17.0 17.2

15.7 16.0 16.0 15.9

15.7 16.0 15.8 15.5

15.7 15.9 15.9 15.5

5

n

i

g

7

H

w

(a)

Sptm Costs 1977 thrwgh 2050 discanted a t 4.58 7.58 10.0\

Lmlircd Systa k Wi(ills/Yrhr) i n

IS4

132

1S4

r Colts

tOD0 2015 2025 2035

Qulative U 0 hsq.tim (r(illim Tm2)8thmugh 2025 2049

biu &uual b r i c h e n t bpircMY) Bullion W.l/y~)

Pnt th-

16.7

18.8

15.8 16.1 15.8 14.9

2.72 4.55

2.31 2.71

2.94 5.36

2.52 b.32

2.51 4.57

2.32 3.66

2.30

5.20

2.10

2.m 3.57

4.55 5.86

3.67 5.56

2.65 2.?4

4.n 5.89

3.59 4.76

3.57 4.85

2.71 4.30

2.64 2.73

2.85 3.77

64 (2051)

47

47

(2009)

(2009)

(2009)

5.10

42 55 (2009j5) (2011)

47

(2oos)

18.0

9 6 6 4 (20SS) (2029)

46

b i l a t i v e hrichrnt (Billim SUJ) M -' 25

2049

u308 Utilization mms u 3 0 p e ) in(') 2025 2049

1.57 2.10

1.80

2.34

1.62 1.95

2.05 4.011

1.75 5.06

1.74 5.09

1.65 2.59

1.62 1.94

1.67 2.40

4225 396

3562 2894

2572 1398

4093 3030

3490 2431

3470 2475

26%

2195

2562 1592

1924

1.75 1.22

1.58 1.m

1.99 2.10

1.70 1.56

1.69 1.58

1.58 1.52

1.57 .99

1.6.2

FIT.

sc;

n:

sr;

27-13

Ehrichmt Utiliutim (Nillion SNJ/~XO)(~)

in 1.52 1.40

2025 2049

1.25

Qulative k l c a r C I p r i t y hilt

(*'2;;;"ph

1029 1320

1029 1959

1029 1959

1029 1794

1029 1924

1029 1844

1029 1959

1029 1959

1029 1959

487 214 128

486 214 129

486 214 129

515 223

487 214 129

486 214

486 214 129

486

129

214 I29

486 214 129

16.4 17.9 18.6 19.0

15.0 14.9 14.8 14.2

15.0 15.0 15.0

16.2 16.6 16.7 lb.5

15.1 15.2 15.6 15.8

15.0 15.5 15.7 16.0

15.0 14.9 14.9 14.7

15.0 14.9 14.9 14.7

15.0 14.9 15.0 14.7

2025 2049

3.23 5.41

2.32 4.23

4.22

2.30

2.58 4.741

2.52 4.35

2.U 4.65

2.23 4.19

2.23 4.19

2.26 4.24

2049

4.73 5.91

5.20 5.26

3.17 5.25

5.57 5.69

S.20 5.51

3.29 5.64

5.26

5.26

3.09

3.15 5.29

70 (2049)

go (2039)

(16

76

(2047)

(20411

74 (2047)

74 (2047)

75 (2047)

2049 S y s t m Costs

(a)1977 thrmgh

2050 discanted at

Levelired Systa Rwr Costs Willslurhr) in 2000

2015 2025 2035 Qulatiw U 0 w

i

m

plillim Tcd)'rhgh

Maxhual hrnrichpnt wiremmt Ulmugh 2050 B u l l i m SU/yr) Qulatirs

2049

uSo8 Utiliratim rims u ~ o ~ w ? in(') ) 2049

(20%')

(24:)

3.09

thrichmt (Billim w )

-25

2025

14.8

133

2.11 3.81

1.62 5.16

1.60 3.13

1.99 5.97

1.64 5.41

1.65 3.27

1.55 5.10

1.55 3.10

1.57 5.16

4597 4478

3105 2683

3081

S472 5172

3108

2h30

2865

3198 3055

so05 2683

3004 2682

3057 2698

1.58 1.61

1.54 1.60

1.95 2.21

1.59 1.77

1.58 1.77

1.11 1.58

1.51 1.58

1.53 1.62

m r i h n t Utiliutim willion SNJ/~X~)(~)

in

2025 2049

(1) (2) (3)

2.05 2.89

Cumulative UBOs consumed through year 2050 (including forward comitments) per cumulative nuclear capacity b u i l t through 2050. Cumulative enrichment requirements through 2050 per cumulative nuclear capacity b u i l t through 2050. Year i n which maximum enrichment requirements occur.


C-16

Table C-8 (cont.) Installed Carurcitv GWeVLevelized Fwer Cost Mills/Kwhrl in year 202s

3L 703/20.6

-

30/20.S 69S/l9.8

-. 216h6.8

S23/17.0 460/1S.8

--

254/14.9 2m2.s

41 -

SUL -

57/18.8

S41A6.9

439/19.0 243/17,8

3/17.8 7V17.0

-

8L 7L 6L sn S44/16.3

SSl/lS.8 SSl/lS.8 553/16.0 0/16.3 72/13.2

0/16.3 72/13.2

0/16.8 73/14.2

122/16.2 122DS.8 103112.5

103D2.S

103/14.1

13/10.7

13/10.7

72/16.4

Om.5 4s 3s 1s - 2s

109/20.3

8V16.4

llS/lS.4

5us -

S7/17.9 184hS.6 380/17.6 0/17.3

m -

6s -

418D6.0

346/1S.O

8s 7s -

184/15.6 184/15.6 184D5.6 184h5.6 0/17.3 0/17.3 0/17.3

239/1S.7 279/14.4 630/19.5

-. 0/17.8 500/1S.S 302h7.1 128/1s.s 128/1S.S

FBR-FU-U/lh FBR-FU-RJlh FBR-u3-u/Th

0/16.3

1H

w ft -

4H -

mi -

6H 51)l -

0/16.3

0i18.4

0/19.8

8H ni 357/1S.8 410/15.7

-

-

-

FBR-h-U/IJ FBR-FU-U/Th

FBR-Pu-Th/Th

19/15.1

m-u3-u/Th

x 1C -

3G 4G -

201/20.3 472/14.6 477/14.9 0/16.1 Oll6.3 S39/19.2

-

28P4.S

-

-

14/15.1

SUG -

5TG -

193/17.9 40S/16.1 S18/16.2 0/17.7 0/17.1 0/17.2 471/16.7

109D6.2

76/15.9

WlS.7

-

63/13.4 63/14.S 176/15.3 17S/lS.7

-

lO/ll.l

-

4S/lS.6

*System w i t h standard

0/21.5 oiia.9 0/24.1 0/17.6 0/17.3

-

182/lS.8 148/15.4

7G - 464/1S.1 471/15.1 8G

6G

466hS.1 Oh6.3

0/16.3

W16.3

mS.3

7/1S.3

W15.4

71D5.0

7VlS.O

94/1S.O

172/16.0 176D6.2 148/16.S 50/12.8

148/16.S 132/16.6 SO/l2.7 0/16.l

,

-

0/16.3

232D9.S 480/17.8 359D6.1 666h9.5 592/17.4 S87h7.4 37V16.1 0/23.0 0/23.0 0121.7 0/22.9 0123.9 0/22.2 0/2S.4 0/21.7 0/20.1 0/20.1 o/ig.o 507/20.4 0/20.6 0/19.1 0/19.3 0/21.8 0/26.2 0/24.4 63/21.7 0/2s.1 0/17.4 10/20.4 0/17.3 0/17.5 0/17.S 0/18.8 259/19.6 147/17.2 1w17.3 0/17.0 0/18.8 380/14.6 364/15.9 363/15.2

-

c

c

i;

300/1S.S 500/1s.s Joo/1s.s 300/1S.S 128/15.S 128/1S.S 128/1S.S 128/1S.S 128/lS.S 128/1S.S 128DS.5 128/15.5

0/14.1

FBR-h-u/IJ

11/12.8 0/16.2

t:

-

28/12.6 0/16.6

u D

c f:

L

c

LUR only.

t:


D-1

Appendix D.

L

CALCULATIONS OF NUCLEAR AND FOSSIL PLANT COMPETITION BASED ON ECONOMICS

M. R. Shay, D. R. Haffner, W. E. Black, T. M. Helm, W. G. J o l l y , R. W. Hardie, and R. P. Omberg Hanford Engineering Development Laboratory

I n a s e r i e s o f c a l c u l a t i o n s t h a t preceded those reported i n Chapter 6 f o r nuclear power systems, the same a n a l y t i c a l model was used t o evaluate power systems t h a t i n c l u d e both nuclear power p l a n t s and c o a l - f i r e d power plants, w i t h t h e two types o f p l a n t s being c

i n economic competition.

As was s t a t e d i n Chapter 6, t h e r e s u l t s o f these c a l c u l a t i o n s

i n d i c a t e d t h a t a t U3O8 p r i c e s above $160/lb, nuclear power p l a n t s do n o t compete w e l l f o r

c

t h e assumptions used i n t h i s study. Therefore, f o r t h e a l l - n u c l e a r systems i t was decided t o l i m i t t h e uranium resources t o those a v a i l a b l e a t p r i c e s below $160/lb. This appendix describes the i n i t i a l s e t o f c a l c u l a t i o n s .

The nuclear p l a n t s used

were LWRs, w i t h and w i t h o u t recycle, and they correspond t o Cases l L , 2L,. t e r 6.

...8L

i n Chap-

The primary d i f f e r e n c e s between the c a l c u l a t i o n s presented i n Chapter 6 (and i n

Appendix C) and t h e c a l c u l a t i o n s described here are as f o l l o w s :

b -

(1) I n s t e a d o f a nuclear energy growth p r o j e c t i o n , a t o t a l e l e c t r i c a energy growth p r o j e c t i o n was used. (2) I n a d d i t i o n t o nuclear plants, coal p l a n t s were a v a i l a b l e t o s a t i s f y t h e t o t a l e l e c t r i c a l energy demand.

i

ii

L

No p r i c e c o n s t r a i n t on o r e existed. Instead i t was assumed t h a t a d d i t i o n a l uranium ore was always a v a i l a b l e a t i n c r e a s i n g l y higher costs. As w i t h t h e a l l - n u c l e a r systems, two d i f f e r e n t U308 p r i c e s t r u c t u r e s were used.

(3)

(4) Power p l a n t s e l e c t i o n was based on economics i n s t e a d o f U308 u t i l i z a t i o n . The e l e c t r i c a l energy demand t h a t was used f o r these c a l c u l a t i o n s i s shown i n Table D-1. This p r o j e c t e d demand assumes a 5.6% per year growth r a t e u n t i l 1980, and a 5.1% per year growth r a t e from 1980 t o 1990. The growth r a t e decreases each decade u n t i l year 2030, a f t e r which a constant 2.5% p e r year growth r a t e i s assumed. The marginal c o s t o f uranium as a f u n c t i o n o f t h e cumulative q u a n t i t y mined was shown i n Table B-7 o f Appendix B. supply are denoted as cases l L , 2L,

I n t h i s appendix cases t h a t use t h e high-cost uranium

..., w h i l e cases t h a t use t h e intermediate-cost uranium .... As has already been emphasized, i t was assumed

supply a r e denoted as cases lLU, 2LU,

r

f o r these c a l c u l a t i o n s t h a t t h e q u a n t i t y o f a v a i l a b l e uranium was unlimited.

The o n l y


-~

L D-2

Year 1975 1980 1990 2000 2010 2020 2030

also

r e s t r i c t i o n on uranium consumption was based on economics t h a t i s , the marginal cost o f an a d d i t i o n a l pound o f U308 increases as more uranium i s consumed.

-

. Table D-1. Projected Total E l e c t r i c a l Generation

Electrical Energy ( lo1* kWh)

Electrical Energy Growth Rate (%per year)

1.9) 2.5i

5.6 5.1

4.1,

4.1 3.5

Fossil-fueled power plants were represented by nine d i f f e r e n t coal p l a n t types which are i n d i c a t i v e o f d i f f e r e n t coal regions. The p r i n c i p a l differences

between coal p l a n t types are the coal price, the coal energy content, and the s i z e of , 6.1; the demand t h a t can be s a t i s f i e d by each 3.0 coal p l a n t type. The maximum f r a c t i o n of 2.5 14.9 the t o t a l e l e c t r i c a l energy demand t h a t can be s a t i s f i e d by each regional coal p l a n t type i s shown i n Table 0-2. This t a b l e gives the heat content o f the coal f o r each region.

llS6I

The c a p i t a l cost associated w i t h b u i l d i n g a coal p l a n t was assumed t o be 12% lower than the c a p i t a l cost of a LWR, o r $550/kWe ( i n 1/1/77 d o l l a r s ) . Therefore, for nuclear plants t o be b u i l t instead o f coal plants, the f u e l costs o f the nuclear plants must be enough lower than the f u e l cost o f f o s s i l plants t o override t h i s c a p i t a l cost d i f f e r e n t i a l . I f nuclear plants are l e s s expensive than coal plants f o r a l l regions, then a l l o f the new plants b u i l t w i l l be nuclear. Figure D - 1 shows how the nuclear market f r a c t i o n decreases as nuclear plants become more expensive. I f nuclear plants increase i n p r i c e by 20% over the p r i c e where a l l o f the market would be nuclear, the nuclear market f r a c t i o n decreases t o 0.75. An increase o f about 35% i n the p r i c e o f a nuclear u n i t reduces the nuclear market f r a c t i o n t o about 0.34, while a 57% increase r e s u l t s i n a l l o f the new plants b u i l t being f o s s i l - f u e l e d plants. Nuclear power growth projections f o r the LWR on the throwaway cycle are shown f o r both uranium supplies in.Fig. D-2a. For the high-cost uranium supply case, nuclear power peaks a t 500 GWe o f i n s t a l l e d capacity around the year 2005 and then phases out t o about 100 GWe i n 2040. On the other hand, i f t h e intermediate-cost uranium supply i s assumed, nuclear power continues t o grow u n t i l about 2015 t o almost 900 GWe, and then decreases t o about 300 GWe i n 2040. As a r e s u l t , nuclear i s more competitive w i t h coal and captures a l a r g e r share o f t h e market. Figure D-2b shows t h a t r e c y c l i n g plutonium i n LWRs (Case 2L) increases the nuclear power market even more than the assumption o f a l a r g e r uranium supply, and introducing the Pu/U-fueled FBR w i t h recycle (Case 3L) f u r t h e r increases the nuclear market t o 1300 GWe o f i n s t a l l e d nuclear capacity i n the year 2040. The u& u t i l i z a t i o n , defined as the

L i:

L L L t'

D 6 L3 LI

I' 1


D-3

L L

Table D -2. Maximum E l e c t r i c a l Energy Demand S a t i s f i e d by Regional Coal Plants

t o t a l U308 consumed p l u s committed per GWe o f nuclear power constructed through the year 2050, i s a l s o given f o r these

E l e c t r i c%aOl f Sales New England (NE)

Heat ( B t u‘Ontent /lb)

3.9

13,500

13.1

11,783

cases. As noted, r e c y c l i n g plutonium i n LWRs reduces U30s usage by 38% per GWe, w h i l e i n t r o d u c i n g t h e FBR r e s u l t s i n a 62% reduction.

Middle A t l a n t i c

(MA)

L L

supply, 1300 GWe f o r t h e FBR case becomes

East North Central (ENC)

19.5

10,711

West North Central (WNC)

6.6

9 ,408

South A t l a n t i c (SA) East South Central (ESC)

16.6

11,855

almost 1800 GWe i n 2040 (see Case 3LU i n Fig. D-2c).

The nuclear power peak f o r

each o f t h e ore supplies occurs around t h e year 2040, although t h e i n s t a l l e d nuclear capacity i s very f l a t a t t h i s

9.6

11,006

Moutain (MT)

12.2 4.9

6 ,583 9,637

P a c i f i c (PA)

13.5

8,101

West South Central (WSC)

With t h e intermediate-cost U308

point. The disadvantage o f c l a s s i c a l plutonium r e c y c l e i n FBRs i s demonstrated i n Fig. D -2d f o r Case 3L. Here t h e two Pu-fueled reactors are i n s i d e t h e energy center and t h e LEU-LWR i s outside t h e

center. I t can be seen t h a t a f t e r about 2020, t h e r a t i o o f reactors t h a t can be located outside t h e center t o those i n s i d e i s l e s s than u n i t y and r a p i d l y decreasing. I n f a c t , as

RELATIVE POWER COST

[i

i”

HEDL 7805090.50

Fig. D-1.

E f f e c t o f Changing Nuclear Power Costs on the Nuclear Market Fraction.


D-4

25m

2m

d G

I

-

I

I

I

I

2Mo

I

I

I

I

I

b)

I.I

NUQEAR-FOSSIL COMPETITON

HIGH-CnST u& SUPPLY NUCLEAR-FOSSIL COMPETITION

-

U p s UTILRATION 6T U p $ G W $

a oim a ipoo

1

-

l a -

S

L

3

3 1LU

- LWR. INTERMEDIATE COST UJosSUPPLY

-

5mI1

0

wm

I 1990

-

LW THROWAWAY

I

I

zm

mi0

~

I

I

I

m

ma

YEAR

YEAR

am

I ?ow

I

I

I

(c)

NUCLUR-FPSIL COMPETnlON

.m

-

-

L

~ L ULWR F I R Pu RECYCLE. INTERMEDIATE-COST U30, SUPPLY

-

"1

Bl

YEAR

zsm

I

I

I

I

I

25m

1-1

I

1

I

If)

WIGH-COST u+a WPPLV. NUCLEAR-FOSSIL WHPETITION

CONVERTER Punh TRANSMUTER

a

i

NUCLEARdOSYL COMPETITION

YIO

P 2880

6L 1680

$2+

2 ism

S

-

0 1900

YEAR

5LU

I

IWO

I

'loo0

INTERMEDIATE-COST U 3 0 SUPPLY ~

I

I

2010

2020

I 2030

YEAR

Fig. D-2. Installed Nuclear Capacities During Years 1980-2040 (or 2050) for Various Power Systems Including Both Nuclear and Coal Power Plants.

2040

c


D-5

Fig. D-2

25W

(cont.) I

I

-

2mo

I

I

I

I

(I)

FBR Pwqh TQANSMUTER NUCLEAR-FOSSIL COMPETITION

5 J

1 Imo-

0 19m

1

I

1990

2mo

I 2010 YEAR

I 2020

I 2030

2Mo YEAR


D- 6

the system becomes less and'less dependent upon uranium ore and more and more upon plutonium, the energy support r a t i o w i l l approach zero.

*

The denatured f u e l cycle Cases 4L, 5L, and 6L are compared w i t h the throwaway cycle i n Fig. D-2e. Nuclear market penetration f o r plutonium throwaway (Case 4L) i s n o t subs t a n t i a l l y greater than f o r the throwaway cycle (Case 1L). The peak penetration i s about 630 GWe o f i n s t a l l e d nuclear capacity versus 500 GWe f o r the throwaway cycle. However, i f the plutonium i s u t i l i z e d i n an LWR Pu/Th converter (Case 5L), the maximum nuclear penetration i s 1000 GWe, which i s a f a c t o r o f two greater than f o r the throwaway cycle and, furthermore, the peak does n o t occur u n t i l more than 10 years l a t e r . Introduction o f the FBR w i t h a Pu-U core and thorium blankets (Case 6L) r e s u l t s i n a peak penetration o f 1250 GWe i n about 2025. A f t e r 2025, the nuclear market f r a c t i o n i s constant because the system i s e s s e n t i a l l y independent o f uranium, which i s becoming increasingly more expensive.

L

L L

With respect t o U308 u t i l i z a t i o n , Fig. D-2e shows t h a t the Pu/Th converter case has s l i g h t l y b e t t e r ore u t i l i z a t i o n (by 7%) than c l a s s i c a l plutonium recycle i n LWRs (Case 2L i n Fig. D-2b). Furthermore, plutonium "transmutation' i n Pu-U FBRs also has b e t t e r U308 u t i l i z a t i o n (by 12%) than c l a s s i c a l plutonium recycle i n FBRs (compare Cases 3L and 6L). The reason f o r these trends i s t h a t the 233U f u e l t h a t i s being bred i s worth more as a fuel i n thermal reactors than the plutonium t h a t i s being destroyed. The e f f e c t o f a l a r g e r uranium supply on the market penetration f o r converters and FRBs t h a t produce * 3 % i s shown i n Figs. D-2f and D-29. For both cases (5 and 6), the large uranium supply increased the maximum nuclear penetration by about 450 GWe. Case 7L introduced a denatured 233U-fueled FER t o the 6L case, and Case 8L i s i d e n t i c a l t o Case 7L except t h a t the FBR w i t h a Pu-U core i s replaced w i t h an FBR w i t h a Pu-Th core. The maximum nuclear penetration f o r Cases 7L and 8L are compared w i t h 6L i n Fig. D-2h. The denatured 23%-fueled FBR doesn't have any impact because t h i s reactor i s competing w i t h less expensive 233U-fueled LWRs and therefore i s n ' t b u i l t . The nuclear market penetration f o r Case 8L i s seen t o decrease a f t e r about 2020. This i s because the neutronics properties o f FBRs fueled w i t h Pu-Th are degraded s i g n i f i c a n t l y from those fueled w i t h Pu-U. As a r e s u l t , the doubling time o f these reactors i s longer and the cost i s higher. The degraded neutronics o f the Pu-Th FBRs are r e f l e c t e d i n the U308 u t i l i z a t i o n o f Case 8L where the ore usage per GWe i s almost 50% higher than f o r Case 6L. The objective i n b u i l d i n g FRBs w i t h Pu-Th cores i s t o increase the

23%

production

and therefore the r a t i o o f reactors located outside the energy center t o those i n s i d e the

*

c

The nuclear reactors t h a t are available i n Case 5L w i t h nuclear-fossil competition are s i m i l a r t o Case 5UL described i n the other sections o f t h i s report. However, i n 5L the denatured 235U-fueled LWR i s n ' t b u i l t because o f economics. Therefore, the s o l u t i o n more c l o s e l y resembles Case 5TL.

I;


D-7

energy center. It can be seen from t h e nuclear power growth p a t t e r n s f o r Cases 6L and 8L, shown i n Figs. D - 2 i and D-2j, t h a t t h e energy support r a t i o f o r Case 8L i s higher. The degraded neutronics of t h e FBRs f u e l e d w i t h Pu-Th are r e f l e c t e d i n t h e U308 u t i l i z a t i o n o f Case 8L where the o r e usage per GWe i s almost 50% higher than f o r Case 6L (see Fig. 0-2h). However, f o r most years t h e t o t a l amount o f energy t h a t i s a v a i l a b l e t o be b u i l t i n t h e energy centers i s about t h e same f o r Case 8L as i t i s f o r Case 6L because the t o t a l amount o f nuclear energy i s lower. Key selected r e s u l t s from t h e n u c l e a r - f o s s i l competition c a l c u l a t i o n s are presented i n Tables D-3 and D-4 f o r high-cost and intermediate-cost U,08 supplies r e s p e c t i v e l y . Each t a b l e presents t h e cumulative capacity o f nuclear and f o s s i l p l a n t s b u i l t through year 2050, t h e t o t a l system costs, t h e annual coal consumption i n 2025, data on uranium and enrichment u t i l i z a t i o n , t h e i n s t a l l e d capacity o f each r e a c t o r type i n year 2026, and t h e l e v e l i z e d power c o s t of each r e a c t o r type f o r a r e a c t o r s t a r t i n g up i n year 2025. The most s t r i k i n g conclusion t h a t can be drawn from t h e comparison o f l e v e l i z e d power costs o f

[i

each r e a c t o r type i s t h a t t h e r e i s n ' t a l a r g e d i f f e r e n c e .

The reason, o f course, i s t h a t t h e t o t a l amount o f uranium consumed doesn't vary much from case t o case because when

uranium becomes expensive, f o s s i l p l a n t s are constructed i n place o f nuclear plants.

This

p o i n t i s demonstrated i n Table D-5, which shows t h e time behavior o f t h e U308 p r i c e . I t can be seen from t h i s t a b l e t h a t t h e d i f f e r e n c e s i n t h e p r i c e o f U308 f o r t h e d i f f e r e n t nuclear systems a r e n o t large.


D-8

Table D-3. Summary o f Results f o r Cases Assuming High-Cost U308 Supply, an E l e c t r i c a l Energy Growth Projection, and Power Systems Including Both Nuclear and Coal Power Plants 4L 81 5L 6L lL 2L 3L n -

Cmulative Capacity Built (me)through 2050 liuclear

705 4611

FaSSil *tern Costs (SB) 1977 thmugh 2050 Discomted B 4 1/21 7 1/21 101 Annual coal connmpcim in 2025 (10' tans)

2050 Total c"$"ittdU308 through 2050 (10 tosls)

2663 2653

933 4383

1684 3632

2597 2719

2595 2721

1909 3407

1804 787 479

1733 764 470

1701 758 468

1806 791 483

1724 761 468

1703 760 469

1703 760 469

in8 761 469

5.22

3.72

3.15

4.79

3.59

2.91

2.91

3.25

2.92 3.42

3.50 4.75

3.56 4.60

2.88 3.13

3.62 4.75

3.68 4.33

3.68 4.33

3.69 4.70

3.55

4.92

5.06

3.18

4.85

4.37

4.37

4.77

65 (20ll)

73 (2009)

72 (ZOOS)

75 (2015)

80 (2011)

80 (20W

79 (rn15)

2.12 5.04

3.11 3.10

2.89 1.90

2.53 3.41

3.40 2.88

3.11 1.68

3.11 1.68

3.37 2.50

3.01

1.96

1.09

2.71

2.02

1.20

1.20

1.77

54

(ZOOS) (3)

'

L

t]

I n s t a l l e d Capacity (GWe) i n Year 2026/Levelited Power Costs (Mill/Kwhr) i n Year 2025 Reactor LWR-U5(LE)/U

US{LE)/u-EE

1L 36/23.2 225/22.3

8L n 6L SL 4L 3L 2L 579/21.1

us m1/um u3 (DE)/U/Th

wu

336/22.3

513/20.8

113/21.6

661/21.2

594/20.7

594/20.7

668/20.8

-

189/22.5 157/20.0

0/23.5 120/20.6

0/23.2 190/19.6

0/23.2 190/19.6

0/23.1 230/20.8

-

181/20.1

52/22.1

52/22.1

102/23.0

-

408/19.4

40W19.4

-

196/19.5

pu/Th

-

FBR-PU-U/u

444/18.4

-

P U U m

Pu-mm

-

u3-um

Fossil Total Nuclear

1934

1280

1042

261

915

llS3

1736 459

1233 962

951 1244

0/23.0 951 1244

104/22.6

0/25.0 1091 1104

Cumulative U308 comsumed through 2050 (including forward comnitments) per cumulative nuclear capacity b u i l t through 2050. (2) Cumulative enrichment requirements through 2050 per cumulative nuclear capacity b u i l t through 2050. (3) Year i n which maximum enrichment requirements occur. (1)

E L

c


0-9

Table D-4. Summary o f Results for Cases Assuming Intermediate-Cost U308 Supply, an Electrical Energy Growth Projection, and Power Systems Including . Both Nuclear and Coal Power Plants

L

Cunulative Capacity Built (ab) through 2050 Nuclear Fossil System Costs [$Bl 1977 through 2050 Discounted e 4 1/2\ 7 1/28 101 Annual Coal Consunption in 2025 (lo9 tons) ative U 0 Consunption

L

1LU 2LU 5LU 4LU 5LU 6LU 7LU 8LU 1257 4059

2523 2793

3415 1901

1815 3501

2701 2615

3296 2020

3338 1978

2727 2589

1732 759 466

1652 738 459

1622 734 458

1743 770 474

1643 755 458

1624 735 459

1624 735 459

1638 736 458

4.13

2.28

1.92

3.41

2.22

1.82

1.77

2.01

4.75 6.10

4.60 7.44

4.43 6.29

4.41 5.75

4.63 7.40

4.48 5.75

4.50 5.75

4.60 6.62

6.28

7.88

6.90

5.94

7.99

5.87

5.89

6.84

93 (2013)

103 (2025)

93 (2011)

119 (2011)

111 (2023)

101 (2011)

102 (2011)

103 (2017)

3.80

4.87

3.96

4.78

5.26

4.12

4.12

4.73

5.00

3.12

2.02

3.27

2.96

1.78

1.76

2.51

3.02

1.93

1.16

2.63

1.95

1.25

1.23

1.73

3 k n s ) tiLough

2026 2050 T o t a l Cngnitted U308 throu& 2050 (10 tons) Maximm Annual Enrichment throu& 2050

zFwem$

Cunulati e Enridment through 2050 (106 sku) u308 utilization(') Enr'chment

(10' s I \ u / a e )

Installed Capacity (GWe) in Year 2026/Levelized Power Costs (Mills/Kwhr) in Year 2050 Reactor

1LU -

LwR-US(LE)/U

61/22.4

US(LE)/U-EE

i il

2LU 1028/19.8

3LU 827/19.4

6LU

-

235/20.6 1108/19.9

4LU,

874/19.2

872/19.2 1028/19.7

489/21.6 336/20.4

0/21.9 143/19.5

0/21.3 219/19.6

0/21.3 221/19.6

0/21.7 280/19.7

-

235/18.9

63/20.8

56/20.8

119/21.3

486/19.2

509/19.2

5LU

675/21.6

u5 W)/U/Th U3(E)/U/Th 441/19.2

pu/v

269/18.7

PUm -

FBR-h-U/U PU-U/Th

516/17.3

-

-

-

Pu-Th/Th-

us-urn Fossil Total Nuclear

(1)

L

tW

8LU -

0/23.7

136/20.6 0/23.5

1458

725

583

1135

7lO

553

537

632

736

1470

1612

1060

1485

1642

1658

1563

Cumulative U3O8 consumed through 2050 (including forward c o m i tments) per cumulative nuclear capacity bui 1 t through 2050. (2) Cumulative enrichment requirements through 2050 per cumulative nuclear capacity built through 2050. (3) Year in which maximum enrichment requirements occur,


D-10

Table D-5.

Variation o f U308 Price with Time f o r Various Nuclear Cases

Year

1L

2L

3L

U308 Price ($/lb) 6L 5L 4L -

1987

76

81

83

73

82

83

83

82

1997

104

112

114

99

113

114

114

113

2007

136

150

153

130

150

153

153

151

2017

157

177

175

151

177

175

175

175

2027

167

185

179

158

184

180

180

180

2037

172

189

180

158

186

180

180

180

2047

173

195

180

158

189

180

180

180

-

7L -

8L -

L

f

c c


-

0 RNL 5388

Di s t

. Category UC-80

Internal Distribution

-Lid

1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14. 15. 16. 17. 18. 19. 20. 21. 22. 23. 24. 25. 26. 27. 28. 29. 30. 31.

L. R. T. W. S.

D. H.

J. R. W. T. W. J. T. A. J. J. D. G. M. E. P. P. W. J. R. D.

3. D. P. H.

M. W. A. 35. R. 36. F. 37. B. 32. 33. 34.

S. Abbott G. Alsmiller D. Anderson B. Arthur, ORGDP Baron E. Bartine I. Bowers T. Bradbury, ORGDP E. Brooksbank D. Burch J. Burns L. Carter C. Cleveland E. Cole G. Croff G. Delene R. Engel E. Ferguson F. Flanagan H. Fontana H. Gift M. Greebler (Consultant) M. Haas 0. Harms F. Harvey F. Hibbs T. Ingersoll D. Jenkins R. Johnson R. Kasten E. Knee Levenson B. Loewenstein (Consultant) L. Lotts S. towrie C. Maienschein F. Maskewitz

38. 39. 40. 41. 42. 43. 44. 45. 46. 47. 48. 49. 50. 51. 52. 53. 54. 55-75. 76. 77. 78. 79. 80. 81. 82. 83. 84. 85. 86, 87. 88. 89-90. 91. 92-131. 132-133. 134.

H. J. F. K. A. R. H. R. B. R. J. P. T.

R. Meyer D. McGaugh R. Mynatt J. Nott R. Olsen W. Peelle Postma T. Prim E. Prince H. Rainey P. Renier S. Rohwer H. Row T. Santoro L. Selby L. Simard R. Smolen Spiewak G. Stockdale E. T i l l B. Trauger E. Trammel1 E. Uhrig (Consultant) E. Vath L. Vondra R. Vondy R. Weisbin R. White Wilson (Consultant)

R. D. R. G. I. W. J. D. H. R. J. B. D. C. 3. R. R. G. Wymer

A. Zucker Central Research Library Document Ref. Section

EPD Reports Office Laboratory Records Dept. Laboratory Records , RC

External Distribution DOE, Washington, D.C.

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135. 136. 137. 138. 139.

20545

W. W. Ballard, Asst. Director, Fuel Cycle Development E. S. Beckjord, INFCE Coordinator Harold Bengelsdorf, Office of Nuclear Affairs S. T. Brewer, Director, Program Planning and Analysis P h i l l i p Clark, Assoc. Director for Reactors, Div. o f Naval Re a c tor s


DOE, Washington, D.C.

140- 142. 143. 144. 145. 146. 147. 148. 149. 150. 151. 152. 153. 154. 155. 156. 157. 158. 159. 160-180. 181. 182.

20545 (contd.)

E. G. DeLaney, NASAP Control Office, Office of Fuel Cycle Eval uati on D. E. Erb, Division of Reactor Research and Technology H. Feinroth, Division o f Nuclear Power Development Neil Goldenberg, Division of Advanced Systems and Materials Production E. J . Hanrahan, Director, Office of Fuel Cycle Evaluation J. R. Humphreys, Program P l a n n i n g and Analysis Hugh Kendrick, Office of Fuel Cycle Evaluation M. W. Koehlinger, Program Planning and Analysis P. M. Lang, Asst. Director, L i g h t Water Reactor Programs K. 0. Laughon, Asst. Director, Gas Cooled Reactor Programs D. E. Mathes, Office of Fuel Cycle Evaluation W. H. McVey, Division of Nuclear Power Development Marvin Moss, Office of Energy Research C. W. Newstead, Office of Energy Research J. A. Patterson, Chief, Supply Eval uati on Branch, Division of Urani um Resources and Enrichment A. Pressesky, Director, Division of Nuclear Power Development W. F. Savage, Division of Advanced Systems and Materials Production W. S . Scheib, Division of Nuclear Power Development C. Sege, Office of Fuel Cycle Evaluation S . Strauch, Office of Fuel Cycle Evaluation K. A. Trickett, Office of the Director, Division of Nuclear Power Development

DOE, Oak Ridge

183. Asst. Manager f o r Energy Research and Development 184. Director, Nuclear Research and Development Division Federal Agencies 185. 186.

187. 188. 189. 190. 191. 192.

D. L. Bell, Tennessee Valley Authority, 503 Power Bldg., Chattanooga, TN 37401 John Boright, Director, Office of Energy & Technology, Department of State, Rm. 78-30, Washington, DC 20520 D. T. BradshBw, Tennessee Valley Authority, 503 Power Bldg., Chattanooga, TN 37401 Greg Canavan, Office of Chief o f S t a f f , Air Force, Pentagon, Washington, DC 20301 John Depres, CIA Headquarters, 7E 47, Washington, DC 20505 H. L. Falkenberry, Tennessee Valley Authority, 503 Power Bldg., Chattanooga, TN 37401 Joseph Kearney, Office of Management and Budget, 17th and H S t r e e t , NW, Washington, DC 20036 S . N. Keeney, Arms Control and Disarmament Agency, Rm. 5934, New State Bldg., Washington, DC 20451


It

e

Federal Agenci es (con t d . ) 193.

Louis V. Nosenzo, OES/NET 7830, Department of S t a t e ,

194.

Joseph Nye, Department of S t a t e , Rm. T-7208, Washington,

195. 196.

L L

197. 198.

Washington , DC 20520

DC 20520 Robert Rochlin, Arms Control and Disarmament Agency, Rm. 4930, New S t a t e Bldg., Washington, DC 20451 Lawrence Scheinman, Department of S t a t e , Rm. T-7208, Washington, DC 20520 James Sheaks, Arms Control and Disarmament Agency, Rm. 4933, New S t a t e Bldg., Washington, DC 20451 Charles Van Doren, Asst. Director, Arms Control and Disarmament Agency, Rm. 4930, New S t a t e Bldg., Washington, DC 20451

Outside Organizations 199. 200. 201.

L

202. 203. 204. '

205. 206.

t

207. 208. 209.

u

210. 211. 212. 21 3. 214. 21 5.

I' f' ki

216. 217.

W. E. Black , Hanford Engineering Development Laboratory, P. 0. Box 1970, Richland, WA 99352 George Bunn, University of Wisconsin, Madison, WI 53706 A. Carnesale, Harvard University, #9 Divinity Avenue, Cambridge, MA 02138 Y. Chang, Argonne National Laboratory , 9700 South Cass Avenue , Argonne, IL 60439 Thomas Cochran , National Resources Defense Counci 1 , 91 7-1 5th St., NW, Washington, DC 20005 Gordon Corey, Commonwealth Edison E l e c t r i c , P. 0. Box 767, Chicago, IL 60690 Russel 1 Crowther , General E l ectri c Company , 175 Curtner Ave. , San Jose, CA 95125 Joe Cupo, Westinghouse E l e c t r i c Corporation, P. 0. Box 355, Pittsburgh, PA 15230 W. K. Davis, Bechtel Power Corporation, 50/11/813, P. 0. Box 3965, San Francisco, CA 93119 J. M. de Montmollin, Sandia Laboratories, Dept. 1760-A, A1 buquerque, NM 87185 M. J. Driscoll, Massachusetts I n s t i t u t e of Technology, 138 Albany S t r e e t , Cambridge, MA 02139 M. C. Edlund, Virginia Polytechnic I n s t i t u t e , Blacksburg, VA 24060 J. F. Foran, Resource Planning Associates, Inc. , 3 Enbarcadero Center, Suite 2080, San Francisco, CA 94111 T. K. Glennan, 11483 Waterview Cluster, Reston, VA 22090 H. Goldstein, Columbia University, 520 W. 120th S t . , New York, NY 10027 L. Gordon, Resources f o r the Future, 1977 Massachusetts Ave., Washington , DC 20036 D. R. Haffner , Hanford Engineering Development Laboratory , P. 0. Box 1970, Richland, WA 99352 R. W. Hardie , Hanford Engi neeri ng Development Laboratory, P. 0. Box 1970, Richland, WA 99352 William Harris, RAND Corporation, 1700 Main Street, Santa Monica, CA 90406


Outsi de Organizations (contd. ) 218. 219. 220. 221. 222. 223. 224. 225. 226. 227. 228.

229. 230. 231. 232. 233. 234. 235. 236. 237. 238. 239. 240. â&#x20AC;&#x2DC;241. 242. 243. 244.

Chuck Hebel , Xerox Paâ&#x20AC;&#x2122;lo Alto Research Center, 3333 Coyote Hill Rd. , Palo Alto, CA 94303 T. M. Helm, Hanford Engineering Development Laboratory, P. 0. Box 1970, Richland, WA 99352 M. Higgins, Science Applications, Inc., 8400 Westpark Dr. , McLean, VA 22101 William Higinbotham, Brookhaven National Laboratory, Associated Universities, Inc. , Upton, NY 11973 Fred Hoffman, RAND Corporation, 1700 Main S t r e e t , Santa Monica, CA 90406 William W. Hogan, Stanford University, Stanford, CA 94305 W. G. Jolly, Hanford Engineering Development Laboratory. P. 0. Box 1970, Richland, WA 99352 3. M. Kallfelz, Georgia I n s t i t u t e of Technology, Atlanta, GA 30332 Arthur Kantrowi t z , AVCO Everett Research Lab. , 2385 Revere Beach Parkway, Everett, MA 02149 Walter Kato, Argonne National Laboratory, 9700 South Cass Ave. , Argonne, IL 60439 John Kearney, Edison Electric I n s t i t u t e , 1140 Connecticut Ave. , Washington , DC 20036 Herbert Kouts , Brookhaven National Laboratory, Associated Universities, Inc. , Upton, NY 11973 F. W. Kramer, Westinghouse Nuclear Fuel Division, P. 0. Box 355, Pittsburgh, PA 15230 Myron Kratzer, International Energy -Associates , 2600 Virginia Ave. , NW, Suite 200, Washington, DC 20037 Id. C. Lipinski, Argonne National Laboratory, 9700 South Cass Ave. , Argonne, IL 60439 A. S. Manne, Terman Eng. 432A, Stanford University, Stanford, CA 94305 G. Nugent, Burns 81 Roe, P. 0. Box 663, Route 17, South, Paramus, NJ 07652 R. P. Omberg, Hanford Engineering Development Laboratory, P. 0. Box 1970, Richland, WA 99352 6. Pasternak, Booz-Allen and Hamilton, 4330 East-West Highway, Bethesda, MD 20014 R. G. Post, University of Arizona, Tucson, AZ 85721 Richard Richels, E l e c t r i c Power Research I n s t i t u t e , P. 0. Box 10412, Palo Alto, CA 94303 C. L. Rickard, General Atomic Company, P. 0. Box 81608, San Diego, CA 92138 David Rossin, Commonwealth Edison Electric Co., P. 0. Box 767, Chicago, IL 60690 H. S. Rowen, Stanford University, Stanford, CA 94305 Thomas Schell ing, JFK School of Government , Harvard University, Cambridge, MA 02138 N. L. Shapiro, Combustion Engineering, 1000 Prospect Hill Rd., Windsor, CT 06095 M. R. Shay, Hanford Engineering Development Laboratory, P. 0. Box 1970, Richland, WA 99352

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Outside Organizations (contd. ) 245. 246. 247. 248. 249. 250. 251. 252. 253.

u

254. 255. 256. 257-396.

B. I . Spinrad, Dept. of Nuclear Engineering, Oregon S t a t e Uni versi t y , Corval 1is , OR 97331 Chauncey S t a r r , Electric Power Research I n s t i tute , 341 2 H i 11 View Ave., Palo Alto, CA 94303 H. B. Stewart, Nuclear Technology Evaluations Company, 4040 Sorrento Valley Blvd., Suite F, San Diego, CA 92121 S. M. S t o l l e r , S. M. S t o l l e r Corporation, 1250 Broadway, New York, NY 10001 E. Straker , Science Appl i cations , Inc. , 8400 Wes tpark Dri ve , McLean, VA 22101 C. E. T i l l , Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 James Tulenko, Babcock and Wilcox, P. 0. Box 1260, Lynchburg, VA 24505 R. F. Turner, General Atomic Company, P. 0. Box 81608, San Diego, CA 92138 Frank von Hippel , Program on Nuclear Pol icy A1 t e r n a t i ves , Center f o r Environmental Studies, Princeton University, Princeton, NJ 08540 Eugene Weinstock, Brookhaven National Laboratory, Associated Universities, Inc. , Upton, NY 11973 A. Weitzberg, Science Applications, Inc. , 8400 Westpark Drive, McLean, VA 22101 Albert Wohlstetter, 518 S. Hyde Park Blvd. , Chicago, IL 60615 For Distribution as Shown inTID-4500 under UC-80, General Reactor Techno1ogy

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*US. GOVERNMENT PRINTING OFFICE: 1979-640-079/374

ORNL-5388  

Interim Assessment of the Denatured 233UFuel Cycle: Feasibility and Nonproliferation Characteristics

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