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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:

I

!

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.

L

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 '

L

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.


I;

6-14

UZJS

USILEINS

WYI

SWU

HM

7

I

THROWAWAY

SWU

U5ILEIN-

HM

L L

I:

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

h.'


i

6-15

n

L L

?

1'

fa P" Urn UM

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.)

,

L

n

UEDL7801.7&5

i'

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

n

c

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:

L

Y9

An

exception t o t h i s