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ORNL TM-1060

Contract No. W-7405-eng-26

REACTOR DrJISION

MOLTEN SALT CONVERTER REACTOR

Design Study and Power Cost Estimates f o r a 1000 Mwe S t a t i o n L. G. Alexander, W. L. Carter, C. W. Craven, D. B. Janney, T. W. Kerlin, and R. Van Winkle

SEPTEMBER 1965

W

OAK RIDGE NATIONAL LABORATORY O a k Ridge, Tennessee mer ated by UNION CARBIDE CORPORATION f o r the U. S. ATOMIC ENERGY COMMISSION


iii

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FOREWORD

i i

A molten-salt-breeder reactor was evaluated at the Oak Ridge National

Laboratory beginning in 1959. Because a number of the features postulated had not been demonstrated at that time, the realization of a breeder appeared to lie rather far in the future. Accordingly, the study of the near-term, one-region, one-fluid molten-salt convert.er described in this report was begun in July 1961 and completed in December 1962. Since then, several advances have been made in molten-salt technology which make the

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breeder reactor m w h less remote and modify some of the conclusions in this report. Briefly, these advances include: 1. Progress in core graphite design which greatly simplifies previous problems of separating the core into two regions - one for the uraniumbearing fuel salt and one for the thorium-bearing blanket salt. The new design utilizes a liquid-lead seal around the tops of graphite tubes containing fuel salt that allows the tubes to expand or contract freely while maintaining an absolute seal between fuel and blanket fluids.* The addition of a blanket results in a much better conversion than obtained in this report and leads directly to an attractive breeder. 2. Thermal engineering studies which show that the Loeffler boiler

4

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system can advantageously be replaced by a supercritical boiler. Thermal stress problems are reduced, overall thermodynamic efficiency is increased, and capital costs are considerably reduced. In addition, studies of sodium metal and of mixtures f alkali carbonates show that if either of these inexpensive materials can be safely used for the intermediate coolant in place of the costly li ium-beryllium fluoride mixtures postulated in this study, then f’urther 1 e cost reductions can be realized?

*E. S . Bettis, Oak Ridge National Laboratory, personal communication with L. G. Alexander, Oak Ridge National Laboratory, January 1965.

t ~ .W. Collins, oak Ri tional Laboratory, personal communication with L. G. Alexander, Oak Ridge National Laboratory, January 1965.


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74

iv 3. ‘

A f u e l p u r i f i c a t i o n process based on simple d i s t i l l a t i o n ? which

not only reduces processing costs§ but permits reuse of the c a r r i e r

salts

- an advantage not assumed i n t h i s

study.

As a r e s u l t of these developments, we believe t h a t f u e l cycle costs f o r a two-region breeder based on 1965 technology w i l l be only 0.3 t o 0.4 mill=

compared t o the 0.68 mill/kwhr sham i n Table 6.10 f o r the IGCR.

?M. J. Kelley, Oak Ridge National Laboratory, personal communication with L. G. Alexander, Oak Ridge National Laboratory, January 1965. SW- L. Carter, Oak Ridge National Laboratory, personal comunication with L. G. Alexander, Oak Ridge National Laboratory, January 1965. *H. F. Bauman, Oak Ridge National Laboratory, personal communication with L. G. Alexander, Oak Ridge National Laboratory, January 1965.


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CONTEXTS Page

........................................................... 1. SUMMARY ........................................................ 1.1 Description ................................................ 1.2 Fuel Reprocessing .......................................... 1.3 Nucleas and Thermal Performance ............................ 1.4 Fuel Cycle Cos$ ............................................ 1.5 Power Costs ................................................ 1.6 Advanced IGCR .............................................. 1.7 Post S c r i p t - January 1965 ................................. 2. INTRODUCTION ................................................... 2 . 1 Purpose, Scope, and Method of Approach ..................... 2.1.1 Figure of Merit ...................................... 2.1.2 Reactor Concept ...................................... 2.1.3 Procedure ............................................ 2.2 Status of Molten S a l t Reactor Development .................. 2.2.1 Early Work ........................................... 2.2.2 The Molten S a l t Reactor Program ...................... 2.2.3 Fuel Development ..................................... ABSTRACT

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1 3 3

3 4

5 6

7 8 8

8

8 8 9

9

10 10

2.2.4 2.2.5

Container Development Moderator Development

2.2.6

Component

2.2.7

Reactor V

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

14

2.2.8

Molten S

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

14

2.2.9

Molten S

12

-..

15

14

2.2.14

..................... Molten t a t i o n and Special .................................. Remote Maint ................................... Chemical of Molten S a l t Fuels ............. Fluoride V o l a t i l i t y and HF' Solution Processes ........

18

2.2.15

Thorex Process

19

2.2.10 2.2.11 2.2.12 2.2.13

hd

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

1

Freeze Valve

Freeze Flanges

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

15

16 16

18


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CONTENTS (continued)

2.2.16 2.2.17 2.2.18

.................. ..................................... Molten Salt Reactor Studies ......................... Fractional Crystallization Process Other Processes

20 20 21

............................ 22 .......................................... 23 3. 3.1 Design Bases .............................................. 23 Reactor Concept ..................................... 23 3.1.1 3.1.2 Design Calculations ................................. 23 Station Power ....................................... 23 3.1.3 Plant Utilization Factor ............................ 24 3.1.4 3.1.5 Therm1 Efficiency .................................. 24 . 3.1.6 Fueling Cycle ....................................... 24 3.1.7 Processing .......................................... 24 3.1.8 Feed and Recycle .................................... 25 3.1.9 Isotopic Composition of Lithium ..................... 25 3.1.10 Energy Conversion System ............................ 25 3.1.11 Primary Heat Exchanger Requirements .................. 26 3.1.12 Minimum Salt Temperatures ........................... 26 3.2 Cost Bases ................................................ 26 3.2.1 Value of Fissile Isotopes ........................... 26 3.2.2 Value of Thorium .................................... 27 Value of LiF(99.995$ ‘Li) ........................... 27 3.2.3 3.2.4 Value of BeF2 ....................................... 27 3.2.5 Value of Base Salt .................................. 27 3.2.6 Cost of Compounding and F’urifying Fuel Salt ......... 27 27 3.2.7 ..INOR -8 Cost ......................................... Moderator Graphite Cost ............................. 28 3.2.8 3.2.9 A n n u a l Fixed Charges ................................ 28 3.2.10 Central Fluoride Volatility Plant Processing Charges ............................................. 29 3 . 3 Special Assumptions ....................................... 29 3.3.1 Permeation of.Graphiteby Salt ...................... 29 Molten Salt Reactor Experiment BASES AND ASSuMpTIONs

2.3

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vi i CONTENTS (continued) Page

.................. 3.3.3 Corrosion Products .................................. 3.3.4 Approach t o Equilibrium ............................. 4 . DESCRIPTION OF MSCR CONCEPT ................................... 4.1 General Description ....................................... 3.3-2

................................................. 4.3 Structures ................................................ 4.4 Primary System Components ................................. Reactor Vessel ...................................... 4.4.1 4.4.2 Moderator Structure ................................. 4.4.3 Fuel-Salt Circulating Pumps ......................... Primary Heat Exchanger .............................. 4.4.4 4.5 Intermediate Cooling System ............................... 4.5.1 Introduction ........................................ 4.5.2 Coolant S a l t Pumps .................................. 4.5.3 Steam Superheaters .................................. 4.5.4 Steam Reheaters ..................................... 4.6 Power Generation System ................................... 4.6.1 Introduction ........................................ 4.6.2 Loeffler Boiler System .............................. 4.6.3 Steam Circulators ................................... 4.2

S i t e Plan

...................................... 4.7 Reactor Control System ................................... 4.7.1 Introduction ........................................ 4.7.2 Shim Control ........................................ 4.7.3 Emergency Control ................................... 4.8 S a l t Handling Systems ..................................... 4.8.1 Introduction ... ............................... 4.8.2 m e 1 S a l t Prepax ............................... 4.8.3 Coolant S a l t Preparation ............................

m

4.6.4

CI

6.) a 4

3e

w

Permeation of Graphite by 135Xenon

!Turbogenerator

4.8.4

Reactor S a l t P u r i f i c a t i o n

4.8.5

Coolant-Salt Purification

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

30

30 31 32 32

32

40 44 44 45 45 48 51 51

51

51 53 53 53 54 55

57

57 57 58 58 59 59 59 61

61 . 62


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CONTENTS (continued) Page

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

4.8.6 Reactor Salt Charging System 4.8.7 .IntermediateCoolant Charging System 4.8:8 UF'4 Addition Facility 4.8.9 Fuel Salt Drain and Storage System 4.8.10 Coolant Salt Drain System 4.8.11 Spent F'uel Withdrawal System 4.8.12 High Level Radioactive Salt Sampler 4.8.13 Coolant Salt Sampl5ng 4.8.14 Freeze Valves 4-9 Auxiliasy Services and Equipment 4.9.1 Introduction

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

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

63 63 63 64 66 66 67 69 69 70 70 70 72 72 73

.... 4-9.3 Reagent Gas Supply and Disposal System ............. 4.9.4 Waste Gas System ................................... 4.9.5 Liquid Waste Disposal System ....................... 4.9.6 Coolant Pump Lubricating O i l Systems ............... 74 4.9.7 Beheating System .................................. 74 4.9.8 Auxiliary Power .................................... 74 4.9.9 Service Water System ............................... 75 4.9.10 Control and Station Air Systems .................... 75 4.9.11 Cranes and Hoists .................................. 75 4.9.12 Instrumentation and Control ........................ 76 4.9.13 Plant Utilities .................................... 76 4.10 MSCR Design Specifications .............................. 77 5. PROCESSING .............................................. 91 5.1 Reprocessing System ..................................... 91 5.2 Fluoride Volatility Central Plant ....................... 91 5.2.1 Process Design ..................................... 92 5.2.2 Shipping ........................................... 94 5.2.3 Prefluorination Storage Tanks ...................... 94 5.2.4 Fluorinator ........................................ 96 5.2.5 CRP Trap and NaF Absorbers ......................... 98 4-9.2

Helium Cover Gas Supply and Distribution System

B


ix CONTENTS (continued) Page

.......................................... 5.2.7 Reduction Reactor ................................... 5.2.8 Transfer Tanks ...................................... 5.2.9 Waste Storage Tanks ................................. 5.2.10 Freeze Valves ....................................... 5.2.11 Samplers ............................................ 5.2.12 Biological Shield ................................... 5.2.13 Process Equipment Layout ............................ 5.2.14 Plant Layout ........................................ 5.2.15 Capital Cost Estimate ............................... 5.2.16 Operating and Maintenance Cost Estimates ............ 5.2.17 MSCR Irradiated Fuel Shipping Cost .................. 5.2.18 MSCR Unit Processing Cost ........................... 5.3 Thorex Central Plant ...................................... 5.3.1 Head-End Treatment .................................. 5.3.2 Solvent Extraction .................................. 5.3.3 Tail-End Treatment .................................. 5.3.4 Processing Costs .................................... 5.4 Cornpaxison of Processing Cost Estimates ................... 6. FUEL CYCLE ANALYSIS ........................................... 6.1 Analysis of Nuclear System ................................ 5.2.6

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Cold Traps

................................... ............................... .................................. . ................. ............................ ............................ ....... ................................. 6.2.4 Fuel Volume ......................................... 6.3 Analysis of Chemical System ............................... 6.3.1 Thorium-232 ......................................... 6.1.1 Computer Programs 6.1.2 Reactor Physics Model 6.1.3 Cross Section Data 2 Analysis of Therma chanical System 6.2.1 Maximum Fuel ture 6.2.2 Minimum Fuel atme .2.3 Velocity

6.3.2

Protactinium-233

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

100

100 100 100 102 102 102 102 105 105 113 113 117 118 119 121 121

123 125 127

127 127 129 129 130 131 131 131

133 134 134 135


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W' CONTENTS (continued)

......................................... ......................................... ......................................... ......................................... ....................................... ......................................... ................ ................................................ ...................... ........................ .............................. .................................. ................................... .................................. ...................................... ..................................... 6.5.3 Inventories and Processing Rates .................... 6.5.4 Fuel Cycle Cost ..................................... 6.6 Parameter Studies ......................................... 6.6.1 Processing Cost as Parameter ........................ 6.6.2 Effect of Xenon Removal ............................. 6.6.3 Effect of Product Sale Without Recycle .............. 6.7 Alternative Design and Cost Bases ......................... 6.7.1 Thorex Processing Cost Estimates .................... 6.7.2 Reactor-Integrated Fluoride Volatility Processing ... 6.7.3 Reactor-Integrated Precipitation Process ............ 6.8 Ebolution of a Self-sustaining MSCR ....................... 6.8.1 Reduction of Leakage ................................ 6.8.2 Reduction of Xenon Captures ......................... 6.8.3 Reduction of Fission Product Poisoning .............. 6.8.4 Improvement of Mean Eta and Reduction of 236U Captures ............................................ 6.3.3 Uranium-233 6.304 Uranim-234 6-395 Uranium-235 6.3-6 Uranium-236 6.3.7 Nept-iiium-237 6.3.8 Uranium-238 6.3-9 Neptunium-239 and Plutonium Isotopes 6-3-10Salt 6-3-11Xenon-135 and Related Isotopes 6.3-12 Noble Metal Fission Products 6-3-13Other Fission Products . 6.3.14 Corrosion Products 6.4 Fuel Cycle Optimization 6.5 Reference Design Reactor 6.5.1 Specifications 6.5.2 Neutron Economy

135 135

136 136 136 136 136 136 137 137 137 137

138

141

142 143

147 148 150

150 152 153 154 154

155 156 156 157 157

158 158

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CONTENTS (continued) Page

................. MSCR CAPIWL INVESTMENT. FIXED CHARGES. AND OPERATING EXPENSE ....................................................... 7.1 Introduction .............................................. 6.8.5

7

7.2 7.3 7.4 t .c

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

Summary of N C R Capital Investment MSCR Fixed Charges

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

MSCR Operating and Maintenance Cost Estimate 7.4.1 Labor aEd Materials 7.4.2

..

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

Operation aEd Maintenance Cost RESULTS AND CONCLUSIOXS 8 8 . 1 Fuel Cost 8.2 Fixed Charges

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Ultimate Breeding Potential of MSCR

................................................. ............................................. 8.3 Operation and Maintenance Expense ......................... 8.4 Cost of Power ............................................. 8.5 Breeding Potential of the MSCR ............................ 8.6 Conclusions ............................................... 8.7 Recommendations ........................................... 8.7.1 Title 1 Design Study of MSCR ........................ Conceptual Design Studies of Advanced Breeder 8.7.2 Reactors ............................................ Fundamental Studies of Alternative Chemical 8.7.3 Processes ........................................... 8.7.4 Engineering Laboratory Study of Precipitation Processing .......................................... 8.7.5 Pilot Plant Study of HF Dissolution Process ......... APPENDICES ............... ................................. Appendix A - MULTIGROUP CR CTIONS FOR MSCR CALCULATIONS ......

- EFFECTIVE THOR ONANCE INTEGRAIS ................ Introduction ................................................... Analysis ....................................................... Sample Calculation .............................................

Appendix B

........................................................ ........................................................ References .....................................................

Results Symbols

.

158 160 160 160 162 164

164 165 168 168 169 170 170

171 173 173 173 174

174 174

175 177 179 196

196 197 200

201 201 204

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CON"?TS

(continued)

..................... Summary ........................................................ References ..................................................... Appendix D . THE MERC-1 EQUILIBRIUM REACTOR CODE .................. Introduction ................................................... Theory ......................................................... References ..................................................... Appendix E . FISSION PRODUCT NUCLEAR DATA ......................... Appendix C

.ENERGY

References

DEPENDENCE OF ETA OF 233U

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

Appendix F -TREATMENT OF DEIAYED NEUTRONS

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

........................................................ References ..................................................... Appendix G - TRFATMENT OF XENON ABSORPTION IN GRAPHITE ............ summary

................................................... ....................................................... Reference ...................................................... Appendix H -THE EQUILIaRIUM STATE AS A BASIS FOR ECONOMIC EVALUATION OF' THORIUM REACTORS .................................... Introduction ................................................... Methods ........................................................ 1. Fission Products ........................................ 2. . Uranium-233 ............................................. 3 . Uranium-234 ............................................. 4. Uranium-235 ............................................. 5. Uranium-236 ............................................. 6. Uranium-238 ............................................. Results and Conclusions ........................................ 1. Fission Products ........................................ 2. Uranium-233 ............................................. 3. Uranium-234 ............................................. 4 . Uranium-235 ............................................. 5 . Uranium-236 ............................................. 6. Uranium-238 ............................................. Introduction Analysis

205 205 208

209 209

209 220 221 224 225 225 229

230 230 231 234 235 235

236 236

237 239 239 240 241 242 242 249 249 249 249 249


xiii i

'b CONTENTS (continued) Page Appendix I -ESTIMATES OF PHYSICAL PROPERTIES OF LITHIUMBERYLLIUM MSCR FUEL AND COOLANT SALTS Introduction Viscosity

............................. ................................................... ...................................................... Heat Capacity .................................................. Thermal Conductivity ........................................... Density ........................................................ Liquidus Temperature ........................................... References ..................................................... Appendix J . FUEL AND CARRIER SALT C a T BASES ..................... Introduction ................................................... Bases for Establishing Prices .................................. General Comments on Price Quotations ........................... Thorium Fluoride ............................................. Zirconium Fluoride Beryllium Fluoride

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

............................................. Thorium Oxide ............................................... Lithium Fluoride ............................................ Recommended Values for Molten Salt Fuel ........................ References ..................................................... Sodium Fluoride

Appendix K . B C R POWER LIMITATION RESULTING FROM MODERATOrR THERMAL STRESS

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.................................................... Summary ........................................................ References ............... ................................... Appendix L - VOLUMES OF FUEL SALT AND INTERMEDIATE COOIANT SALT FOR 1000 Mwe MOLTEN SALT CONVERTER REACTOR ........................ Introduction ................................................... Fuel Salt Volume ............................................... Coolant Salt Volume ..... ................................... Appendix M - EVALUATION OF A SALT CONVERTER REACTOR ............................................ Introduction

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

252 252 253 256 256 258 258 260 261 261 269 269 269 269 270 270 270 270 271

272 273 273 279 280 280 280 282 285 285

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xiv CONTENTS (continued)

........................................................ Conclusions .................................................... Appendix N -DETAILED ESTIMAT'E OF 1000 Mwe MSCR CAPITAL INVES........................................................ Sumnary ........................................................ Investment Requirements ........................................ Appendix 0 -DESIGN REQFOR THE M3CR MODERATOR ........... Introduction ................................................... Moderator ...................................................... Void Fraction .................................................. Permeation of Graphite by Salt ................................. Results

Graphite Shrinkage

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

........................................... Differential EsLpansion ......................................... References ..................................................... BIBLIOGRAPHY ...................................................... Graphite Replacement

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286

288 289

289 290 328 328 330 330 331 332

333 335 336 337


xv ACKNOWLEDGMENTS The comments, suggestions, and c r i t i c a l reviews of R. B. Briggs, P. R. Kasten, J. A. Lane, H. G. MacPherson, E. S. B e t t i s , A. M. Perry, and S. E. Beall a r e g r a t e f u l l y acknowledged.

The computer programming

was performed by J. Lucius, ORGDP Computing Center, and many of the drawings were prepared by H. MacColl.

The c a p i t a l c o s t s were estimated by C. A. Hatstat of Sargent and Lundy, Engineers. R. P. Milford and

W. G. Stockdale, ORNL Chemical Technology Division, a s s i s t e d t h e chemic a l processing cost studies. Roy Robertson and I. Spiewak, ORNL Reactor Division, a s s i s t e d the design and analysis of t h e energy conversion systems.

Special studies reported i n the Appendix were performed by

R. H. Chapman, J. W. Miller, and C. W. Nestor of t h e ORNL Reactor D i -

vision, and D. B. Janney of ORGDP.


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MOLTEN SALT CONVERTER REACTOR

Design Study and Power Cost Estimates for a 1000 Mwe Station

L. G. Alexander, W. L. Carter, C. W. Craven, D. B. Janney, T. W. Kerlin, and R. Van Winkle

ABSTRACT The MSCR is a one-region, one-fluid, graphite-moderated converter reactor fueled with a mixture of the fluorides of thorium uranium, lithium-7, and beryllium which is circulated through the 20-ft-dim core to an external heat exchanger. Heat is transferred through an intermediate salt-coolant to steam at 2400 psi, 1000째F in a Loeffler boiler system having a net thermal efficiency of 41.5$. Spent fuel is processed by fluorination (at 0.08 mill/kwhe) for recycle of isotopes of uranium. The stripped salt is discarded. A capital investment of $143/kwe (3.0 millslkwhe), an operation and maintenance annual expense of $2.1 million (0.3 mill/kwhe), and a minimum fuel cycle cost of 0.7 mill/ kwhe (optimum conversion ratio is -0.9) were estimated, giving a net power cost of 4.0 mills/kwhe. All costs were based on 1962 bases ground rules.

Second generation plants may have capital costs as low as $125/kwe. Conversion ratios slightly greater than one can be obtained in advanced designs. This study was completed in December 1962 and does not reflect increased feasibility and superior performance of two-region, two-fluid molten salt breeder reactors made possible by recent (January 1965) advances in core design, heat transfer, and fuel-salt processing.

1. S m Y

The Molten Salt Converter Reacior (MSICR) is a one-region, one-fluid,

near-term reactor that does not require any technology beyond the scale-up of that already developed at ORNL or to be demonstrated in the MSRE.-,Salient characteristics are given in Table 1.1.


2

Table 1.1, Characteristics of the Molten Salt Converter Reactor Thermal capability Net thermal efficiency Diameter and height of core Moderator Volume fraction of fuel in core Composition of fuel carrier salt (mole-percentages) Density of fuel salt Heat capacity of fuel salt Velocity of fuel salt Inlet temperature Outlet temperatwe Flow rate Volume of circulating stream Parer density in core (av) Pawer density in fuel salt (av) Thorium specific power Fissile material specific parer Fertile material exposure Intermediate coolant (mole-percentages)

2500 Mw 41-54 20 x 20 ft Graphite 0.10 68-LiFy 22-BeF2, 9-ThF4, LUF4

190 lb/ft' 0.35 Btu/lb "F

-

6 fps 1100째F 1300째F 160 ft3/sec 2500 ft3 14 w/cm3 35 w/cm3 30 Mwtltonne 0.9 Mwt/kg ,47Mw days/& 63-ZiFy 37-BeF2

Mean neutron productions (92)

2400 psi, 1000째F -300 -30 2.21

Optimum conversion ratio

099

Stem conditions C: Th atom ratio Th: U atom ratio


3 1.1 Description

The reactor v e s s e l i s fabricated of INOR-8 a l l o y and i s f i l l e d w i t h c y l i n d r i c a l graphite logs 8 inches i n diameter and 24 inches long.

The

f u e l , a mixture of the fluorides of 7 L i , Be, Th, and U flows upward through the passages around the logs and i s discharged through eight pumps t o an equal number of heat exchangers where the heat is transferred t o an i n t e r mediate-salt coolant.

Saturated steam i s superheated i n a shell-and-tube

exchanger; p a r t of the steam i s routed t o the turbines; the r e s t i s recirculated t o Loeffler boilers where saturated steam i s generated by inj e c t i n g the superheated steam i n t o water.

Thus, thermal contact of the

coolant s a l t with subcooled, boiling water i s avoided, and thermal s t r e s s i n the tube walls i s tolerable.

The thermal efficiency i s i n excess of

Twenty-five hundred Mw of heat a r e extracted from a single core a t 4%. average power d e n s i t i e s i n the f u e l s a l t of not more than 35 w/cm3. 1.2

Fuel Reprocessinq

I r r a d i a t e d f u e l i s removed from the reactor daily, collected i n t o processing batches, and t r e a t e d with fluorine f o r recovery of isotopes of uranium ( f u l l y decontaminated) as the hexafluoride. discarded.

The stripped s a l t i s

Recovered UF6 i s reduced t o UF4, blended with f r e s h s a l t , and

recycled t o the reactor.

Net burnup and l o s s of f i s s i l e material a r e com-

pensated by addition of 95% enriched 235U. 1.3 Nuclear and Thermal Performance The l i m i t i n g c r i t e r i a ( e . g . , m a x i m u m allowable fuel temperature, maxi-

mum allowable thermal stress i n graphite, e t c . ) were chosen conservatively throughout, and provide consid

l e margin f o r improvement i n l a t e r de-

The key variables (core diameter, volume f r a c t i o n of f u e l i n core, thorium r a t i o , and processing r a t e ) were optimized with respect t o the f u e l cycle cost.

Characteristics of the optimized system are l i s t e d

i n Table 1.1 where it i s seen t h a t the optimum conversion r a t i o i s 0.9, w i t h s l i g h t l y permeable graphite t h a t absorbs 135Xe only slowly.


4 1.4 Fuel Cycle Cost The estimation of inventory and replacement charges for the MSCR is straightforward. Processing costs are less well defined; however, the processing contributes only e, small part of the total fuel cost, and the aggregate is not sensitive to large errors in the processing cost estimates. A central Fluoride Volatility facility capable of processing 30 ft3/

day of salt was designed and costed. Only isotopes of uranium are recovered; carrier salt and thorium a r e discarded along with fission products. Unit costs and the components of the fuel cycle cost are listed in Table 1.2. Table 1.2. Fuel Cycle Cost in 1000 Mwe Molten Salt Converter Reactor Plant Cost Bases Capital investment in processing plant: $26 million Annual operating expense: $2 million "urn-around-time: 2 days Batch size: 6000 kg Unit processing cost: $27/kg Th Shipping costs: $lO/kg Th Purchase price ThF4: $19/kg Th Carrier salt purchase price: $1130/ft3 Fissile isotopes: $l2/gram Charges, mills/kwhre Material m232

pa23 $33 u235

Total Salt

Total Inventory

Replacement

0.033 0.001f

0.043

0.181 0.037 -0.262 0.062

Totai charges, mills/kwhre

Processing

0.082

0- 156

-

-

0.199

0.082

0.079

0.54 0.14 0.7


5

1.5 Power Costs The cost of power was obtained by combining fuel-cycle costs with estimates of capital charges prepared by Sargent and Lundy, Engineers (95,96), from a design study conducted at ORNL.

Equipment was sized and

specified in sufficient detail that costs might be estimated by usual procedures. Plant arrangement drawings were prepared from which costs of buildings, piping, services, etc. were estimated. Operation and maintenance costs were estimated according to standard procedures ( 5 2 ) . A summ w y of the principal items is given in Table 1 . 3 . r

Table 1 . 3 .

1000-Mwe Molten Salt Converter Reactor Construction Costs

Direct construction costs Structures and improvements Reactor plant equipment Turbine-generator units Accessory electric equipment Miscellaneous power plant equipment

$ 5,997,950 51,324,350 26,843,700 4,375,300 799,900

Total direct construction costs Indirect costs

89,341,200

Engineering design and inspection costs

15,080,300

Miscellaneous charges GRAND TOTAL Net station power Unit capital cost

35,370,800

9,083,300

$148,875,600

1038 Mwe

$l43/Kwe

The fixed charges (14.46%)on the capital investment contribute 3.0 mills/kwhre to the power cost. The uncertainty in this cost might run as high as 15-204, and the fixed charges might range up to 3.5 mills/kwhre. Operation and maintenance contribute 0.3 mills/kwhre to the total power cost (Table 1.4). Because of the many uncertainties, this estimate may be low, and the cost might run as high as 0.5 mills/kwhre.


6

Table 1.4. 1000-Mwe Molten Salt Convert= Reactor Operating and Maintenance Cost Wages and salaries Routine materials Maintenance Management Total

$ 872,000 220,000 800,000 262,000 $2,154,000

The various contributions to the cost of power have been swmned in Table 1.5. Table 1.5. Cost of Parer in a 1000-Mwe Molten Salt Converter Reactor

Item

.Fuel cycle cost Fixed charges Operation and maintenance Cost of power, mills/hke

0-7

3.0 0.3 4=0

Taking the upper bound on these three items estimated above (fuel cast -1.0, fixed charges -3.5, operation and maintenance -0.5) gives an upper limit on the cost of power of 5.0 mills/kwhre. 1.6 Advanced WCR

The system evaluated above was based on the scale-up of current technology, and was conservatively designed in every respect. There are several obvious impruvements that could be incorporated into a "second gen-. eration" design. If the design criteria were relaxed, metallic sodium could be substituted for the intermediate salt coolant (saving about $10 million in capital costs. This would also permit the use of "conventional"


i

‘u

7 once-through sodium-heated b o i l e r s and reduce the cost of the energy conversion system by about another $10 million. -$l25/kwe.

The t o t a l cost would then be

By careful design and development the f u e l volume might be r e -

duced from 2500 f t 3 t o 1800 ft3. Separated 92M0 could be used t o clad the graphite and so reduce absorption of xenon therein and a l s o a s a struct u r a l material by means of which a blanket of ThF4 bearing salt could be added a t the periphery of the core t o reduce neutron leakage.

The use of

Fluoride V o l a t i l i t y coupled with the Hl? Solution Process t o remove r a r e earths could reduce the f i s s i o n product poisoning t o very low l e v e l s while permitting recycle of c a r r i e r s a l t (but not thorium).

Preliminary calcu-

l a t i o n s show t h a t these improvements ( a l l within reach of modest development programs) might increase the conversion r a t i o above 1.0, and, w i t h .

C

I

the reduction i n c a p i t a l c o s t s noted, r e s u l t i n a power cost of 3 . 4 m i l l s / kwhre.

1.7 Post S c r i p t

- January

1965

This study was completed i n December 1962, and does not r e f l e c , fncreased f e a s i b i l i t y and superior performance of two-region, two-fluid molten s a l t breeders made possible by the recent advances (January 1965) i n core design, heat transfer, and f u e l - s a l t processing alluded t o i n the Foreword.

T


8 2.

INTRODUCTION

Purpose, Scope, and Methoc, of Approach The purpose of t h i s study was t o evaluate the economic p o t e n t i a l of a near-term molten salt power reactor.

"Near-term" characterizes a system

which u t i l i z e s only techniques or equipment currently under development. 2.1.1

Figure of Merit The economic p o t e n t i a l of power reactors i s measured by the net cost

of e l e c t r i c power.

Fuel cycle cost, although not d e f i n i t i v e , i s a l s o an important index

of economic p o t e n t i a l .

Moreover, the optimization of the f u e l cycle i s a

required f i r s t s t e p i n the d e t a i l e d design of both reactor and e l e c t r i c I n t h i s study, t h e reactor and i t s associated heat t r a n s f e r sys-

plants.

tem, the energy conversion system, and the f u e l reprocessing p l a n t were designed i n d e t a i l s u f f i c i e n t t o permit the optimization of the f u e l cycle. 2.1.2

Reactor Concept A concept w a s selected f o r evaluation, which, judging from previous

experience, would s a t i s f y the "near-term" requirement and y e t would exh i b i t a t t r a c t i v e f u e l costs:

A single-fluid, single-region, graphite-

moderated molten-salt reactor generically r e l a t e d t o the Molten S a l t Reactor Experiment.

Since the breeding r a t i o was expected t o be l e s s than

unity, the system was designated the "Molten S a l t Converter Reactor"

-'*%-

(mcR)* 2.1.3

Procedure I n a s e r i e s of preliminary calculations, the l i m i t a t i o n s on reactor

design imposed by consideration of allowable temperature, pressures, vel o c i t i e s , thermal s t r e s s , etc., were determined.

Design and cost bases

were established, and t h e f u e l cycle cost w a s minimized by optimization of the key variables, which i n the M3CX are the core diameter, carbon/ thorium r a t i o , volume f r a c t i o n of f u e l i n t h e core, and spent f u e l


processing r a t e .

For the optimum conditions, the f i e 1 cycle costs r e s u l t -

ing from a l t e r n a t e bases and assumptions (e.g., removal of xenon) were determined.

Finally, t h e ultimate performance r e s u l t i n g from a concatena-

tior- of a l l favorable assumptions and potentially l o w processing costs w a s estimated. 2.2

2.2.1

Status of Molten S a l t Reactor Development

Early Work Molten s a l t f u e l s were conceived originally as a means of s a t i s f y i n g

the requirements f o r very high temperature and extremely high power density necessary f o r a i r c r a f t propulsion.

A very large amount of work on the

physical, chemical, and engineering c h a r a c t e r i s t i c s of uranium and thorium bearing molten fluorides was carried out as p a r t of the ANP program a t

Oak Ridge National Laboratory. The technology of molten s a l t reactors w a s f i r s t introduced i n t o the open l i t e r a t u r e i n 1957 by Briant and Weinberg (14).

Papers by B e t t i s

e t a l . (6,7) and Ergen e t al. (31) reported the Aircraft Reactor Experiment, a beryllium-moderated reactor fueled with UF4 dissolved i n a mixt u r e of the fluorides of sodium and zirconium, and contained i n Inconel. The reactor was successfully operated i n 1954 f o r about 90,000 kwhr w i t h out incident a t powers up t o 2.5 Mwt and temperatures a s high a s 1650째FThe p o t e n t i a l usefulness

recognized from the start.

of molten s a l t f u e l s far c i v i l i a n power w a s

The f e a t u r e s that a t t r a c t e d a t t e n t i o n were

the high temperature of the f u e l (permitting use of modern steam technology and attainment of high thermal efficiency) combined with a l o w vapor pres-

sure, the high s t a b i l i t y of halide salts under radiation, and t h e advantages t h a t a f l u i d f u e l provides.

These include a negative temperature

coefficient of r e a c t i v i t y , absence of the need f o r i n i t i a l excess react i v i t y and of neutron wastage i n control elements, no l i m i t a t i o n t o fuel exposure due t o r a d i a t i o n damage or f i e 1 burnup, the absence of a complicated s t r u c t u r e i n the reactor core, removal of t h e heat t r a n s f e r operat i o n from the core t o an external heat exchanger, and the p o t e n t i a l f o r a low-cost f u e l cycle.

I n addition, s u i t a b l e molten s a l t mixtures exhibit a


10 solubility for thorium fluoride sufficient for all reactor applications; morewer, these mixtures may be economically and rapidly processed for the recovery of 233U by means of the well-developed Fluoride Volatility Process. Studies of power reactors utilizing molten salts have been reported by Wehmeyer (log), Jarvis (49),Davies (27), and Bulmer (15). Davidson and Robb (26) conceived many of the features of one-region thorium converter reactors and anticipated some of the development problems.

2.2.2 The Molten Salt Reactor Program The molten salt reactor program was inaugurated at ORNL in 1956 ( 5 7 , 5 8 ) to exploit the technology of molten salt fuels for purposes of economic civilian power. Several parts of the program were: (a) a reactor evaluation study to select the most promising concepts for civilian power and to pinpoint specific development problems: (b) an extensive materials development program for fuels, containers, and moderators; ( c ) an equally extensive program for the development of components, especially pumps, valves, and flanges suitable for extended use with mblten salts at 1300"E; (d) a modest program for the discwery of supplementary chemical processes for recovering valuable components (other than uranium) from spent fuel; (e) a program for the development and definitive demonstration of the feasibility of completely remote maintenance of molten salt reactor systems; and presently (f) the Molten Salt Reactor Experiment (MSRE). 2.2.3 Fuel Development The program for the development of molten salt fuels in the Reactor Chemistry Division at ORNL has been highly successful (56). The five-component mixtures (fluorides of Li, Be, Th, U, and Zr) developed for the M3RE (12) have many exceptional features. They have melting points well below 1000째F, with ample solubility for UF4, ThF'4, and fission product

.-.

fluorides. They are thermodynamically stable with vapor pressures less than 0.1 atm at temperatures well above 2000"F, and, being ionic liquids, are not subject to permanent radiation damage ( e . g . , radiolytic dissociation) when in the liquid state. The pasasitic capture cross sections of the base elements (7Li,r'Be, qnd 19F) are satisfactorily low, arid 7Li


11 i s available a t a t t r a c t i v e p r i c e s i n grades containing as l i t t l e a s 0.005$ Li-6.

The high volumetric heat capacities of salt mixtures make them

b e t t e r heat transfer media than most l i q u i d metals i n s p i t e of the higher film conductances obtainable with the l a t t e r . These mixtures do not appreciably attack the container material

(INOR-8),

corrosion rates being l e s s than 1 mil/yeax (possibly as l o w as

1/2 mil/year) a t temperatures below 1300째F (28).

Although it i s not now

anticipated t h a t it w i l l be necessary t o use INOR i n the neutron active zone, since the moderator material (graphite 1 i s suitably s e l f -supporting, experiments have shown t h a t the corrosion i s not appreciably accelerated by radiation.

A long l i f e (10-30 years) is predicted f o r a l l components

constructed of INOR (reactor vessel, pumps, heat exchangers, e t c . ) because resistance t o corrosion does not depend on maintenance of a protect i v e f i l m but stems from the inertness of the base metal toward the s a l t . Molten s a l t f u e l mixtures are compatible with graphite.

Tests of a

t y p i c a l grade show t h a t the s a l t does not wet the graphite and penetrat i o n i s mostly confined t o the surface layers ( 8 4 , p. 93).

Some CF4 has

been observed i n post-irradiation examination of i n - p i l e experiments. Since CF4 i s thermodynamically unstable with respect t o the salt, it i s thought t h a t i t s formation r e s u l t e d from a t t a c k on graphite by f r e e fluor i n e produced by r a d i o l y s i s of s o l i d salt.

Since the f u e l - s a l t must be

maintained i n t h e l i q u i d s t a t e f o r other reasons, f r e e fluorine would not normally be present i n the c i r c u l a t i n g stream. Xenon i s not adsorbed appreciably on graphite (17) a t reactor temperatures, though it w i l l s a t u r a t e the voids present because of i t s extremely low s o l u b i l i t y i n s a l t (107). However, it may be possible t o exclude Xe

,from the graphite by t r e a t i n-g the surface t o close the pores there and render i n t e r i o r pores inaccessible (5).

Purging the salt with a stream

of helium i n the pump bowl or i n a s p e c i a l contactor would then maintain the Xe concentration a t a very low l e v e l (.Section 6 . 8 ) .

Iodine remains

i n the ionic s t a t e and i s not absorbed. Noble metal f i s s i o n products a r e expected t o be reduced by INOR outside the core. The phase behavior of' a great many mixtures has been investigated (108). Proposed mixtures containing up t o 40 mole $ BeF2 have v i s c o s i t i e s


c

3.2 \

adequately l o w and dissolve heavy metal fluorides (W4, ThF4, or ZrF4) in concentrations up to I 5 mole $ with liquidus temperatures less than 1000째F ( 5 6 ) . Additions of 5 1 mole $ of ZrF4 to the base salt satisfactorily re-

w-

duces the sensitivity of the fuel mixture toward precipitation of U02 by oxygenated contaminants (e.g., air, water, lubricating oils) which will be difficult to exclude entirely *om a large reactor system. Graphite is readily de-oxygenated by in situ decomposition of m F - H F vapor, which shows negligible attack on the INOR. Thermophysical properties of the important salt mixtures have been measured (8,24)in detail sufficient to permit reliable calculation of pumping and heat transfer characteristics, which are good. No evidence of the deposition of scale or dendrites in the heat exchangers has been fO m d

-

2.2.4 Container Development The development of nickel-molybdenum base alloys (INOR series) for containment of molten fluorides was conducted jointly by OFWL and International Nickel Company. In addition to the resistance to corrosion mentioned abwe, the alloys have good-to-excellent mechanical and thermal characteristics, (superior to those of many austenitic stainless steels) and are virtually unaffected by long-term exposye to salts or to air at 1300째F (12). The alloy has been made by several major manufacturing companies, and it is presently available on a limited commercial basis in the form of tubing, plates, bars, forgings, and castings. Exhaustive tests

-

at ORNL have sham that its tensile properties, ductility, creep strength,

-

cyclic fatigue strength (both thermal and mechanical) are adequate for molten salt reactor applications when ,-judged in accordance with criteria used in the ASME Boiler Code (75-87). INOR is weldable by conventional techniques using welding rods of the same composition as the base metal. A gold-nickel alloy has been developed at ORNL suitable for remote brazing

of react02 components. INOR begins to soften abwe 2000째F and melts at 2500째F. The thermal conductivity is about 3.2 Btu/hr.ft*"F at 1200PF. NO major difficulties have been encountered in the design and fabrication of reactor components, including pumps and heat exchangers (12)

?

w

5

P


13 2,2.5

Moderator Development Graphite, because of its good moderating properties, low neutron cap-

ture cross section, compatibility with fluoride salts and INOR, and excellent high-temperature physical properties is a superior moderator for molten salt reactors. "he graphite proposed for use in the MSRE has a density of 1.'8 g/cc and a kerosene-accessible porosity of 6%. About half the pore volume is accessible from the surface. However, as mentioned above, molten fluorides do not wet graphite and permeation of MSRE grade graphite by the salt is less than 0.5% by volume at 150 psi ( 8 4 , p. 9 3 ) . The coefficient of permeability by helium at 30째C is loc5 cm2/sec, and Xe will be adsorbed rapidly. However, techniques for reducing permeability

are being developed. Samples of high-density graphite having permeabilik ties at least two orders of magnitude lower have been made (107). Developent of graphites and graphite bodies is being carried out cooperatively with National Carbon Company. Pieces of graphite are presently available in sections up to 20 in. square and 20 ft long. Graphite having outstanding mechanical properties is available in the form of readily machinable rods, tubes, slabs, and spheres. The effects of nuclear radiations on this material are not fully known. The thermal conductivity declines, but probably not below 15 Btu/hr*ft*OF. Thermal stress considerations thus affect the design of moderator elements; the allowable stress is thought to be at least 2000 psi and the allowable strain at least O.l$. These limits appear to be compatible with the thermal and nuclear requirements of optimum core design. However, experimental verification of these values is needed. At the temperatures encountered in molten salt reactors, graphite will shrink during exposure to fast neutrons. Where large gradients in the fast neutron flux exist, the resulting differential shrinkage will result in deformations, or, if these are restrained, in stresses. "he problem of designing a long-lived core structure of large pieces of graphite is presently unresolved. The bowing of graphite stringers might be restrained by use of molybdenum hoops, but this solution may not be suitable for large power reactors.

L,


14 2.2.6

Component Development

Development of components for molten salt reactors has been in progress for over ten years. The most notable achievements to date are:ih demonstration of the long-term reliability of pumps operating at 1300째F, including pumps having molten-salt-lubricated bearings, and the demonstration of the reliability and maintainability of remotely operated freeze flanges and freeze valves. 2.2.7

Reactor Vessel

No difficulties were encountered in the design or fabrication of the reactor vessel for the MSRE.

In large power reactors provision to limit thermal stress by means of thermal shields may be necessary, but mechanical ~

stresses are not important because pressures greater than 200 psi are not encountered anywhere in the systems.

Corrosion does not appear to be a

problem. 2.2.8

Molten Salt Pumps

Molten salt pumps have been operated continuously for 33.months at temperatures above 1200'F. A sump-type pump having one salt-lubricated journal bearing has logged more than 12,000 hours of operation at I225'F, I200 rpm, and 75 gpm. After it was stopped and restarted 82 times, exami;. nation of the bearings disclosed no discernible attack. The use of saltlubricated bearings will enable the shaft to be lengthened so that shielding may be interposed between the pump bowl and the motor with its oillubricated bearings. The impellers of these pumps also withstand attack indefinitely under operating conditions. It is believed that pumps of the types developed can be made in large sizes for use in large molten salt reactor plants and that these can operate at the temperatures required. 2.2.9

Molten Salt Heat &changers

and Steam Boilers

i

The design and fabrication of exchangers for transferring heat from

fuel salt to an intermediate coolant salt are straightforward. Heat


15 t r a n s f e r experiments conducted a t ORNL w i t h unirradiated salt v e r i f y the correlations used t o predict the performance.

Scale d i d not form on the

heat t r a n s f e r surfaces. The Loeffler boiler seems especially suited f o r use with molten salts. Here dry saturated steam i s superheated i n a salt-to-steam exchanger; p a r t of the superheated steam i s routed t o the turbines, and p a r t i s recircul a t e d through an evaporator producing saturated steam f o r recycle t o the exchanger.

Problems i n boiling burnout, thermal s t r e s s i n the exchanger

tubes, and freezing of the s a l t are thas avoided. However, a f u e l - s a l t boiler presently i n the conceptual stage has

I n t h i s concept, the f u e l downcomer annulus inside the reactor v e s s e l i s widened t c accommodate several hurrdred 33oR thimbles. Bayonet tubes, i n t o which water i s introduced, a r e inserted

many p o t e n t i a l advantages.

i n t o the thimbles, but are separated from the thimble walls by a narrow annulus f i l l e d with an i n e r t s a l t .

Calculations show t h a t the heat trans-

f e r i s adequate t o produce steam at 1000째F and 2000 p s i . steam systems a r e i s o l a t e d from d i r e c t contact under negligible pressure.

alnd

Yet the s a l t and

the salt- system i s

Should e i t h e r system leak, t h i s would be de-

tected immediately by monitors i n the i n e r t salt system. Such a b o i l e r has many advantages, including the complete elimination of one cooling loop and i t s associated pumps, heat exchanger, e t c .

In

addition, the f u e l c i r c u i t i s appreciably shortened i n comparrison t o a "spread-out''

system.

The steam produced w i l l be considerably l e s s radio-

active than that produced i n a d i r e c t cycle boiling-water reactor.

2.2.10

Freeze Valves and Freeze Flanges

Although t h e high melting point of a molten s a l t reactor f u e l (800-

1000째F) i s a disadvantag n t h a t the system m u s t be preheated before f i l l i n g and provision must be made t o avoid freezing, there a r e a l s o benef i t s t h a t accr-ue. Among these i s the f a c t t h a t i f a leak does occur there

i s l i t t l e tendency f o r the material t o disperse rapidly. Noble gas f i s sion products do not accumulate i n the liquid, and the fluorides of the remaining f i s s i o n products have negligible vapor pressure and a r e retained.


The ready solidification of salts has also been put to use in the development of flanges and valves. The remote manipulation of reliable freeze flanges has been successfully demonstrated in many tests and in a remote maintenance development facility. Freeze valves have no moving parts, no seals, and have been demonshated to be satisfactory infsalt transfer and drain pipes.

2.2.11 Molten Salt Instrumentation and Special Equipment Conventional equipment is adequate for measuring the nucleax behavior of molten salt reactors; hawever, special equipment for handling molten salts was developed at ORNL for the MSRE. For measuring liquid level in the pump bawls, for example, a ball-float suspending an iron bob whose position is sensed by an external induction coil was developed. A single point electrical probe device has a l s o been developed for use in the filland-drain tanks to calibrate the weighing system. A sampler-enricher device is being tested whreby fresh fuel may be added to the fuel stream during operation, and a sample of spent fuel may be removed without contamination of the fuel stream by air or water vapor and without the uncontrolled escape of any radioactive material fYom the reactor. Clam-shell electrical pipe heaters for lines carrying molten salt have been developed.

2-2-12Remote Maintenance Because of fission-product contamination and induced activity in components and piping, the fuel-containing portions of molten salt reactors cannot be approached for direct maintenance even after draining and flushing. Semi-direct maintenance through a shield plug with longhandled tools is possible for some items, but it is necessary to develop completely remote tools and methods for many of the larger components. These include tools, techniques, and procedures for removing and replacing all major reactor components, including the heat exchanger, primary fuelpump and motor, reactor vessel, and fill-and-drain tank. Such equipment and techniques successfully demonstrated in the Molten Salt Remote Maintenance Development Facility at OFiNL ( 6 5 ) .

This facility simulated a


17 20-Mwt molten salt reactor system and comprised a mockup of the reactor vessel, a mockup of the heat exchanger, together with full-scale pumps, flanges, valves, electrical heaters, thermocouples, etc. A l l maintenance operations were performed by a single operator from a remotely located control center, using closed-circuit stereo-television for viewing. The manipulator was a general purpose, medium duty, electro-mechanical "arm" which performed a variety of functions easily and efficiently. It was used to connect and disconnect tube and electrical connections, to carry loads weighing up to 750 lbs and to manipulate tools. Eight basic mo- . tions, five for the arm and three for the crane bridge, were controlled independently by two pistol-grip handles on the control coosole. Two types of remotely interchangeable grasping devices permitted a variety of objects to be handled. Tools developed for remote manipulation included impact wrenches, a torque tool and bolt runner, screw jacks on the heat exchanger for working the freeze flanges, and miscellaneous devices such as lifting slings, socket extensions, hooks, fingers, etc. A l l these were operated by the manipulator. In addition, a reactor-lifting jig, a pump-lifting eye, and socket extensions for the torque tool and bolt runner were positioned by the manipulator, but operated by the crane or by their own powerThe installation of microphones at strategic locations inside the reactor cell to enable the operator to listen to pneumatic and electric motor sounds was found to be helpful. Reliable, quickly acting disconnects f o r electric, pneumatic, oil, and other services were adapted or developed. The components of the Remote Maintenance Facility were removed and replaced several times before he system was filled with salt in order to develop procedures and test the tools. Finally, the system was filled with salt, brought to temperature with salt circulating freely, then shut down and drained. All equipment was then removed and replaced remotely, and tested. The salt was replaced and brought to temperature again. Items 11

maintained" in this way included the pump motor, the fuel pump, the re-

actor vessel, the heat exchanger, the fill-and-drain tank, electrical pipe heaters, and thermocouples. The demonstration was entirely successful.


18 Maintenance of the MSRE will be accomplished by means of the tech-

-.

WA

niques and tools developed and supplemented with same semi-direct maintenance operations through a portable shield having a rotatable plug. Long-handled tools may be inserted through this plug and manipulated by hand. These means of maintenance w i l l be thoroughly tested in a fullscale mockup of the MSRE now being constructed at ORNL. 2.2.13

Chemical Processing of Molten Salt Fuels

The use of fluid fuels in nuclear reactors provides an opportunity for continuously removing fission products and replacing fissile isotopes at power. Thus, it is possible to hold fission-product neutron losses to low levels and to eliminate capture of neutrons in control rods. The "Fluoride Volatility Process" is in an advanced stage of development; a pilot plant for general application is now in operation at ORNL. Other processes are being sought, and prospects are good that simple and economic means can be found to separate fission products continuously from spent fuel salt.

2.2.14

Fluoride Volatility and HF Solution Processes

While the fluoride volatility process was not developed specifically for use with molten salt fuels, it has been verified in laboratory experiments conducted at ORNL that it is applicable for removal of uranium from fluoride mixtures containing W 4 (16). In this process, elemental fluorine, diluted with an inert gas, is bubbled through the salt. UF4 is

-a

converted to U F 6 which is volatile at the temperature of operation (500700째C) and passes out of the contactor to be absorbed reversibly in a bed

of sodium fluoride. The off-gas is cooled, stripped of F2 in a scrubber, and passed through charcoal beds where fission product gases are absorbed. The fluorides of a few of the fission products are also volatile but these are irreversibly absorbed in the sodium fluoride beds. Thus, by heating the beds, u F 6 is brought Over in a very pure state, completely decontaminated and with losses less than 0.1%. The UF6 is reduced to UF4 in a hydrogen-fluorine flame, and is collected as a powder in a cyclone separator backed up by gas filters. Losses

P

ci

=sb


19

routinely are smaller than random errors in the assays, and the process has been used successfully for many years in the manufacture of enriched 235U from natural uranium in the production plants at Oak Ridge.

The Fluoride Volatility FYocess alone is seficient for the economical operation of a malten salt converter reactor. Spent fuel containing UF4, ThF4, 233Pa, as well as fission products is removed from the reactor periodically and fluorinated for recovery of uranium isotopes. The

stripped salt is discarded (stored in INOR cylinders indefinitely) to purge the system of fission products. Although the discwded salt contains valuable components (7Li, Be, 232Th, 233Pa), the cost of discarding these is'offsetby the'.imprwementin conversion satid The steps described above appear to be especially attractive for integration with the reactor plant. That is, they are all high-temperature, non-aqueous processes, and could conveniently be carried out in the reactor cell, utilizing the same shielding and sharing in the use of remote maintenance equipment. The waste product (fuel salt stripped of itotopes of uranium) is in a form conveniently stored for decay of radioactivity. After a period measured in years, the waste codd conveniently be removed to another location for recovery of thorium, lithium, beryllium, and other valuable components in a relatively low-level-radiation facility

.

The HF Solution Process (16) under study at ORNL provides one means of separating rare earths (which constitute the bulk of important nonvolatile fission products, including isotopes of samarium) from the base salt, after uranium has been removed. The separation is effected by dissolving solidified salt in liquid HF containing up to 1% water. The rare earths, thorium, and related materials precipitate and may be separated by filtration or decantation,.permitting reuse of the salt. The HF Solution Process is presently in the laboratory stage of development.

2.2.15

Thorex Process

While the Fluoride Volatility process appears attractive if integrated with the reactor plant, it i s not obvious that it is superior in a central facility to alternative modes of processing, such as Thorex.


20 This uncertainty is due in part to paucity of reliable information on costs of on-site and central Fluoride Volatility process plants, and in part to the limitations of the method in respect to recovery of lithium and thorium. On the other hand, the costs of Thorex pknts are rather better known, and, with suitable modifications, Thorex appears to permit economic recovery of all valuable components of the fuel salt only moderately contaminated with certain fission products (e.g., cesium). "he costs associated with a modified Thorex process as described in Section 5.3 were used in an alternate evaluation of the MSCR.

2-2-16Fractional Crystallization Process Studies by Ward et al. (108, lw, 80,,p. 80).pravide a basis for evaluating the feasibility of removing rare earth fluorides from the fuel salt by partial freezing. A brief description is given in Section 6.7030 The process is not suitable for a breeder reactor inasmuch as the fission product concentration cannot be lowered much below 0.2 mole $; however, much higher concentrations can be tolerated in a converter. In the reference design studied here, the concentration is approximately 0.5 mole $-

2.2.17 Other Processes Solvents which will selectively dissolve either 'I"4 or rare earth fluorides are being sought at CIRNL. Solutions of SbF5 in HE' show some promise. The capture of a neutron by an atom of 233Pa results in a double loss - that of the neutron and of the fissile atom of 233U that would have been formed by decay of the Pa. A process is needed that can quickly and economically remove 233Pa from the circulating salt stream SO that it may be held outside the reactor until it decays to 233U. There i s a POSsibility that exposing the fertile stream to beds of Tho2 pellets might accomplish this. There is some evidence that thorium from the beds will exchange with Pa in the solution, and the latter will be immobilized until it decays, after which it might, as 233U, exchange with thorium in the salt, and so become available for recovery by fluorination. Other oxides, e-g., BeO, are also under study.


c

21

2.2.18

Molten S a l t Reactor Studies

The s t a t u s of the Molten S a l t Reactor Program was reviewed i n 1958 f o r t h e second Geneva Conference by MacPherson e t a l . (56).

A t that

time a homogeneous molten s a l t reactor having only a limited capability for f u e l regeneration was under consideration.

Further studies of t h i s

system were reported by Alexander e t a l . (l), and a 30-Mwt experimental reactor was described (2).

Also, i n 1958, good indications were obtained t h a t the system INORgraphite-salt i s chemically s t a b l e i n radiation f i e l d s and a t t e n t i o n was accordingly s h i f t e d t o graphite-moderated systems.

MacPherson e t a l .

(60) described a one-region single-fluid reactor u t i l i z i n g s l i g h t l y en-

riched uranium and a highly enriched feed.

Many f e a t u r e s of h i s concept

were incorporated i n the present study. The p o t e n t i a l of graphite-moderated molten-salt reactors f o r breeding i n t h e Th-233U cycle was investigated and the associated development problems were i d e n t i f i e d by MacPherson i n a s e r i e s of papers (61-63)Several conceptual designsfor one- andtwo-region breeders were proposed. One of these ( t h e I’GBR) was evaluated i n comparison w i t h four other thermal breeders by the Thorium Breeder Reactor Evaluation Group a t ORNL (3); t h i s system employed a f u e l s a l t (contained i n graphite bayonet tubes and

circulated through external heat exchangers) together with a f e r t i l e s a l t stream (containing a l l the thorium) surrounding the moderated core region. The major problems associated w i t h t h i s concept were the development of a

r e l i a b l e graphite-metal j o i n t f o r connecting the bayonet tubes t o an INCJR

T

header and the uncertain behavior of t h e core structure f o r long periods under i r r a d i a t i o n a t high power densities. It was estimated t h a t the MSBR could achieve f u e l yields up t o about 7$/year

(doubling time about 14 years) a t f u e l cycle costs not greater

than 1.5 mills/kwhr; and t h a t f u e l c o s t s as l o w as 0.7 mills/kwhr could be achieved by s a c r i f i c i n g the f u e l yield i n favor of lower processing costs (3).

.

bd


22 2.3

Molten S a l t Reactor Experiment

The favlorable r e s u l t s obtained i n the various evaluation and development programs l e d t o the i n i t i a t i o n i n May 1960 of preliminary design of t h e Molten S a l t Reactor Experiment (12,5).

Construction and i n s t a l l a t i o n

of the e n t i r e system are scheduled f o r campletion i n mid-1964 and c r i t i c a l i t y late i n 1964, or e a r l y 1965. The MSRE i s expected t o demonstrate the long-term r e l i a b i l i t y of components and the compatibility of materiais under a c t u a l operating conditions, including t h e dimensional s t a b i l i t y of the graphite and i t s r e sistance t o permeation by f u e l salt i n the presence of radiations and the maintainability of the system after operatiozl a t power. The reactor w i l l produce up t o 10 megawatts of heat i n a f u e l consisting of a solution of highly enriched 235U'F4 dissolved i n a m i x t u r e

of the f l u o r i d e s of lithium (99.990$ 7 L i ) , bery$lium, ,and zbconium

having a liquidus temperature of 842째F. The s a l t enters a volute around the upper p a r t of the c y l i n d r i c a l v e s s e l a t 1175째F and flows a t the r a t e of 1200 gpm down through an annular plenum between the wall of the v e s s e l and up the graphite core-matrix.

This i s constructed by pinning E-5n;

square b a r s loosely t o INOR beams l y i n g across the bottom of the vessel. The s a l t flows up among t h e bars a t a v e l o c i t y of 0.7 f t / s e c .

(Reynolds

number 1000) and e x i t s a t 1225째F. The f u e l pump, a sumptype having a bowl 36 i n . i n diameter and

l2 i n . high, i s driven by a 75 hp motor and develops a head of 48.5 f t a t 1200 gpn.

A l l p a r t s a r e constructed of INaR.

The heat exchanger, a l s o constructed of INOR, has 165 tubes 14 f t long by 1/2 i n . OD with w a l l s 0.042 in. thick, and provides 259 f k 2 of heat t r a n s f e r surface (heat flux 130,000 Btu/hr*ft2 a t a IMTD of 133째F). The reactor heat i s transferred t o a secondary salt coolant from whence

it i s discharged t o t h e atmosphere i n an air-cooled r a d i a t o r . I n i t i a l l y , the E R E w i l l contain no thorium, since t h e power l e v e l i s too l o w f o r s i g n i f i c a n t amounts of 233U t o be produced i n a reasonable time.

Thorium may be added later t o permit v e r i f i c a t i o n of nuclear calcu-

l a t i o n s of c r i t i c a l mass, etc., and t o discover i f there a r e any unforeseen compatibility or s t a b i l i t y problems.


23 BASES AND ASSUl@TIONS

3,

3.1 Design Bases 3.1.1 Reactor Concept

The concept selected for study was mQdeled closely after that proposed by MacPherson et al. ( 6 0 ) , and is essentially a scale'up) of the 9

Molten Salt Reactor Experiment (12,5) plug necessary auxiliary equipment for generation of electricity, etc. Briefly, the core consists of a vertical bundle of unclad graphite logs contained in an INOR vessel. Fuel salt containing thorium and uranium flows up through the bundle into a plenum, thence through several pumps in parallel to the shell side of multiple shell-and-tube heat exchangers, and then back t o the reactor. 3.1.2

Design Calculations These were performed only in sufficient detail to permit the estima-

tion of the capital cost. Problems of control, shielding, hazards analysis, etc., were ignored. Attention was centered on the nuclear performance and processing costs. The energy conversion system was designed to provide a basis for estimating the volume of the fuel salt circulating in the primary heat system, the net thermal efficiency, and the capital investment. 3.1.3 Station Power

An electrical capability of 1000 Mw was selected to permit direct comparison with systems previously evaluated at the same plant capacity. Preliminary calculations indicated that the core should be -20 ft in diameter for satisfactory nuclear performance. At a power of 1000 &e, power densities are only 14 kw/liter of core and 35 kw/liter of salt (average).

A lower plant output would result in inefficient utilization of the fuel inventory.


24 3-1.4 Plant Utilization Factor The standard factor of 0.8 was used as recammended in the "Guide"

(52). 3.1.5 c'

Thermal Efficiency Several different energy conversion schemes were considered in suf-

ficient detail (see Section 4.3) to show that even the least efficient system (Loeffler boiler) would have, when f u l l y optimized, a thermal efficiency not less than 40$. This efficiency was therefore adopted for use in the fuel cost optimization calculations. 3.1.6 Fueling Cycle For the purposes of optimization calculations, it was assumed that make-up fuel was added and spent fuel was removed quasi-continuously, and that, with three exceptions, the concentrations of the various nuclides in the circulating salt system were in equilibrium with respect to feed rates, nuclear reactions, and processing rates. The exceptions were 234U and 238U (which are initially present in amounts substantially lower than the equilibrium value, and whose concentrations increase with time) and 236U (the concentration of which starts at zero and reaches only about 3/4 of its equilibrium values in 30 years). For these three isotopes, concentrations that approximated the average Over a life of

30 years starting with the reactor charged with 235U (95% enrichment) were used. Other important isotopes appear to approach their equilibrium concentrations in times short compared to the reactor life. "he use of equilibrium concentrations for these, especially for slowly equilibrating fission products, is discussed in Appendix H. 3.1.7 Processing The processing rate was optimized with respect to the fuel cycle cost. In the selected process, spent fuel is accumulated, shipped to a central Fluoride Volatility Plant, cooled for a minimum of 90 days, and treated for recovery of uranium. Undecayed 233Pa, along with 232!L'h, 7Li, and 'Be are lost in the waste.


25

3-1.8 Feed and Recycle

In the optimization calculations, it was assumed that isotopes of uranium recovered from irradiated fuel are recycled, and that deficiencies in the breeding ratio are compensated by additions of 95$-enriched 235u. The effects of a few feed and recycle schemes on the optimum reactor were studied (Section 6.9), such as the use of feeds containing a mixture of uranium isotopes (e.g., spent Pael from the Consolidated Edison Reactor at Indian Point, New York). The sale of irradiated fuel to the AEC as an alternate to recycle was also investigated. 3.1.9 Isotopic Composition of Lithium It was assumed that lithium (as the hydroxide) would be available in grades containing up to 99.995% 7Li at a price no greater than that quoted in reference 67 ($120/kg of lithium). The choice of this composition (rather than one having a lower cost) resulted from a compromise between cost of neutron losses to 6Li and the cost of discarding the salt enriched in 7Li with a processing rate of about 2 ft3/day. 3.1.10

Energy Conversion System

Although it would be difficult to establish a complete set of requirements for coupling of the reactor system with the energy conversion system prior to the preparation of a detailed design, nevertheless, it is necessaxy t o f i x some of these in order that the f u e l cycle cost may be estimated.

The most important requirement appears to be a necessity to iso-

late the fuel salt from the thermodynamic fluid, at least when that fluid is water. The hazards associated with the possibility that high pressure I’ steam might leak into the fuel system cannot be tolerated, since such leakage would result in the rapid formation of U02 (81, p. 63). This is only slightly soluble in the base salt, although its solubility can be increased somewhat by additions of ZrF4 and of ThF4 ( 8 4 , p. 96). Isolation of the steam and fuel systems is achieved by interposing a compatible third fluid, either as a stagnant layer or as a separate stream circulated between primary and secondary heat exchangers.


26 The intermediate coolant ( t h i r d f l u i d ) must be chemically compatible with f u e l salt, and i n addition, it i s desirable that it be i n e r t with

Also, it should e i t h e r not be a nuclear poison, or e l s e it should be r e a d i l y removable from f u e l salt. For the reference

respect t o steam.

design, a s a l t 66 mole $ LiF (99.9959 7 L i ) and 34 mole $ BeF2 was selected (Table 3.4) as the intermediate f l u i d . 3.1-11 Primary Heat Exchanger Requirements

It i s important t h a t the external portion of the f u e l salt c i r c u l a t ing system s h a l l have as small a volume as possible i n order t o reduce the inventory of valuable materials. However, the r e l i a b i l i t y and maint a i n a b i l i t y of the system cannot be compromised i n favor of s m a l l volume. A requirement f o r maintainability, which includes replaceability, implies

t h a t the primasy heat exchanger s h a l l be drainable of f u e l salt.

This re-

quirement i s most easily and c e r t a i n l y met by putting the f u e l s a l t i n the shell-sides of the heat exchangers and grouping these about the react o r i n a v e r t i c a l position so t h a t the heads may be removed and t h e tube bundles l i f t e d out easilS;. 3.1.12

Minimum S a l t Temperatures

To provide a margin of safety i n regard t o possible freezing of both f u e l s a l t and intermediate coolant salt, it w a s decided t h a t the operating temperature of any salt stream should not be a t a temperature l e s s than t h e liquidus temperature of the f u e l salt. 3.2 3.2.1

Cost Bases

Value of F i s s i l e Isotopes Unirradiated, highly enriched 235U w a s valued a t $ U . O l / g r a m

tained 235U (52).

of con-

Mixtures of isotopes were valued according t o the

formula V = f ( E ) $=/gram

of contained f i s s i l e isotope (233U, 235U), where

f ( E ) i s an enrichment f a c t o r found by dividing the value of enriched 235u

having the same composition as the mixture i n question by $12.0l/gram. The enrichment, E, of the mixture is found by dividing the sum of the


1

27 atomic concentrations of 235U and 233U (and 233Pa, if any) by the Sum atomic concentrations of all isotopes of uranium in the mixture (thus lumping 234U and 236U with 238U as diluents).

Of

3.2.2 Value of Thorium Inquiries directed to several vendors elicited only one reply (Appendix J); however, the quotation given ($6/lb of ThF4) agreed well with a 1959 estimate by Orrosion (89) and led to the adoption of a price of $19.00/kg of thorium as ThFc, ($6.5O/lb ThFt,).

3.2.3 Value of LiF(99.9955 7Li)'This was taken to be $l20/kg of contained lithium (Appendix J) or $32.30/kg of LiF.

3.2.4 Value of BeF2 Inquiries cited in Appendix J led to adoption of a price of $15.40/kg of BeF2. 3.2.5

Value of Base Salt This varied with the composition, but the base salt in the optimum

reactor contained 68 moles of LiF per 23 moles of BeF2 giving a value of $25.97/kg. 3.2.6 Cost of Compounding and Purifying Fuel Salt

The operation of blending recycle uranium with make-up uraniurp and fresh lithium, beryllium, and thorium fluoride and purifying is to be performed on-site. The cost was therefore excluded from the operating and capital chazges of the processing plant and included in the capital and operating charges of the reactor plant.

3.2.7 INOR-8 Cost The following cost information supplied by A. Taboada of CJRNL is

4iJ

based on quantity production. Manufacturing experience to date with fabrication of the listed forms has not indicated the existence of any


28

n

WL serious problems and therefore pricing safety f a c t o r s i n the costs shown may be pessimistic. Plate

$3 per l b

Round Rod Welding Rod

$4.25 per l b $8 per l b

Pipe (Seamless)

$10 per l b

Pipe (Welded)

$5 per l b

Tubing (Seamless)

$12 per l b

Tubing (Welded) Simple Forgings

$6 Per l b $4.50 per lb (e.g., tube sheets)

Fabricated Plate

$10 per l b (e.g.,

Dished Heads

$5.50 per l b

Forged Pipe F i t t i n g s

$50 per l b

Castings

$2 per l b

pressure vessel shells3

Moderator Graphite Cost

3.2.8

The cost of graphite such as would be used i n the MSCR core bas been established a t $6.00 per l b . dor s

This i s from informal discussion w i t h ven-

.

3.2.9

A n n u a l Fixed Charges For f i s s i l e isotopes, the use charge was taken a t 4.75$/yr

cordance with the "Guide" (52).

i n ac-

Other components of the f u e l mixture were 4

carried as depreciating a s s e t s (since only the isotopes of uranium and thorium a r e recoverable).

For such the "Guide" recornends (Table 3.1) an

annual r a t e of 14.464 f o r an investor-owned public u t i l i t y (IOPU).

This

r a t e , however, includes 0.354 f o r interim replacement when the r a t e of r e placement i s not known.

I n t h e present instance, the replacement r a t e s

f o r base s a l t were calculated and the corresponding c o s t s l i s t e d separ a t e l y ; therefore, the annual charge for the above items w a s s e t a t

14.-U$/yr. This included a l s o l.ll$/yr

~

.

f o r amortization by means of a

30-yr sinking fund with cost of money a t 6.75$~/yr; hence, a charge f o r

replacement of salt a t the end of 30 years was not made e i t h e r separately or as p a r t of the f i n a l processing t o recover the uranium inventory.

i - - <

u

L


29 .-

3 - 2 - 1 0 Central Fluoride Volatility Plant Processing Charges

The schedule given below was extracted from the estimates presented in Table 5.9 and apply to a plant capable of processing 30 ft3 of salt per day (about lOOO/kg day of thorium for the reference design salt) for recovery of isotopes of uranium. The barren salt is discarded. Capital investment ($25.5 million) was estimated by scaling from a study by Carter, Milford, and Stockdale ( 2 1 ) of two smaller on-site plants (1.2 and 12 ft3/day), and adding costs of other facilities required in a central plant (receiving, outside utilities, land improvements, etc. ).

The

plant is large enough to service about fifteen 1000 Mwe molten salt converter reactor plants. A turn-around-time of two days was allowed. Shipping charges ($10.30/kg thorium) were estimated separately (Table 5 . 8 )

-

Table 3.1. MSCR Reference Design One Ton/Day Central Fluoride Volatility Plant Cost Schedule

Production Rate kg/day of Thorium from Reactor

Processing Cost* $/kg Thorium

320 160 80 40 53.3 40 26.7

23.0 24.0 25.3 26.1

26.6 27.6 30.0

*hcluding shipping. 3.3

Special Assumptions

3.3.1 Permeation of Graphite by Salt Tests with MSRE fie1 salt at 1300째F and 150 psi in MSRE graphite showed penetrations of the order of 0.02% in 100 hours (86, p. 9 3 ) . Most

of the absorbed salt was contained in pockets lying at the surface of the graphite, and presumably in cammunication with bulk liquid. From a metallographic examination of thin sections, it was concluded that penetrations c 1


t

30 considerably less than 0.12s would be encountered i n t h e MSRE a t the

maximum pressure of 65 psia.

For t h e purposes of evaluating the "R,

it was assumed the penetration would be O.l$, and t h a t only pores lying a t the surface would contain salt. Thus, i n a core 90 volume $ graphite the volume of salt absorbed i n the graphite would be s l i g h t l y less than l$of the volume of salt i n the core. This absorbed salt was assumed t o have the same composition as the c i r c u l a t i n g stream. 3.3.2

Permeation of Graphite by 135Xe.nont The s o l u b i l i t y of xenon and other noble f i s s i o n product gases i n f u e l

salt i s very low (107); also, t h e i r adsorption on graphite a t 1200째F appears t o be negligible (17).

f

However, there remains the p o s s i b i l i t y t h a t *

gaseous xenon may d i f f u s e i n t o t h e pores i n graphite a t a rate large cornpared t o that at which it can be removed from the s a l t by sparging or spraying. The mathematical treatment of the case a t hand has been presented by Watson, e t . a l . (107), who a l s o established probable ranges f o r the diffusion coefficient.

For the purposes of a reference calcula-

t i o n having a reasonable degree of p l a u s i b i l i t y , a value of

(cm2/ s e e ) w a s selected f o r the diffusion c o e f f i c i e n t and a value of 0.01 f o r the porosity of graphite t o noble gases.

Further, it w a s assuuied t o be

f e a s i b l e t o by-pass 1%of the f u e l s a l t (16 f t 3 / s e c ) through the pump bowls or through a sparge chamber, and t h a t t h i s by-pass steam would give up s u b s t a n t i a l l y a l l of i t s xenon t o the sweep gas. i

3.3.3

Corrosion Products Tests i n a forced corxection INCX loop using a s a l t (62-LiF, 36.5-

BeF2, 0.5-UF4, 1.0-ThF4) very s i m i l a r (except f o r thorium content) t o t h a t proposed f o r the MSCR, show t h a t after a period of i n i t i a l a t t a c k (occurring generally i n t h e equipment i n which t h e batch of salt i s prepared) t h e concentration of structural-element cations reaches equilibrium values (84, p. 79).

The temperature of the salt was 3300째F i n the hot

leg, llOO째F i n t h e cold leg, and w a s circulated f o r a t o t a l of almost 15,000 hours.

The concentration of nickel, a f t e r r i s i n g t o a maximum of

80 p p i n about a thousand hours, reached an equilibrium value of about

r?

u


5

31

'c)

*

50 ppm at 2000 hours. Chromium concentration fluctuated between 400 and 600 ppm, averaging about 500 ppm, while iron averaged about 250 ppm. Molybdenum was said to be negligible and was not reported. Apparently the concentration of chromium is in equilibrium with respect to the rate with which chromium is oxidized by UF4 to CrF2 at the hot metal surfaces and the rate with which it is reduced to O C r at the cold surfaces (75, p. 39). In the MSCR, large areas of INOR are exposed to the salt at all temperatures between 1100째F and 1300째F. Although the rate of diffusion of chromium in INOR has been determined at various teniperatures, it is not possible to calculate the chromium concentration in the salt until the temperature profile is known. In the calculations performed here, a neutron-poison allowance was made for corrosion products, amovnting to 0.008 neutrons per atom of fuel destroyed. "his loss is comparable to the loss that would result if the concentrations of Ni, C r , and Fe were 50, 500, and 250 ppm, as in the loop-corrosion test cited above (Section 6.3). 3.3.4

Approach to Equilibrium

The nuclear performance was calculated by means of MERC-1, an equilibrium reactor code. Thus the performance of the reactor during the approach to equilibrium, when concentrations of isotopes of uranium and

of fission products are changing, was not considered, except in regard to 234U, 236U, and 238U. These were averaged over a fuel lifetime of 30 years; 233U, 235U, and fission products were taken at their equilibrium values. In cases where adequa supplies of 233U are unavailable the reactor would be fueled initially with enriched 235U. This is inferior to 233U in respect to eta and also forms a non-fertile daughter, 236U. These disadvantages are offset by initially-low concentrations of fission products and 236U. While a calculation of the time-dependent behavior is J

desirable in such cases, it does not appear that the error introduced by assuming equilibrium conditions is important. The matter is explored further in Appendix H.


32

4. DESCRIPTION OF MSCR CONCEPT 4.1 General Description The MSCR i s a single-region, unreflected, graphite-moderated f l u i d f u e l reactor u t i l i z i n g a mixture of molten fluorides of lithium, beryll i u m , thorium, and uranium as the f i e 1 and primary coolant. A sketch of the reference design reactor i s shown i n Fig. 4.1. As seen i n t h i s f i g -

ure, the reactor consists of a 20-ft-diam by 20-ft-high c y l i n d r i c a l core made up of 8-in.-diam graphite cylinders. The f u e l s a l t enters through a bottom grid, flows upward through the spaces between the cylinders and

i s discharged i n t o one of eight primary heat removal c i r c u i t s located around the reactor. The arrangement of these c i r c u i t s i s shown i n Fig. 4.2. The heat generated i n f u e l salt i s transferred t o an intermediate coolant s a l t consisting of a mixture of barren lithium and beryllium fluoride containing no uranium or thorium.

The coolant salt i s used t o

superheat saturated steam produced i n a Loeffler b o i l e r and also t o r e heat steam f r a m the turbogenerators.

"he reactor vessel, i n t e r n a l s and

a l l primary and secondary system components i n contact with f u e l s a l t and coolant salt a r e constructed of INOR-8. i n section 4.10.

The specifications a r e tabulated

Part of the superheated steam i s sent t o a high-pressure

turbine and the rest i s injected i n t o t h e Loeffler b o i l e r s t o generate saturated steam.

This saturated steam i s recirculated t o the superheater

by steam-driven axial compressors using steam drawn f r o m the high-pressure turbine discharge.

A flowsheet of t h i s heat removal-power generation sys-

tem i s shuwn i n Fig. 4.3. Design data and operating c h a r a c t e r i s t i c s f o r the reference design

are given i n Table 4.1.

4.2

S i t e Plan,.

The s i t e plan of t h e MSCR p l a n t i s shown i n Fig. 4.4 based on cond i t i o n s specified i n the AEC Cost Evaluation Guide (52).

The 1200-acre

grass-covered s i t e has l e v e l t e r r a i n and i s located on the bank of a river.

Grade l e v e l of the s i t e i s 40 f t above the r i v e r low water l e v e l


33

ORNLDWG. 63-2801Rl


ORNLDWG. 65-7906

'\ REACTOR VESSEL /

PRIMARY HEAT EXCHANGER

1

I('

I

I

I I

COOLANT PUMP

I

I

I I I

,'

COOLANT SALT LINE c

SHIELD

Fig. 4.2.

MSCR Heat Transfer Flow Diagram.


_ I -

Ik I

\

?I

-


0

>

3

A

I

7

c

W

a\

w


37

Table 4.1. Design Characteristics of the 1000 Mwe Molten Salt Converter Reference Reactor General Thermal power Net thermal efficiency Net electrical power Core geometry Moderator Form Dimensions Weight Volume fraction in core Porosity accessible to salt (assumed) Porosity accessible to gas (assumed) Gas diffusion coefficient (assumed) Graphite density Radiation heating (ma. ) Maximum temperature rise Reactor vessel Inside diameter Thickness Maximum temperature Weight, including internals Radiation heating in support plates Radiation heating in vessel wall Maximum temperature rise in wall Fuel stream

, ! ! . ? -

2

I

j

Composition Base salt ( LiF-BeF2 -ThF4 ) U F ~(fissile) Fission products Corrosion products

2500 Mw 41.5% 1038 Mw

Cylindrical, 20 ft x 20 ft Unclad graphite CylindErs 8 in. diam, 24 in. long 335 tons 0.9 0.1% 1.0% cm2isec 10-6 1.9 g/cm

5.2 watts/cm3 520°F

INOR-8 20 ft-2 in. 1.7 in. 1400°F 1 2 5 tons 2 watts/cm3 0.6 watts/cm3 40°F

68-22-9mole $ 0.3 mole % 0.5 mole '$ 750 ppm

-

Liquidus temperature of base salt Density.ofbase salt at 1200°F Mean heat capacity of base salt at 1200°F Fraction of core occupied by fuel salt Fuel stream inlet temperature Fuel stre& outlet temperature Flaw rate Velocity in channels in core (avg. ) Velocity in piping Velocity in heat exchanger Shell side Tube side

887OF 3.045 g/cc 0.383 Btu/lb OF

0.1 ll0O"F 1300°F 160 ft3/sec 6 ft/sec 35 ft/sec 20 ft/sec 31 ft/sec


38 Table 4.1

(continued]

Fuel Stream, (continued) Pressure, p s i a Pump discharge Heat exchanger i n l e t Heat exchanger o u t l e t Reactor i n l e t Reactor o u t l e t Pump suction

190

185 95 80 35 22.. 5

Parer density i n f u e l s a l t I n core ( m a x . ) Average Over e n t i r e f u e l volume

510 w/cc 35 w/cc

Volume of f u e l salt In In In In In In

a c t i v e core top and bottom plena f u e l annulus adjacent t o vessel wall surge tank pumps heat exchangers I n connective piping I n dump tanks and r e a c t i v i t y control tanks

630 540 105 85 130 575 320

ft3 ft’ ft3 ft3

ft’ ft’ ft’ 115 ft’

2500 f t 3

TOTAL

Volume of f u e l i n active core Primary heat t r a n s f e r loop Primary pumps; number and type Pressure a t pump discharge Primary heat exchangers Total heat transfer area Average heat f l u x Material Weight

650 ft3

-

8 S a l t Lubricated 200 ’psi 8 S h e l l and Tube 53,000 f t 2 160,000 Btu/hr*ft2 INOR-8 36,000 l b each

-

Secondary heat t r a n s f e r loop Coolant Coolant Coolant Coolant Coolant Coolant

salt salt salt salt salt salt

composition (mole $) i n l e t temperature outer temperature flow rate pump discharge pressure volume

66-LiF; 34-BeF2 950°F U.00 OF 203 ft3/sec 350 p s i 5600 f t 3


39 Table 4.1 (continued) Energy conversion loop Superheaters Materials Heat flux Heat transfer area Weight Inlet steam temperature Steam flow rate Reheaters Materials Heat flux Heat transfer area Weight (approx.) Inlet steam temperature Steam flow rate $oeffler boilers Length Diameter (ID) Weight (approx ) Inlet steam conditions Steam flow rate Inlet feedwater conditions Feedwater flat rate Discharge steam conditions

.

Steam circulators Flow rate Power Steam temperature Steam pressure Turbine Flat rate Generator output -Steam temperature Steam pressure Exhaust

-

16 U-Shell and U-Tube INOR-8 and alloy steel 52,440 Btu/hr ft2 8,850 ft2 -50,000 lbs 670째F 1.3 X lo6 lb/hr 8 - Shell and U-Tube INOR-8 and alloy steel 37,250 Btu/hr ft2 3,543 ft2 22,000 lbs 635OF 0.7 X lo6 lb/hr

4 100 ft 6 ft 600,000 l b s 2430 psia/100O0F 3.1 x lo6 lb/hr 2520 psia/545OF 2.0 X lo6 lb/W 2400 psia/662"F (sat4 4 (turbine driven) 20.5 X lo6 lb/hr 5,100 BHP

670'F 2,480 psia 1 (CCGF-RH) 8.04 x lo6 lb/hr 1083 Mw l O O O / l O O O ~F 2400/545 psia 1.5 in. Hg

Processing system Processing method Salt processing rate Production rate Cooling time ( average )

W t

Central Fluoride Volatility 1.67 ft3/day 53.3 kg Thor ium/day -90 days


40 Table 4.1

(continued)

Processing sys tem (continued) Hold-up time ( t o t a l ) Processing batch s i z e Processing plant capacity Turn-around time

126 days -6,000 kg Thorium 1,000 kg/day 2 days

and 20 f t above the high water l e v e l .

An adequate source of r a w water

f o r the ultimate s t a t i o n capacity i s assumed t o be provided by the r i v e r w i t h an average maximum temperature of 75째F and an average minimum tem-

perature of

LOOF.' 4.3

Structures

Plan views of the reactor and turbine building are shown i n Figs. 4 - 5 and 4.6 and v e r t i c a l sections i n Fig. 4.7. As seen i n these figures, the reactor building and turbine building are addscent, t h e secondary shield w a l l forming a separation from grade t o the main f l o o r .

"he buildings

are two-level s t r u c t u r e s w i t h t h e grade f l o o r s of the turbine and reactor buildings a t an elevation of one foot above grade, and the main f l o o r a t

36 f e e t above grade.

The secondary shield wall extends t o t h e main f l o o r

and forms the w a l l s of the lower part of the r e a c t o r and a u x i l i a r y building

-

"he turbine building and the upper l e v e l of the reactor and auxiliary

buildings are s t e e l frame structures, w i t h insulated metal panel siding. The arrangement of the equipment within the buildings i s indicated on the general arrangement drawings, Figs. 4.5 and 4.7.

'

A three-level s t e e l frame and insulated metal-pBnel structure ad-

joining the turbine building houses the administrative offices, control

room, switchgear, b a t t e r i e s , plant heating b o i l e r and makeup water demineralization plant.

Lockers, showers, and t o i l e t s f o r plant personnel

are also located i n t h i s building. A 200-ft waste gas stack i s provided f o r d i s p e r s a l of plant venti-

l a t i n g a i r and waste gases from the various reactor and reactor equipment rooms.


! -

.

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i

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44 4.4

Primary System Components

The primary system components consist of the reactor vessel, modera-

tor, fuel salt pumps and fuel salt-to-intermediate coolant salt heat exchanger. The arrangement of these components is shown in Figs. 4 . 1 and 4.2. Design and performance characteristics are summarized in the following paragraphs. 4.4.1

Reactor Vessel The reactor vessel is 20 ft in diameter and 37 ft high (including

expansion dome) as shown in Fig. 4.1. The vessel wall is fabricated of 2-in.-thick INOR-8. No thermal shield is required since the gamma heating of the vessel wall does not exceed 0.6 w/cc.

The thickness of the external vessel insulation and salt flow through the core-vessel annulus can be adjusted as necessary to avoid excessive thermal stresses in the vessel wall and at the same time minimize heat loss to the external containment space.

A 6-ft diam expansion dome at the top of the reactor vessel not only provides surge volume but serves as the fuel salt volatiles purge location, the UF4 pellet injection point and salt sampler location. Circulation of fuel salt in the dome is accomplished by recirculation of salt from each pump discharge through an orificed 2-in. line to a point below the normal liquid level in the dome. This level is maintained by adjusting the pressure of the helium cover gas. To minimize holdup of salt, 5%

of the volume of the dome is occupied by 2-in. d i m sealed INOR-8

tubes. The reactor vessel is provided with eight inlet nozzles which discharge fuel salt radially into the bottom plenum, eight outlet nozzles leading to the pumps, together with bottom and top INOR-8 grids for support and restraint of the graphite core. These grids in turn are supported by columns or stanchions attached to the reactor vessel. The reactor vessel specifications are given in Section 4.10.


45 i i

'U

4-4.2 Moderator Skructure The graphite matrix is composed of cylindrical logs 8.in. in diameter and 24 in. long, as shown in Fig. 4.8. These are stacked in a vertical position and alligned by means of axial pins and sockets. Fuel-salt flows up through the cusp-shaped passages between logs. The "pile" is centered in the vessel by metal pins protruding from the support plates at the bottom and top of the vessel. These are located near the axis of the vessel and mate with a corresponding moderator log. Initially the pile rests on the lower support grid; as the reactor is filled with salt, the pile floats up against the upper grid. The pins, while allowing this vertical motion, keep the central logs centered in the vessel. The remaining logs axe bound to these by means of metal hoops passing around the peripheral logs. These hoops are fabricated from molybdenum, which has about the same coefficient of expansium as graphite. This arrangement allows the support grids to expand independently of the moderator as the reactor is broughk to operating temperature. Also, the increase in height of the vegsel on expansion is accommodated.

r

"he radial profile of temperature in the fuel-salt is flattened by proper distribution of the flow, which is accomplished by orificing the flow channels. The bottom row of moderator logs is machined from hexagonal pieces. A &in. section of the end in contact with the support grid is not machined. If close-packed in a triangular lattice, these ends would block completely the f l o w path of the fie1 salt. Therefore, the corners of the hexagons axe cut away to provide orifices of appropriate diameter for each channel.

4.4.3

Fuel-Salt Circulating Pumps

Circulation of the fuel salt flowing at 9075 gpm is maintained by a centrifugal pump as sham in Fig. 4.9 in each of the eight independent heat exchange circuits. Because the system has no valves or other means of equipment isolation, the pumps are installed at the highest elevation (and the highest temperature region) of the circuit between the reactor and the primary heat exchanger. By this means it is possible to avoid the hazard of seal flooding and reverse rotation at standstill and to


46 ORNL-LR-DWG

Bhck Fig. 4 . 8 .

MSCR Core Configuration.

63-2803

Annulus


47 ORNL-LR-DWG 417651

i

1

7

:

1

VEL

Fig. 4.9.

Molten S a l t Journal Bearing Pump.


48 minimize some design problems r e l a t e d t o thermal expansion, shaft s e a l

and guide bearings, and maintenance a c c e s s i b i l i t y . Locating the pumps SO close t o the reactor, however, introduces the problem that the organic materials used i n the motor winding insulatiaq and bearing lubricant must be shielded from radiation.

Also the pump motor and oil-lubricated bear-

ings must be located a t some distance from the s a l t region.

In t h e design

shown, the motor and oil-bearings are positioned above the pumps f o r easy maintenance and a lower salt lubricated bearing providbd t o take the radial thrust.

Circulation of salt through t h i s bearing i s accamplished

by allowing salt t o leak upwards around the shaft and out a smaU vent line.

A c y l i n d r i c a l casing above the impeller provides expqnsion volume

and a v o l a t i l e f i s s i o n product purging w r f a c e .

The surface of s a l t i n

t h i s casing i s maintained a t the Same leveJ as t h a t i n the expansion dome

under a helium cover gas a t 22 psi.

This gas overpressure a l s o helps

prevent impeller cavitation. Detailed specifications of the MSCR pumps axe given i n Section 4.10-

4.4.4

Primary Heat kchanger The design of the heat exchanger f o r transferring heat f r o m the f u e l

s a l t t o the intermediate coolant salt i s shown i n Fig. 4-10. This s h e l l and tube heat exchanger i s of a U tube configuration designed p a r t i c u l a r l y f o r a c c e s s i b i l i t y t o the tube sheet from above without disturbing connecting piping.

Removal of the tube hundle through the top head i s a l s o

possible w i t h the given design.

With-this configuration, however, the

more valuable f u e l salt, which normally would be c i r c u l a t e d through the tubes, i s put on the s h e l l side t o permit ready drainage.

-

The tube bundle

and associated baffles, which can s l i d e i n t o the s h e l l from above, hang from the tube-sheet which r e s t s on a shelf machined i n t o the s h e l l .

A

circumferential s e a l weld j o i n s t h e tube sheet t o the s h e l l and separates the coolant from the fâ&#x20AC;&#x2122;uel salt.

A Wid-down r i n g keeps the tube sheet i n

place i n case pressure on the s h e l l side should exceed pressure on the tube side.

Normal design conditions require that the coolant salt (tube

s i d e ) be kept a t a pressure higher than t h a t of t h e fâ&#x20AC;&#x2122;uel salt.

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bid i


49 ORNL-LR-DWG

Fuel Salt

II Y

i

Fig. 4.10. ,

ciis

MSCR Primary Heat Exchanger.

63-2802


-

50 "he end-closure may be sealed against leakage of coolant salt t o the outside bp several ways.

A bearing and sealing surface (much l i k e a

valve s e a t ) i s provided at point "A" shown on Fig. 4.10. s e a l i s maintained i n the annulus above "A".

A frozen s a l t

A s e a l weld could be made

a t the top between the flange on the s h e l l side and the inverted head. O r l a s t l y , double-gaskets on c i r c l e s inside and outside of the s e a l weld

l i p could confine leakage i f the s e a l weld, or other s e a l s ahead of the s e a l weld, should f a i l .

A massive hold-down flange bolted t o the top

flange provides the strength necessary t o hold t h e inverted head against the several hundred pound pressure of the coolant salt. A p a r t i t i o n attached t o the inverted heat f i t s closely i n t o a dia-

metrical s l o t i n the tube sheet t o separgte t h e i n l e t coolant salt plenum

from the o u t l e t plenum.

A s m a l l amount of by-pass flaw may occur thraugh

the very small gap between the s l o t and the p a r t i t i o n .

The inverted head

i s shaped so as t o minimize the volume of coolant s a l t i n the tube-sheet d i s t r i b u t i o n plenums. The i n l e t and o u t l e t coolant nozzles are noncircul a r i n cross section ( f l a t t e n e d and broadened so as t o minimize head-space without s a c r i f i c i n g flow area). Not shown are cooling and heating p r w i s i o n s f o r $he top head, heati n g provisions f o r the shell ar t h e insulation which must be provided.

I n order t o avoid excsssive thermal s t r e s s e s i n the tube sheet it may be necessasy t o reduce the thermal gradient w r o s s the tube sheet by providing insulation between it and the f u e l salt. .

An appreciable heel of coolant salt w i l l remain i n the tubes a f t e r draining.

I n order that the salt-containing i n t e s n a l s of the heat ex-

changer be exposed only t o i n e r t atmospheres when the top closure i s r e moved, the upper portion of the heat exchanger (above the coolant nozzle)

i s prorrided w i t h a gas-tight caisson t h a t extends from the heat exchanger t o the l o c a l i t y of the maintenance equipment. The specifications given i n Section 4.10 are based on eight heat exchangers per reactor, each w i t h a heat load of 1.006

X

lo9

Btu/hr per u n i t .

6 a

*

i


51 4.5 4.5.1

Intermediate Coaling System

Introduction Leakage of steam i n t o the f u e l s a l t would res-alt i n precipitation of

uranium; therefore, a material which has good heat transfer properties and

i s compatible with both f u e l s a l t and steam must be interposed between these two materials.

Of t h e various possible choices, a barren salt of

LiF-BeF2 appears t o be the most promising and i s used as a basis f o r the present MSCR design involving a Loeffler steam generation system.

In this

design the intermediate cooling system consists of the heat exchanger a l ready described, coolant s a l t pumps, steam superheaters and steam reheaters.

4.5.2

Coolant S a l t Pumps The coolant s a l t pumps a r e s i m i l a r i n design t o the f u e l s a l t pumps

but have a higher capacity and head.

Eight pumps serve the sixteen super-

heaters and four reheaters and each pump c i r c u l a t e s 13,900 gpm of coolant s a l t a t a temperature of 950째F and the primary heat exchanger i n l e t pressure of 300 psia.

The developed head of the pump i s approximately 270 f t .

Additional specifications f o r the coolant pumps a r e given i n Section 4.10.

4.5.3

S t e m Superheaters Steam from the superheaters is divided among several streams.

2%

About

i s used t o drive the turbogenerators f o r the production of e l e c t r i c i t y ;

a much smaller f r a c t i o n i s used t o drive other turbines which power feedi

-

water pumps and the steam circulators.

The balance i s mixed w i t h feedwater

i n the Loeffler b o i l e r t o produce saturated s t e m required i n the superheater. A U-tube i n U-shell arrangement w i t h fixed tube sheets and counter-

current flow of coolant s a l t and steam i s shown i n Fig. 4.ll.

Coolant

s a l t i s on the shell-side, and steam flows through the tubes.

The unit i s

mounted on i t s side t o minimize f l o o r space. t o minimize pressure drop.

The shell-side i s unbaffled

The heat t r a n s f e r coefficient with near-laminar

flow of coolant s a l t i s about 15% l e s s than f o r baffled flow, and requires ?


52

ORNL-LR-DWG 39169A

COOLANT SALT IN

A

TUBE SHEET

SANRATED

STEW IN

U t

COOLANT SALT OUT

Fig. 4.11. MSCR Steam Superheater.


53

more tubing.

To prevent freezing of the s a l t , the superheater i s brought

up t o operating temperature by steam generated i n an auxiliary, o i l - f i r e d boiler. Coolant s a l t operating between 950 and 1100째F c i r c u l a t e s through the shell-side.

Steam enters a t 670째F and 2490 p s i , and leaves a t 1000째F and

2465 p s i . 4.5.4

Steam Reheaters I n order t o provide greater energy a v a i l a b i l i t y for the turbine work

cycle and t o provide lower pressure steam t o drive auxiliary turbines, the high pressure turbine exhaust steam i s directed t o the reheaters, the auxiliary turbines, and the second feedwater heaters.

The same general

design c r i t e r i a apply t o the reheaters as t o the superheaters, although the steam i s a t a lower pressure.

As with the superheaters, the reheaters a r e constructed of INOR-8 and alloy s t e e l .

Eight u n i t s provide the desired heat transfer surface.

Again, the coolant s a l t flows i n the s h e l l countercurrently t o the steam passing through the tubes.

Comparable pressure drop and heat transfer

conditions e x i s t i n the reheater a s i n the superheaters, aJthough b a f f l i n g ( t o improve heat t r a n s f e r ) on the s h e l l introduces a pressure drop of about 40 p s i . 4.6

4.6.1

Power Generation System

Introduction Before selecting the power generation system f o r the MSCR a number of

a l t e r n a t i v e approaches were considered i n some d e t a i l .

These included the

use of ( a ) a mercury boiler, ( b ) a "Kinyon" boiler, ( c ) a "Bettis" boiler, and ( d ) a "Loeffler" boiler.

The c h a r a c t e r i s t i c s of these a l t e r n a t i v e

steam generation systems a r e summarized as follows. The use of binary mercury-steam power cycles f o r power generation

-

from s o l i d f u e l reactors has been studied by Bradfute et-Lal, (13) and Randall e t al. (92). Kinyon and Romie (55) considered application of the binary cycle t o molten s a l t reactors.

The binary system, i n addition t o


54 mitigating the thermal s t r e s s problem, has the f'urther advantage of a higher thermal efficiency (95).

On the other hand, bimetallic heat ex-

changers are required a t t h e high tempersture end, and it i s not obvious that the f u e l salt volume w i l l be small compared t o that involved i n the Loeffler system. "he "Kinyon" b o i l e r (53) employs bayonet elements composed of three concentric tubes.

Water is introduced i n the annular passage surrounding

the innermost tube and i s botled by heat transferred f r o m superheated Vapor i s separated from t h e

steam f l w i n g downward i n the outer annulus.

l i q u i d a t the top of the inner annulus; the l i q u i d i s routed down the c e n t r a l tube t o a plenum from which it i s recycled t o the boiling annulus. Vapor passes down t h e outer annulus; it i s superheated by salt surrounding the bayonet element and i n turn b o i l s the water r i s i n g i n the inner annulus.

This arrangement results i n t o l e r a b l e thermal s t r e s s e s

tubing. . The system avoids the use of steam r e c i r c u l a t o r s .

i n the

Furthermore,

the bayonet tube b o i l e r may be more compact and l e s s eppensive than the equivalent superheaters and b o i l e r s i n the Loeffler system.

It i s not

c l e w , however, t h a t the required rate of vapor-liquid separation can be achieved with present technology. The "Bettis" b o i l e r i s a modification of the "Lewis" b o i l e r (59,55) wherein a bayonet tube containing water and steam i s i n s e r t e d i n t o a thimble extending down i n t o the f u e l salt.

An i n e r t buffer s a l t occupies

the annular space between the pressure tube and the thimble, providing

thermal contact along w i t h physical isolation.

This system avoids use of

steam c i r c u l a t o r s and a l s o eliminates the necessity f o r pumping the intermediate s a l t .

4.6.2

Loeffler Boiler System I n the system selected for the reference design, four b o i l e r s gen-.

e r a t e steam by d i r e c t contact between the steam from the superheaters and the feed water from the turbine cycle.

Four steam compressws c i r c u l a t e

the saturated steam from the b o i l e r s through the superheaters of the i n e r t

salt system, providing steam a t 2400 psia, 1000째F a t the turbine t h r o t t l e and a t the superheated steam i n l e t of the Loeffler b o i l e r s .


55

These boilers, sham in Fig. 4.12, consist of cylindrical drums with an inside diameter of about 72 in., with distribution pipes for super- ', heated steam and feed water located on the vertical centerline and below the horizontal centerline of each drum. Conventional cyclone steam separators, located above the horizontal centerline and arranged in four parallel rows along the inside of the drum, separate the steam from the boiling water. A system of scrubber-type steam driers occupies the upper part of the drum. In operation, superheated steam and feed water are mixed as they leave the two distribution pipes and boiling occurs. The boiling water is guided by internal baffles through the cyclone separators where the water is removed by centrifugal force. Steam and a small amount of moisture flow through the scrubbers, where the steam is dried to approximately 99.7 quality. The drums are fabricated of carbon steel and are designed for a pressure of 2625 psia. A wall thickness of 7 in. is required for an inside diameter of 72 in. and a temperature of 650째F. drum is about 100 ft in length and has hemiellipsoidal heads. 4.6.3

Each

Steam Circulators \

The steam circulators are single-stage axial compressors suitable for a steam flow of approximately five million lb/hr at a discharge pressure of 2480 psia. The circulators are driven by steam turbines, using steam f r m the cold reheat lines of the main turbine-generator. At design conditions, each circulator requires a power input of 4750 BHP. Steam from the turbines is returned to the cycle through the feed-water heaters. One circulator has a motor drive suitable for full flow at design pressure to be used during stat-up shutdown. An oil-fired package boiler, designed for a saturated steam flow of 50,000 lb/hr at 300 psia, is provided for stetrtup purposes. The system is designed t operate as four 1/4 capacity units; each unit consists Qf one Loeffler boiler, one steam circulator and four sbperheaters. Valves are located according to this philosophy. Interconnections between various units are not provided except at the inlet to the boilers and the discharge from the superheaters.


._.

ORNL DWG., 65-7908

l4 in. Sch. 160 Steam Header Cyclone Separat

uperheated Steam Headers Operating Conditions (one unit): 2 X lo6 lb/hr @ 545OF, 2500 p s i I n l e t Water 3.1 x lo6 lb/hr @ 100째F, 2450 p s i I n l e t Steam 5.1 X lo6 l b / h r , sat. @ 2400 p s i Outlet Steam

Fig. 4.12.

MSCR Loeffler Boiler.


57

4.6.4 Turbogenerator The turbogenerator is a standard CC6F-RH unit operating at 2400 psia 1000°F and producing a gross electrical output of 1083 Mw. As shown in Fig. 4.3 the units consist of three turbines: two double-flow high pressure units and one double-flow intermediate pressure unit on one shaft with a 667 Mw net output plus three double-flow low pressure units on another shaft with a 416 Mw net output. The plant auxiliaries have a maximum coincidental loading of 45,000 kw; this power results in a net station output of 1038 Mwe to the station main power transformers.

Other characteristics of the power generation system are given in Fig. 4.3 and summarized in Section 4.10.

4.7 Reactor Control System 4.7.1 Introduction The specifications for the MSCR control system depend on the unique properties of its fluid-fuel. The main mode of shim control is by adding

f u e l salt with a salt containing no fissile isotopes to decrease reactivity. This control is supplemented by operational control with BF3 gas, which is sufficiently soluble that adequate concentrations may be obtained in the salt by regulating the partial pressure of BF3 over the salt surface in the dome. Fuel is added to compensate burnup and fission product accumulation. Greater reactivity control is required during startup and shutdown. The startup procedure involves a gradual increase in power, during which the rising xenon poisoning (controlled by the flux level) is compensated by the addition of new fuel. After shutdown, the reactor is sub-critical for forty hours while 135X grows in. If longer shutdm is required, then fifty cubic feet of fâ&#x20AC;&#x2122;uel salt must be removed and replaced by nonfissile salt. Emergency shutdown, such as that required in a loss-of-flow accident, UF4 to increase reactivity and by replacement of

is provided by injecting BF3 gas through tubes opening into the core just above the bottom support grid. The BF3 will displace fuel, will rapidly

, ,


* .

58

-4

dissolve in the salt, and will diffuse into the graphite. A l l of these processes will result in a rapid and large decrease in reactivity. J

4.7.2

Shim Control For shim control by fuel addition or removal, two reactivity-control

drain tanks of 50 cubic ft capacity each and a shim-control salt-addition tank of 50 cubic ft capacity are located in the reactor containment cell.

All tanks are provided with vents to the reactor dome, gas connections to the helium pressurizing system, and several100 kw electric heaters for initial vessel heating and maintaining stored salt tempersture. The addition tank is essentially in parallel with the 2 cubic ft salt-addition metering tank. The drain tank is connected to the reactor vessel upper plenum and is installed with respect to the reactor dome liquid level so that overfilling will not be possible. Freeze valves are used to isolate the tanks from the reactor. The reactivity control drain tanks must have, in addition to heating coils, decay heat removal facilities similar to the fuel salt drain tanks. They have an independent steam condenser and heat removal system which will be 1/20 the size of the fuel drain tank heat removal system. .

4.7.3

Ehnergency Control

The BF3 addition system comprises pressure cylinders manifolded to a heater. BF3 flow control and quick-acting injection valves are incorpo-rated in the supply line connected to the bottom plenum of the reactor vessel. The flow control facility is operated from the shim control instrumentation and the quick-acting injection system is operated from the scram" circuits.

I1

The BF3 gas is stored in six 330 cubic ft, 2000 psi cylinders located outside and adjacent to the east wall of the reactor building. The specifications for the reactivity-control drain tanks and the reactivity-control salt-addition tank are given in Section 4.10.

w:


z

*c

59

4.8 4.8.1

Salt Handling Systems

Introduction Facilities for handling both fresh and spent fuel salt and inter-

mediate coolant salt constitute the heart of the MSCR complex. The functions which must be performed include melting, purifying, charging, removal, storage, and sampling of the salt. These are accomplished by means of the following systems and facilities shown in Fig. 4.13. a. Fuel salt preparation b.

Coolant salt preparation

C.

Reactor salt purification

d. Coolant salt purification Reactor salt charging system

e.

f. UF4 addition facility

g* Fuel salt drain and storage system h. Coolant salt drain system i.

Spent fuel withdrawal system

3.

High level radioactive sampling

k. Coolant salt sampling 4.8.2

Fuel Salt Preparation Fuel salt is received in solid form and must be liquified prior to

processing.

Barren salt, having the same composition as the intermediate

coolant salt, i.e. , 66% LiF and 34% BeF2, with a liquidus temperature of 851'F

is used for preoperational testing and system flushing before and

after maintenance activity. A solid mixture having this composition is charged initially. Sufficient LiF and ThF4 are added to form composition 68% LiF, 23% BeF2 and 9% ThF4. All mixtures are processed through the same equipment. After routine operation is established, flush salt is the base material for the lithium/thorium addition process. In this man-

c

ner, flush salt contaminated by fission products may be consumed and special treatment for removal of fission products will be unnecessary.

A 12-ft-highJ150 cubic ft tank enclosed by an 8-ft-highJ150 kw furnace receives solid salt for melting.

The furnace, located in the


0

cn


61 f l u s h - s a l t storage area, i s designed f o r melting approximately 7 f t 3 / h r . Molten s a l t i s transferred by pressure siphoning. Solid s a l t i s added by means of a tube extending from t h e f l o o r above the storage area t o the melt tank. sealed.

The top of the tube i s flanged and

A valve i s located j u s t below the flange.

The tube takes a

devious route and i s provided w i t h a helium purge t o prevent back-flow of radioactive material and intolerable radiation l e v e l s a t the open feed point.

A check valve j u s t below the flange allows purge gas t o be di.-

rected i n t o the melt tank.

A portable hopper i s connected t o the flange

face t o receive s a l t from the shipping containers.

A portable shed covers

the work area t o prevent the spread of toxic dust r e s u l t i n g from the feed operat ion.

4.8.3

Coolant S a l t Preparation Equipment and system design a r e s i m i l a r t o t h a t of f u e l - s a l t melt

tank and s o l i d salt feed.

The melt tank i s located i n the heat exchanger

room near the chemical treatment tank. above.

Solid salt i s fed from the f l o o r

4.8.4 Reactor S a l t Purification The reactor f u e l and f l u s h s a l t s are p u r i f i e d i n the molten s t a t e p r i o r t o charging or storing.

This i s done i n i t i a l l y t o remove oxides,

and subsequently, t o remove oxides and other contaminants. I n normal operation, a chemical treatment tank of 100 cubic f t c a t pacity located i n the f l u s h - s a l t storage area receives the molten salt. After a one-day holdup f o r purification, the salt i s transferred t o s t o r age tanks.

By t h e addition of LiF and ThF4 t o f l u s h s a l t i n the melt

tank, f e r t i l e s a l t i s formed; t h i s is then transferred t o t h e f e r t i l e s a l t p u r i f i c a t i o n tank where gaseous HEâ&#x20AC;&#x2122; and H2 a r e bubbled through the l i q u i d t o remove oxides.

Subsequently

t h e t r e a t e d s a l t i s transferred t o t h e

100 cubic f t salt-storage tank i n s t a l l e d a t tde same location. Flush-salt make-up is processed i n the same manner, and through the same equipment. After treatment, it i s transferred t o a spare f l u s h - s a l t tank.

For the i n i t i a l charges of f l u s h s a l t and f e r t i l e salt for the react o r system, greater purifying capacity i s required i n order t o avoid delay


62 i n preoperational t e s t i n g and power production.

n

To f u l f i l l t h i s condi-

tion, HF' and H2 bubbling f a c i l i t i e s are provid6d i n t h e f e r t i l e s a l t s t o r age tank and temporary flanged l i n e s arranged so t h a t t h e f e r t i l e s a l t

tank may receive molten s a l t from the melt tank, perform p a r t i a l p u r i f i cation and allow t r a n s f e r of the s a l t t o the normal treatment tank. I n this manner the two tanks, f e r t i l e storage and chemical treatment, are placed i n s e r i e s t o double both the p u r i f i c a t i o n r a t e and the system s a l t

charging r a t e .

This procedure does not increase t h e sampling or analysis

requirements over those necessary f o r normal operation.

4.8.5

Coolant-Salt Purification 5

"he purpose of the coolant salt purification system i s t o remove impurities such as corrosion products or oxides which could cause fouling

c

of surfaces and plugging of l i n e s and tubes i f allowed t o accumulate..

It i s unlikely that a single coolant charge could be used f o r the whole lifetime of the reactor without exceeding permissible concentrations of oxides or corrosion products; however, it probably w i l l not be neces-' sary to employ a continuous treatment of coolant salt.

Oxides may g e t i n t o

the coolant s a l t by accidental exposure t o air or water vapor, or from oxygen present a s an impurity i n the c w e r gas used t o pressurize ishe coolant-salt systems.

Although the r a t e of corrosion of INOR-8 surfaces

by coolant s a l t i s low, the area of metal surface i n contact w i t h coolant

salt i s very large, so that corrosion products a r e c e r t a i n t o accumulate. When it becomes necessary t o repurify the coolant s a l t , it i s done batchwise, one coolant c i r c u i t per batch, a t infrequent i n t e r v a l s during p e r i - ' ods of reactor shutdown.

If the coolant salt should become contaminated w i t h f i s s i o n products

or uranium as a r e s u l t of a leak i n the primary heat exchanger (an event not l i k e l y t o occur because the coolant salt system i s kept a t a pressure higher than t h a t of the f u e l s a l t ) , the contaminated coolant s a l t is drained from the affected c i r c u i t t o a drain tank, and then transferred t o the chemical processing p l a n t f o r disposal. J-.

G â&#x201A;Ź9

t


63

4.8.6

Reactor Salt Charging System

Fuel salt without uranium is injected into the primary reactor circuit by manual control from a small salt addition tank to replenish the fuel salt withdrawn for chemical processing (two cubic ft per day). This addition is made remotely and with assurance that fuel salt will not flow back into the addition system. The make-up is fed into the reactor in a molten condition by gravity flow or under pressure. Sufficient electrical heating to maintain the salt in a molten condition prior to injection is provided

.

A two-cubic-ft salt addition tank (metering tank) is located on the wall of the biological shield at a level somewhat above the liquid level in the reactor. The tank is supplied from the fertile-salt storage tank situated in the flush salt storage area. Makeup is accomplished by opening the freeze valve and the vent valve between the tank and reactor dome. The freeze-valve in the line leading to the reackor dome must be placed at the bottom of a loop so that when the tank is empty a heel of salt will remain in the valve.

4.8.7

Intermediate Coolant Charging System

Molten coolant salt is injected into any one of the eight intermedi-; ate coolant loops from the coolant drain tanks at a rate up to 600 ft3/hr. The drain lines from several locations in each of the eight intermediate c o d a n t loops also serve as charging l i n e s f o r t h i s operation.

4.8.8

UFr, Addition Facility

During steady-state ope are added to the reactor to c of 2 ft3/day. The amount and

on at equilibrium, about 6 kg/day of vF4 ensate for burnup and a fuel withdrawal of addition of UF4 to the reactor is governed by reactivity and tempe e requirements. Valves in the vF4 addition system must be kept gas-tight when closed even though they must pass solids when opened, but since hey will be in a region of relatively law radiation level, plastic seats (for non-scoring properties) may be used.


. 64 /-

The UF4 charged i n t o the reactor system i s prepared by mixing r e -

W'I

i n diam, fabr i c a t e d by casting IJkb i n an i n e r t atmosphere, are charged i n t o a gascycled uranium w i t h fresh uranium.

P e l l e t s about 0.75-in.

t i g h t shielded shipping container, containing an atmosphere of dry helium

a t approximately 1 atmosphere absolute pressure.

A t the reactor, the

shielded shipping container i s mounted on the f u e l charging machine (which i s located above the r e a c t o r ) , and the mated assembly of t h e shipping container and f u e l charging machine i s made gas-tight by bolting a gasketed flange.

Figure 4.14 shows a schematic sketch of the UFI, addition f a c i l i t y . Fuel S a l t Drain and Storage System

4.8.9

F a c i l i t i e s are provided f o r the drainage and storage of f u e l salt during periods of maintenance and emergency shutdown. on the flow diagram i n Fig. 4.2.

The system i s shown

A l l s a l t c i r c u i t s have drain connec-

tions; this includes the primary heat exchangers a s weU as the reactor vessel.

The draining of the f i e 1 salt i s accomplished quickly, even under

emergency condition without steam flow and a u x i l i a r y power, since without heat removal the s a l t temperatures r i s e 300400째F i n the f i r s t half hour. Four equally spaced 2-in. drain l i n e s connected r a d i a l l y tQt h e bottom plenum of the reactor vessel, w i t h 1-in. interconnections t o the primary heat exchangers drain connections, will drain the reactor system i n approximately one-half hour after the freeze valves a r e opened. The tanks a r e capable of receiving the f u l l inventory of t h e reactor

-

within half an hour, maintaining t h e stored Volume of salt at a temperat u r e of less than 1400째F, and recharging it t o the reactor.

Gravity drain

-

and gas pressure are used f o r t r a n s f e r r i n g salt. Because forced c i r c u l a t i o n may not be available under emergency conditions, n a t u r a l convection cooling i s provided i n small-diam c y l i n d r i c a l vessels.

Fifty-four v e r t i c a l l y mounted 35-in. ID cylinders 8 f t long a r e

located i n a l o o f t trench around the inside periphery of the reactor containment c e l l . bonate bath.

Each cylinder i s immersed i n a molten alkali metal carThe carbonate mixture i s contained i n a double-walled tank

which serves a s a b o i l e r f o r the removal of decay heat from the f u e l salt. Fifteen-kilowatt e l e c t r i c heaters a r e arranged within the carbonate m i x t u r e annulus f o r f u e l - s a l t mel4ing following long( storage periods and,:

.ip, -f


65 ORNL- LR- DWG 75326

Shielded Shipping Container

UF4 Pellet Container

M Flange with Soft

Metal Gasket

I cu.ft. Pellet Bln and Dispancer

Metering Cham bar UF, Pellet

Top of Reactor Shield

Ma intenance Valve

JV

UF, Addition Pipe

Vent Opening

I

\c- Reactor Dome

-

f&L @9'

Gas Space Salt Level Perforated U F4 Dissolver Tube

Fig. 4.14. MSCR UFQ Addition Facility.


!

- 1 i

66

c

W' temperature maintenance during shorter periods.

The heaters a r e sized t o

bring the s a l t t o llOO째F i n four days.

4.8.10

Coolant S a l t Drain System

These drain tanks have several purposes:

( a ) t o provide storage

space f o r the complete volume of coolant contained i n two (out

Of

eight)

coolant c i r c u i t s during periods when it i s necesswy t o drain the coolant from one c i r c u i t f o r maintenance; ( b ) t o provide the reservoir and the application of motive force f o r transferring coolant salt i n t o the react o r coolant c i r c u i t 6 during i n i t i a l salt-charging; and ( c ) t o serve a s a t r a n s f e r point from which coolant salt may be removed from the system and transferred t o shipping containers in'case it should become necessmy t o process contaminated coolant salt. The drain tanks a r e located a t an elevation which permits the coolant salt t o flow by gravity from the reactor coolant system i n t o the drain

tanks.

"he t r a n s f e r l i n e s m e valved with freeze-valves t o permit flow

t o or *om

any one of the 8 coolant c i r c u i t s .

The t r a n s f e r l i n e s a r e

large enough t o f i l l one coolant c i r c u i t i n one hour,:-using a gas pressure of 50 psig.

4.8.11 Spent Fuel Withdrawal System Fuel s a l t i s drawn from the reactor vessel drain l i n e a t about 30 p s i a i n t o the vented metering tank.

The metering tank i s a d d n . ( I D ) by 25-ft-long c y l i n d r i c a l vessel located a t an elevation such t h a t when flow stops it contains two oubic f t of f u e l s a l t .

Vent connections above the

s a l t i n the tank lead t o the dome on the reactor vessel.

Decay-heat r e -

moval i s accomplished by radiation and convection t o the reactor containment c e l l atmosphere.

Because t h e heat generation r a t e diminishes rapidly,

the r a t e of remwal i s controlled by means of an insulated jacket 4 in. thick surrounding the tank, and separated from it by a small a i r gap.

A

temperature-actuated bellows opens the jacket t o allow excess heat t o be radiated t o the room.

A 12-kw k l e c t r i c heater provides preheating and

aids i n maintaining desired s a l t temperature when the tank i s f u l l .

After a 24-hr holdup, t h e salt i s released by gravity drain t o a transfer tank.

When a 10 f t 3 batch of salt i s accumulated, the s a l t i s

i


67

transferred t o a shipping f l a s k , which i s placed i n a shielded cask and shipped by r a i l t o a central. processing plant.

Spent f u e l i s accumulated

f o r about 120 days a t the processing plant, and i s then processed i n about 6 days.

Should it be necessary or desirable t o hold spent f u e l a t the reactor s i t e (Section 5.3.4), f i v e 35 ft3 intermediate storage tanks a r e provided. Heat i s removed by boiling water i n steam jackets; the steam generated i s condensed i n water-cooled condensers.

4.8.12

High Level Radioactive S a l t Sampler

The high l e v e l radioactive s a l t sampler shown i n Fig. 4.15 i s used

t o obtain samples of f u e l or f l u s h s a l t from several points such a s the reactor ( v i a the f u e l withdrawal tank), the transfer tank i n the shipping area, the f u e l drain tanks, or the f l u s h - s a l t chemical treatment tank, and t o deliver these samples t o an adjoining hot c e l l f o r chemical analysis. Sampling i s accomplished by means of access ducts located a t approp r i a t e points.

These a r e mounted v e r t i c a l l y with no p i t c h l e s s than 50"

from the horizontal.

They a r e . l a r g e enough t o allow the f r e e passage of

the sample capsule, and axe equipped with a sample capsule cage a t the sampling point t o l i m i t the depth i n the salt melt reached by the capsule and prevent the end of the cable from dipping i n the s a l t . The

thief

sampler principle i s employed wherein an open sample

capsule made i n the form of a thick-walled, round-bottomed bucket hanging on a f l e x i b l e cable i s lowered i n t o t h e salt-containing vessel.

The cap-

sule sinks below the surface and i s f i l l e d with salt which s o l i d i f i e s when the sample capsule i s withdrawn t o a cooler region above the sample point. After s o l i d i f i c a t i o n , the capsule i s pulled up through the sampler access duct i n t o the sampler cavity i n which i s located the cable drive mechanism. Once i n the sampler cavity, the capsule i s positioned over a second duct leading t o a hot a n a l y t i c a l c e l l . the second duct and deposit

The sample capsule i s lowered through

'n the hot c e l l .

Using a suitable manipu-

l a t o r , which i s p a r t of t h e hot c e l l equipment, the sample capsule i s detached from the end of the cable, and a new sample capsule is attached. The new sample capsule i s then pulled up i n t o the sampler cavity i n readiness f o r another sampling operation.


68

-

ORNL- LR DWG 75325

Inner Containment Outer Containment

Sampler Duct Port

Valve Drive Motor

Inner Containmeni

-

Sampler Duct Valve

Outer Containment

Sampler Duct Valve

Fig. 4.15.

Radioactive Salt Sampler.


69

The sampler cavity is connected to the hot cell and to each of the four sample points by five separate sample access ducts. Each duct is made of 1 1/2-in. Sch. 40 pipe and contains two 1 1/2-in. double-disc gate valves. The bottom valve is closed only in the event the upper valve is remaved for repair or replacement 4.8.13

Coolant Salt Sampling

Coolant salt sampling connections for a portable sampling device are provided on the eight coolant loops and the chemical treatment tank. Samples are also drawn from the eight coolant-salt pump bowls and from the make-up salt chemical-treatment tank. A portable, single-sample holding device receives the sample in a manner similar to that of the high level radioactive salt sampler, transports the sample to the hot cell area, deposits the sample capsule in the hot cell, and receives a fresh capsule from the hot cell. The connection between the portable sampler and coolant salt system is of a lock type, has a gas-tight fitting and operates in an inert atmosphere (possibly radioactive) from f u l l vacuum to 200 psia at 100°F.

4.8.14

Freeze Valves

Freeze valves are used to close off all lines used for transferring salt fâ&#x20AC;&#x2122;rom one location to another. These valves are formed in any size of pipe up to 2 in. in d i m by pinching the pipe to form a rectangular shaped flow passage. In a 2-in. pipe, the flattened section is about 0.5 in. thick and 2 in. long. Closure of the valve is accomplished by freezing salt inside the pipe. This is done by directing jets of cold air agailrst the top and bottom surfaces of the flattened portion. The air, supplied by a Rootes-Connersville blower, i s controlled to regulate the rate of freezing. Subsequent thawing of the salt plug is achieved in a few minutes by means of a Calrod electric heater (3000 watts for a 2-in. line) bent in a saddle-shaped series of turns conformal to the flat section. This heater is easily remaved for maintenance.


70

4.9 4.9.1

Auxiliary Services and Equipment

Introduction

Auxiliary services necessary for the MSCR plant are as f;ollows: a. Helium cover gas supply and distribution b. Reagent gas supply and disposal Waste gas disposal de Liquid waste disposal e, Coolant pump lubricating oil system f. Preheating system C*

65. Auxiliary power supply h. Service water system

i. Control and station air systems j- Cranes and hoists k. Instrumentation and control system 1. Plant utilities 4.9.2

Helium Cover Gas Supply and Distribution System

The helium inert cover gas system serves a number af functions, viz: (a) as a pressurizing gas in various salt containing vessels for the transfer of salt from one place to another; (b) as a carrier gas for the removal of volatile fission products from the recycle gas purge system; ( c ) as an inert atmosphere to protect against contamination of the salt

in places such as the UFq system, samplers, melt tank atmospheres, etc., which are occasionally opened to the atmosphere; (d) as a gas seal and bearing lubricant in molten salt pumps; (e) to provide the pressures required in the fuel salt and coolant salt circuits to prwent pump cavitation, and to avoid leakage of fuel salt into the coolant salt in case of a leak in the primary heat exchangers; and (f) to pressurize leak-detection devices at various flanged joints and disconnects. Since the helium comes into contact with fuel salt or with intermediate coolant salt it must be free from oxide-containing impurities such

as H20, SO2, etc; therefore, a purification system is necessary as shown in Fig. 4.16. The raw helium is supplied from a trailer having a capacity

W

-


'

I

I

,

!4 0

.-

a

.. ,

..'


of 39,000 s t d . cubic f t i n 30 cylinders a t 2406 p s i .

An emergency supply

of 2400 s t d . cubic f t i s supplied from twelve 200 s t d . cubic ft c y l i n d e r s . A s shown i n t h e flow diagram duplicate l i n e s of' helium p u r i f i c a t i o n u n i t s

%re provided. 4.9.3

Each has a capacity of 1 . 5 scfm.

Reagent Gas Supply and Disposal System

A reagent mixture c o n s i s t i n g of 8% anhydrous kjrdrogen f l u o r i d e and 20% hydrogen i s used t o remove oxide i m p u r i t i e s from r e a c t o r and i n t c r -

mediate coolant s a l t s .

Unreacted reagent gas Ray contain r a d i o a c t i v e

m a t e r i a l , and thus must be handled by both t h e gaseous and l i q u i d waste disposal s y s t e m . The reagent gases a r e supplied from high pressure cylinders provided

with pressure reducers and flow contr.01 instruments s o t h a t t h e flow may be adjusted ( F i g . 4.16).

The mixed reagent gas i s bubbled through molten

salt contained i n t h e f u e l - s d t and t h e c o o l a n t - s a l t chemical treatment tanks.

Unreacted reagent gas passes through a potassium hydroxide (KOH)

scrubber t o remove HI?, and through a hydrogen burner t o remove hydrogen. Sperit c a u s t i c and condensate from the hydrogen burner may contain r a d i o a c t i v e m a t e r i a l and thus must be sent t o the l i q u i d waste d i s p o s a l syszem f o r concentration and u l t i m a t e shipment t o a remote d i s p o s a l area.

Non-

condensibles from t'ne hydrogen burner a r e cooled and vented from t h e system through t h e gaseous waste d i s p o s a l system.

-

4 9.4 Waste Gas System

As mentioned previously, helium i s passed through t h e dome above t h e core t o c a r r y away xenon and other v o l a t i l e f i s s i o n products.

I n addi-

t i o n , pwge gas i s passed down around t h e s h a f t s of t h e f u e l pumps t o sweep away f i s s i o n products d i f f u s i n g toward t h e upper, o i l - l u b r i c a t e d bearings.

A l s o , t h e r e i s helium cover gas i n t h e bowls of t h e pmps, i n

the durlp tanks, i n the f u e l handling system, e t c . , and a11 of these must be purged t o some e x t e n t .

I n order t o limit r a d i o a c t i v i t y i n t h e atmo-

sphere surrounding t h e r e a c t o r s i t e , t h e off-gas from these various systems i s passed through charcoal beds t o t r a p t h e f i s s i o n products u n t i l Yney have decayed, as shown i n Fig. 4.16.

A few long-lived isotopes,


73

particularly 85e, decay but little in the charcoal beds, and hence considerable dilution with air is required to limit the concentration of these in the stack discharge. If the BF3 addition system is ever used to effect a "scram" of the reactor, it will be necessary to remove the BF3 from the reactor system before normal reactor operation can be restored. A stripped unit removes BF3 from the recycle helium stream to avoid saturating the off-gas ad-

sorber as shown in the flow diagram (Fig. 4.16). 4.9.5

Liquid Waste Disposal System

Liquid wastes originate in the drains and sumps of the various equipment cells. Typical sources are hot sinks in the analytical laboratory,

cell drains in the shipping area where the shipping cask will be flushed with decontaminants, the contaminated equipment storage and decontaminak..: tion cell, and spent caustic solutions from the reagent-gas disposal system. "he liquid waste system shown in Fig. 4.16 provides for holdup, concenkration, and storage of active wastes on-site. High level wastes are concentrated by evaporation and stored in underground tanks. Low-level wastes may be held in underground tanks or in a low-level activity pond until they have decayed sufficiently f o r discharge into the river. The capacity of this pond is approximately 25,000 cubic ft (nearly twice the volume of the contaminated equipment storage cell) Intermediate wastes are sent to ten 10,000-gallon stainless steel retention tanks, which are used for>temporary storage until it is possible to send the waste to an evaporator f o r concentration. High level wastes from the plant and from the evaporator are sent to a L0,OOO-gallon waste-storage tank which provides semi-permanent storage of these wastes, which will ultimately be transferred to shielded shipping containers f o r shipment to a permanent disposal area. Spent caustic wastes from the reagent gas disposal system are sent to the intermediate-level storage tank for further concentration and ultimate disposal off-site. Pumps (mechanical, air jet, or steam jet) are provided at every sump, at the pond, at the 10,000-gallon tanks, at the evaporator tank, and at the 1000-gallon tanks to transfer the liquids.


74

The waste evaporator is capable of evaporating 200 gallons of water per hour using the plant heating steam, The condenser drairrs eit3er to the retention tank or to the l o w level waste pond, depending on the activity. A vent on the shell side allows noncondensibles to be vented to the charcoal beds. 4.9.6

Coolant Pump Lubricating Oil Systems

These systems supply oil to the bearings of eight pumps in the fuel circuits and to eight pumps in the intermediate coolant loops. The two groups of eight pumps each are supplied by two independent systems. The lubricant is circulated under oxygen-free conditions, 120 gpm to each pump group. The systems are designed to 120째F supply with an expected oil temperature rise of 20째F. 4.9.7

Preheating System Prior to the admission of salt, all salt-containing piping and equip-

ment must be preheated to 900째F. This is accomplished by the use of several types of resistance heating. In sizes up to about 2 in., piping is heated by passing an alternating current throygh the piping itself. '

All pipes heated this way are electrically insulated. Larger diameter salt-cogtaining pipes are heated by hinged resistance heaters (2000 w/ft ) surrounding the pipe. For large pipes (above 10 in.) a three-section heater assembly is used. Piping fittings are covered by prefabricated heating units. Heaters for pumps are field-fabricated. The reactor vessel and primary heat exchangers are heated by tubular resistance heaters attached to the outer surfaces by means of clips on .welded studs. 4.9.8

Auxiliary Power Auxiliary power is supplied at 4160 volts by a pair of auxiliary

power transformers, type QA/FA, 24-4.16 kv with a m a x i m fan-cooled rating of 25 Mva. A reserve transformer, fed from the switchyard bus, serves as a standby unit. The loads fed at 4160 volts include all motors rated at 150 horsepower or more, and the auxiliary power transformers


75 I

&cI,-.

required for station lighting, 480 volt auxiliaries, and the fuel melting and preheating system. Smaller auxiliaries, of 30 horsepower or less, and those whose continuous operation is not considered vital to station operation, are fed from motor control centers in the vicinity of the load. Two 15 kva 480120 volt transformers supply a 115 volt a-c control and instrumentation bus for each unit. An emergency supply to this bus is provided from a

15 h a inverter motor-generator set which is driven by the unit 250-volt battery. Service Water System

4.9.9

The service water system supplies river water for cooling purposes throughout the plant, including the reactor auxiliaries and the turbine plant components.

*

Water is supplied by three 15,000 gpm, half-capacity, vertical centrifugal pumps. Each pump is driven by a 1250-horsepower motor, and is located in the circulating water intake structure. During normal operatiop two pumps supply the system with water at LOO psig, with the third pump employed as a standby. 4.9.10

Control and Station Air Svstems

The compressed air system for the plant consists of separate, interconnected air supplies for the station and for control purposes. The station air system supplies hose valves far operating and maintenance re-

I

quirements throughout the station. Control air is used primarily for instrument transmitOers and air-operated valves. The two air systems are cross connected so that compressed air may be supplied to the control air system in event of a compressor failure.

The control air system of the plant supplies air-operated control devices at a header pressure of 115 psia and is reduced to 55 psia and 45 psia for supply to various drive units and instrument transmitters. 4.9.11

dr

Cranes and Hoists

A single traveling bridge crane serves the reactor and steam turbine

buildings. Its lifting capacity is based on handling the rotor of the

.

.

.


76 low pressure steam turbine-generator, which weighs about 150 tons. The bridge span is 130 ft, and the crane lift is sufficient to reach the lowest portion of the building. All heavy equipment coming into the building by rail may be handled by the crane. Its capacity is sufficient to allow removal of all components within the reactor primary shield except the reactor.

i

4-.9.12Instrumentation and Control The requirements for instrumentation and control of the turbine systems are similar to those of the turbine system in a conventional fossilfuel power plant. Instrumentation for the reactor system monitors the reactor neutron flux, primary system pressures, temperatures, levels, and.flow rates, Bnd provides control and alarm signals to actuate the appropriate device or to call for operator action when changes occur in the measured quantities, through either changes in load or malf'unction of system components. Control and instrumentation panels are located in the control room, for convenience of reading, recording and operating the most important quantities and components. Other auxiliary control panels or isolated instruments may be located at appropriate places in the plant; area radiation monitors, alarm or warning signals, hydrogen and seal oil controls for the generators, etc.

4.9.13 Plant Utilities The plant utilities include those systems that are provided for-monitoring plant equipment, disposing of nonradioactive wastes, safety of personnel, protection of equipment and for heating, ventilating and air-

L

-

conditioning the plant buildings. These systems do not differ appreciably from those provided for conventional plants.

f


77

4.10 MSCR Design Specifications '

The significant specifications for the MSCR equipment and materials are listed in Tables 4.2 through 4.6. The vessels are described in Table 4.2, heat-transfer equipment in Table 4 . 3 , pumps and circulators in Table 4 . 4 , miscellaneous equipment in Table 4.5, and materials in Table 4 . 6 .

.-A


mble 4.2.

? ;'Unit s 1. Reactor Vessel

1

54

~~~~~~

ft3

630 f u e l

Diameter in.

241

If?oR-8 Salt-Containing Vessels

Height

in.

Wall

in.

Heater Capacity (each u n i t ) kw

240

Design

Design

'F Psi

1300 outlet

100 in 50 out

U O inlet

Weight (each) l b gross, l b tare 1,100,OOO

50

35

96

15

1300

100

1

150

42

144

150

1200

50

F e r t i l e S a l t Chemical 1 Treatment Tank

100

36

168

15

1300

100

25,000 8

5. F e r t i l e S a l t Storage 1

100

36

168

l2

1x)o

100

23,000 8

1

2

12

32

1300

50

5oog

1

50

36

90

lo00

100

1

2

4

300

1300

150

9. f i e l S a l t Reactivity 2

50

36

96

10

1300

100

10. Fuel S a l t Decay 5 Storage Tanks ll. f i e l S a l t Withdrawal 1 Transfer Tank 12. Flush S a l t Storage 5

35

30

90

lo

lo

30

27

3

1300

25

650

96

156

150

1000

80

1

150

42

144

150

1200

50

14. Coolant S a l t Chemical 1

50

42

60

8

864

XK)

1100

80

2.

Fuel S a l t Draln Tanks

3.

F e r t i l e Sal0 Melt Tank

4.

6.

7. 8.

Tank F e r t i l e S a l t Small Addition Metering Tank F e r t i l e S a l t Iarge Addition Metering Tank

f i e 1 S a l t Small Withdrawal Metering Tank Control Drain Tanks

Tanks

13

Coolant S a l t Melt Tank Treatment Tank

15. Coolant S a l t Drain Tanks

2

750

43 *7

1

lo

2350 g 290 t

100,OOOg

10,000 t


0

4

C Table 4.3. Heat Transfer Equipment

Primary Heat Exchangers AEC Account Number Design data Number of units Unit heat rate, Btu/hr Geometry Number of tubes Active area, ft2 Active length, ft Length of longest tube, ft Length of shortest tube, ft Tube OD, in. Tube-wall thickness, in. Lattice pitch, in. Tube material Shell material Shell ID, in. Shell thickness, in. I"D,OF Shell weight, lb Tubing weight, lb Design pressure Heat transfer coefficient, Btu hr ftz.OF Shell-side conditions Fluid Inlet temperature, OF Outlet temperature, O F Flow rate, ft3/sec Pressure, psig Pressure drop, psi Volume, ft3 Tube-side conditions Fluid Inlet temperature, OF Outlet temperature, OF Flowrate, lb/hr Pressure, psig Pressure drop, psi

Steam Superheaters

Steam Reheaters

Fuel Salt Drain Tank Condensers

Loeffler Boilers

221.314

222 * 32

222.322

222.31

223.312

8 1,066 x lo6 Shell & U-Tube 2,025 6,643

8

4

1

Shell & U-Tube 766 3,543

Cyl. Drums None

85 x lo6 Shell & U-Tube 630 1,036

23.6

100

10

28.8

16 467 x 106 Shell & U-Tube 785 8,905 57.8 61.8

23.7

53.8

0.5 0.035 0.625 fnm-8 fnm-8 43.75 1.5 173.7 16,000 10,000

0.75 0.083 INOR-8 INOR-8 31.5 0.5 174.8

Carbon steel 72 7

0.625 0.065 1.5 Admiralty Carbon steel 26 0.25

25,000

187.5 4,500 9,620

924

300

119

Fuel salt no. 2 1,100 1,300 20.25 200 80 61.6

Int. cool. salt

Int. cool. salt

1,100 950 13.7 300 15 150

1,100 950 3.84 300 40 64

Int. cool. salt 950 1,100 31 350 84

Steam 670

Steam

25

132 x lo6

0.75 0.065 1.00

1 1 , m

INOR-8 INOR-8 31

0.5

3,800

6

1.28 x lo6 2,490 25

635 1,000 0.7 x lo6

440 20

Steam 240 120 80,000 10

Water 80 150 1.2 x 106 110

Steam 1.000 636 3.7 x 106 2,460

'Water 545 636 5.2 x lo6 2,450 50


.

.

.~

.

_ , -

. " ....

i - l ~ .

~~

.

~

Table 4.4.

.

.

.

~

. . , .... . .

. .

8 Centrifugal MSCR-2 fie1 salt 1300째F 190 lb/ft3

MSCR Coolant Pump

Discharge pressure (150 ft developed head)

220 psia

8 Centrifu@l Constant salt 1100째F 120 n p t 3 13,900 130 psia 350 psia

Impeller OD Suction ID Discharge ID Over-all pump OD (Approx,)

25 in. 14 in. 12 in.

25 in. 14 in. I 2 in.

50 in. 50 in.

50 in. 50 ' in.

Type Fluid pumped Service temperature Fluid density Fluid flow per pump Suction pressure (17 ft "SH)

9075 gpm 22.5 psia

-

Over-all pump height-suction opening to pump housing flange face (Approx.) Pump motor rating Pump motor speed ( syncbonous ) Pump motor type

.

.. .

.

.

. ..

.

.

Aunps and Circulators

bSCR Fuel Pump Number of units

.

Steam Circulators 4

Axi a1

Steam 1000"F 5.3 x. lo6 lb/hr

2430 psig

03 0

2490 psig

-n

1600 hp, 4160 volt 900 rpm Totally enclosed, 3 phase water cooled

2000 hp, 4160 volt 900 rpm Totally enclosed, 3 phase water cooled

,


81 Table 4.5.

Miscellaneous Equipment

Thermal shield 1 . 5 x 24 x 24

Dimensions, ft Material Shield Headers

2 in. plate carbon steel 20 in. x 4 in. carbon steel

(1/2 in. wall thickness)

Weight, lb g r o s s (water filled) Inlet/outlet pipe Heat removal rate, lo6 Btu/hr Water flow, gpm

250,000

Water temperature/pressure, "F/psig

90-110/15

Shield cooling system Heat load, Mwt Demineralized water circuit Maximum temperature, OF Minimum temperature, O F Service water conditions Maximum temperature, O F O F

Cell-air cooling system Heat load, Mwt Demineralized water circuit Maximum temperature, O F Minimum temperature, OF Flow rate, gprn Service water conditions Maximum temperature, OF Minimum temperature, OF Flow rate, gpm

18 2000

37.5

180 125 4650

Flow rate, gpm

Minimum temperature, Flow rate, gpm

8 in. Sch. 20 qarbon steel

125 75

5120

5 110 95

2275

95 75 1700


0

82

-.

1

w=I Table 4.5 (continued) High level radioactive salt sampler Sa'mpler cavity linner containment vessel Dimensions Material and thickness Design temperature/pres sure Sampler cavity shielding Material; thickness Approximate weight Sampler cavity outer containment vessel Dimensions Material and thickness Design temperature/pr es sure Sampler ducts

5 ft 4 in.

ID x 6 ft high

3oC, stainless steel, 1/8 in.

125"F/+15 psig Lead; 1 ft thick all around sides an8 top 57 tons

7 ft 8 in. ID x 8 ft high Carbon steel, X/4 in. thick 125OF/15 psig

Design pressure Design temperature

100 psig 200째F

Size Materia1

11/2 in. Sch. 40 pipe

Inconel

Sampler duct valves

Design pressure/temperature Number required

m e

100 psig/l25"F 10 1 1/2 in. Vulcan bellows stem double-disc motar-operated gate valve (supplied by Hoke Valve co. )

Cable drive unit Service environment

Drive unit will operate in an inert but radioactive atmosphere at nprmal temperature. Radiation level will be high only during sampling operation. The cable must be capable of operation at temperatures in the range 100 to 1300째F.

n.

w


83 Table 4.5 (continued) High level radioactive salt sampler (continued) Cable drive unit (continued) Cable size

UF4

addition facility Containment vessel Dimensions, ft Design pressure, psig Material

1/8 in. dim; length sufficient to reach from the sample cavity to the farthest sample point.

3 x 10 +15 304 stainless steel

Pellet bin and dispenser Volume, ft3

1

Metering chamber Configuration

Coiled tube, 1 in. pipe

Tube length, ft Tube material Valves Size (nominal) in. Description UF4

addition pipe

Dissolver tube Configuration

14 Stainless steel 1 Bellows-sealed, 150-lb design, gas-tight, soft-seated 1 in. Sch. 40 stainless steel Pipe ~

Perforated, coiled pipe

Size, nominal pipe size (in.) Material

1

Length, ft Other description

4

Shielding

INOR-8

End blanked off; perforation diameter, 1/8 in. 6 in. of lead


84 Table 4.5 (continued) Reagent gas disposal system data Hydrogen supply Cylinder s t a t i o n capacity, standard cubic f t

3000

Flow r a t e , scfm

2

Hydrogen fluoride supply Cylinder s t a t i o n capacity, standard cubic f t

12,000

Flow r a t e , scfm

8

KOH scrubber f o r HI? disposal

Cooling requirement, Btu/hr

8 118,000

Cooling w a t e r flow rate, gpm

10

KOH solution feed r a t e , l i t e r s/min ( m a x )

1

Flaw r a t e of unreacted hydrogen, scfm

2

Length and diameter, f t

8 and 2.8

Material of construction

Monel

Wall thickness, in.

1/4

WF f l o w r a t e (max), scfm

KOH supply tank

Dimensions

4 f t dim, 8 f t high

Material

Carbgn s t e e l

Chemical feed pump

0-1 liter/min

Mixer

2 hP

Hydrogen burner Design f l o w r a t e , lb-moles H2 per hour

1/3

Heat load, B/hr

41,000

Cooling water f l a w , gpm

4

Design a i r flow r a t e , cfm (max) Dimensions

10

Material

Carbon s t e e l

1 f t dim x 4t f t high


85 Table 4.5 (continued) Reagent gas disposal system data (continued) Hydrogen burner condenser Heat load, B/hr

20,000

Cooling water flow, gpm

4

Exit gas flow r a t e (nitrogen plus unburned oxygen), scfm

9,

Design temperature of e x i t gas, O F

100

Surface area of tubing, f t 2

100

S h e l l material

Carbon s t e e l

Tubing material

Admiralty

Configuration

Shell-and-tube, s t r a i g h t tube

Condensate r a t e , l b H2O per hr

6

.

COver gas p u r i f i c a t i o n system Helium dryer Dessicant used

Molecular sieve 10

Amount, l b Length/diameter

, ft/in.

Pressure r a t i n g , psig Container material

2.5/4

300 Caxbon s t e e l

Helium heater ( e l e c t r i c ) Design temperature gas e x i t ,

1200

"F Helium flow rate, scfm

1.5

Desige pressure, psig

300

Heater r a t i n g ( e l e c t r i c ) , k w

1

Oxygen removal unit Resign flow r a t e , scfm

1.5

Container size, length/diame t e r , in./in. (overall)

26/6 Sch. 40

Active ingredient

Titanium sponge

single pass,


$6 .,

Table 4.5 (continued) Cover gas purification system ( continued ) Oxygen removal u n i t (coptinued) Design pressure/ temperature, psig/"F

3(30/1200

Helium cooler I n l e t / o u t l e t temperature, "F

4 f t long l200/2QQ

Helium f l o w r a t e , scfm

1.5

4 in. finned tube w i t h flanges

Treater helium surge tank Tank volume, f t 3 Design pressure/temperature psig/"F Material

60 300/lOO Carbon s t e e l

Lube o i l system Number of systems

2

Design lube o i l flow, gpm

120

Design lube o i l temperature, OF (in/out)

140/120

Design cooling water temperat u r e , O F (in/out>

75/100

Design cooling water pressure, Psig Design cooling water flow, gpm

25

45

Number of lube o i l pumps

2

Type .- .- -

Rot

Lube o i l pumps head, f t

150 300 20

t

L

Flow, QFan Motor, hp Reservoir, number 2

1

Height, f t

5 5

Capacity, gal.

600

Wall thickness, in.

0.25

ID, f t ( c y l i n d r i c a l )


87 Table 4.5 (continued)

Lube oil system (continued) Reservoir, number 2 (continued) Materia1 Weight (full), lb Cooler, number

Carbon steel 3000

2

TYPe Tube material

2 pass shell, 4 pass tube Inconel

Shell material Overall heat transfer coefficient, Btu/hr*ft2 O F

Carbon steel 20

Tube surface area, ft2

735

Tube size, OD, in. Tube wall thickness, in. Tube pitch (triangular)

518 0.125

-

0.938

Shell diameter, ID, in. Shell length, ft Tube design pressure, p i g Tube design temperature,

OF

Shell design pressure, psig

18 34

25 100 45


88 Table 4.6.

Material Specifications

Properties Assumed for MSCR Graphite MSCR

Items and Units Density, g/cc Thermal conductivity, Btu/hr ' ft' "F At 68'F and with grain

1.9

1.9

80 45

At 68째F and across grain At 1200"F, isotropic, after irradiation

%RE*

15

Coefficient of thermal expansion, across grain, per OF At 68째F At 1200째F, after irradiation Maximum allowable strain, in./in. Porosity Accessible to salt at 150 psi Accessible to gas Poisson's ratio Young' s modulus

1.7 X 10" 3 x 10-6

0.001

0.001 0.01 0.4

With grain Across grain

0.005

1.25 x lo6

3 Y lo6 1.5 x 1Q6

Helium permeability at 3OoC, cm2/ s ec Diffusivity of xenon at 1200*F, cm2/sec Specific heat Btu/lb*"F

1 0 ' 6

0.33

WSRE values are given for comparison and were mostly taken from reference (12).


89

Tsble 4.6 (continued)

Assumed Properties of MSCR Fuel-Salt Mixture at 1200째F 1

2

71-16-13 29-11-60

68-23-9 32-19-49

66-29-5 38-29-33

Mixture No. Composition Mole 4 LiF-BeFz-ThFk Wt $I LiF-BeF2-ThF4

3

Liquidus temperature, OF Molecular weight

941

887

860

66 03

56.2

46.2

Density, lb/ft3 Viscosity, lb/ft*hr

215 6 24.2

190.1

163.0

21.0

18.9

Thermal conductivity, Btu/hr ft O F Heat capacity, Btu/lb*OF

2.67

2.91

3.10

0.318

0.383

0.449

9

Assumed Properties of MSCR Intermediate Coolant Salt Mixture at 1062'F* Composition Mole $ ' LiF-BeF2 Wt 46 LiF-BeF2

66-34

Liquidus temperature, OF

851

Molecular weight Density, lb/ft3 Viscosity, lb/ft*hr

33.14

Thermal conductivity, Btu/hr.ft* "F Heat capacity, Btu/lb*O F

3.5

52-48

120.5 20.0

0.526

*Reference ( 12) of the Bibliography.

Properties of INOR-8* Chemical composition Nickel, min Molybdenum *Reference ( 1 2 ) of the Bibliography.

wt

66-71

15-18

$J

(balance)


! ' i

90

Table 4.6 (continued)

Properties of INOR-8 (continuedl Chemical composition (continued)

Wt $ ' (continued)

i

Chromium

6-8

Iron, max Manganese, max

5

Silicon, max

1

Carbon

0.04-0.08

Miscellaneous, max

2

1

Physical Properties at 1200°F Density, g/cc

8.79

Melting point, OF

247&2 5 55

Thermal conductivity, Btu/hr*ft.OF at 1200°F

11.7

Young's modulus, psi at 1300°F Specific heat, Btu/lb. OF at 1200°F

25 X lo6

Coefficient of thermal expansion, 1/"F at 1200°F

7.8

Maximum allowable stress, psi at 1200°F

6000

Maximum allowable stress, psi at

3500

I30O0F

0.1385 X


91 5. 5.1

FUEL PROCESSING Reprocessing System

Fluorination of spent f u e l from t h e MSCR t o recover isotopes of uranium, followed by discard of the c a r r i e r s a l t (cantaining LiF, BeF2, ThF4, and f i s s i o n product fluorides), was selected as t h e method of processing for several reasons:

( a ) The fluorination process is w e l l adapted f o r

future integration with t h e reactor plant (sharing shielding and maintenance f a c i l i t i e s and personnel) with appreciable p o t e n t i a l reductions i n f u e l cycle cost (Sec. 6.9);

( b ) t h e necessity f o r holding t h e spent f u e l

f o r decay of r a d i o a c t i v i t y is eliminated; ( c ) no development is required; t h e plant may be designed and costed on the basis of current technology. I n order t o use t h e next most applicable process, Thorex, it would be necessary t o develop s p e c i a l head-end and tail-end s t e p s f o r converting spent f u e l from t h e fluoride t o t h e n i t r a t e and back again.

Also it would

be necessary t o hold t h e spent f u e l p r i o r t o processing f o r not less than 120 days (to average t h e equivalent of 90 days of cooling).

Further, t h e r e

is some uncertainty concerning t h e effect of fluoride ion on t h e chemistry

of t h e aqueous separations. On t h e other hand, t h e thorium could be recovered, and perhaps also t h e carrier salt could be recovered free from contamination with rare e a r t h isotopes.

1% would, however, be contaminated

with t h e isotopes Cd, Sr, Ag, C s , Se, Ba, and Te. The processing c o s t s f o r t h e MSCR f u e l have been estimated f o r both processes. 5.2 A central plant capable of processing about 30 f t

day (1 tonne Th/day) was selected â&#x201A;Źor costing, servicing about 20 referenc

3

of f u e l s a l t per

This plant is capable of

esign MSCR's having a t o t a l capability of

20,000 Mwe.

Component design

plant layout, and associated costs for t h e plant

described herein were adapted from a design and cost study of an on-site plant prepared by Carter, Milford, and Stockdale in a p r i o r study (21).


92 Due allowance f o r f u e l transport t o a c e n t r a l location was made, together with other adjustments for difference i n capacities, elimination of protactinium recycle, etc.

5.2.1

Process Design The s t e p s of t h e proposed fluorination process are indicated i n

Fig, 5.1.

Spent f u e l is transferred from t h e shipping containers t o pre-

fluorination storage by applying gas pressure o r by siphoning.

The f u e l

is then introduced batch-wise i n t o t h e f l u o r i n a t o r s where it is treated with elemental fluorine, possibly diluted with some i n e r t gas, a t about 1000째F. The e f f l u e n t UF6 is absorbed i n beds of NaF a t 2OO0F, is l a t e r desorbed a t 7OOOF and collected i n cold t r a p s a t -4SoF, Periodically t h e cold t r a p s are warmed, and t h e decontaminated UF6 i s collected i n cylinders. The fluoride v o l a t i l i t y process does not provide f o r recovering thorium or any of t h e components of t h e carrier s a l t . fluorination t h e LiF-BeF2-ThF4-FP

Consequently, after

m e l t is drained i n t o interim waste s t o r -

age, and l a t e r transferred t o permanent waste storage such as, f o r example, i n a s a l t mine.

The interim storage period has been taken t o be 1100 days,

a value corresponding t o t h e most favorable economic balance between ons i t e and permanent storage charges for t h i s p a r t i c u l a r process. It w i l l be observed t h a t 233Pa is not recovered i n t h i s process.

An

analysis of Pa recovery versus discard disclosed t h a t additional process equipment and building space requirements made t h e recovery of Pa uneconomical. A major problem i n t h e design of vessels which contain i r r a d i a t e d s a l t

is t h a t o f heat removal.

Volumetric heat release rates are high, and t h e

temperature of t h e heat source is considerably greater than t h a t of conventional heat sinks (such as cooling water from r i v e r s or wells). design evolved, heat is transferred across an a i r gap i n t o water,

In the The

principal heat t r a n s f e r mechanism is radiation; convection accounts f o r perhaps 5 t o 10 per cent of t h e t r a n s f e r .

This arrangement, i n addition

t o controlling t h e heat t r a n s f e r rate a t t o l e r a b l e l e v e l s , provides isol a t i o n of t h e coolant from t h e molten s a l t so t h a t a leak of e i t h e r stream through i t s containment wall does not contaminate t h e other stream.


c ORNL-LR NO. 74670

i I

I

I

I

I

I

I I

I

fI

UF. Recycle to

Reactor

i! d

I

4

I b

I I I

Fig. 5.1.

MSCR Fluoride Volatility Fuel Processing Plant.


! i -

94

1

fl

W-b The fluorinators are cooled by c i r c u l a t i n g a i r through t h e cell,

In

t h e case of t h e prefluorination storage tanks , radiation from storage vessels t o a concentric tank i n a water bath provides s u f f i c i e n t cooling f o r 5-day-old f u e l s a l t , Wherever possible, t h e equipment was patterned after t h a t used i n t h e

O W L V o l a t i l i t y P i l o t Plant described by Milford (661, Carr (181, Cathers e t a l , (22, 23). 5.2.2

Shipping Shipment of i r r a d i a t e d f u e l is made i n a lead shielded carrier shown

i n Fig, S 0 2 , The cask is equipped with a water-cooling system which is able t o absorb decay heat radiated from t h e s a l t container and t o d i s s i p a t e

*

t h i s heat t o t h e atmosphere v i a finned tube exchangers fastened t o t h e out-

a

s i d e of t h e cask.

Heat t r a n s f e r may be e i t h e r by boiling t h e water i n t h e

inside jacket followed by condensation i n t h e outside exchanger, or by n a t u r a l convection. The shipping container ( a l s o used for waste disposal), is designed t o hold %lo f t 3 s a l t which is about s i x days accumulation a t t h e r e a c t o r discharge r a t e of 1.67 f t3/day. For t h i s design it has been assumed t h a t t h e processing plant is located 500 miles from t h e reactor s i t e and t h a t shipment w i l l be made by r a i l .

The round t r i p , including f i l l i n g and emptying,

is anticipated t o t a k e 10 days. is approximately 5 days.

The average age of spent f u e l a t shipment

5.2.3

Prefluorination Storage Tanks One hundred and s i x t y 30 f t 3 tanks receive up t o 120 days supply of

spent f u e l from each of a number of r e a c t o r sites.

When a 120-day batch

is completed, t h e tanks are removed from t h e i r cooling jackets t o t h e 3 t r a n s f e r area where t h e material is transferred t o 6-ft metering t r a n s f e r tanks, From these tanks t h e spent f u e l can be transferred by gravity or i n e r t gas pressure t o t h e fluorinators, The cooling jackets f o r t h e prefluorination storage tanks are s t a i n less s t e e l tanks 2,2 f t i n diameter by 1 2 f t high immersed i n a water bath, Cooling is achieved principally by radiation from t h e storage tank t o t h e

h

wc


95

ORNL-DWG. 65-7911

J

Lead Shielding (14.5 t h i c k )

i

Finned Tube Condenser or Heat Exchanger

I

X

Salt 1 Container Fig. 5.2.

t

Jacket for H 2 0 Circulation I r r a d i a t e d Fuel Shipping Cask.


96 The t r a n s f e r tanks are equipped with jackets cooled by c i r c u l a t i n g

jacket,

water. Electric heaters provide preheating p r i o r t o t r a n s f e r operations, should heating be required, 5.2.4

Fluorinator The design shown i n Fig. 5,3 has been successfully operated i n t h e

ORNL Fluoride V o l i t i l i t y P i l o t Plant (18).

chamber is a de-entrainment section,

Surmounting t h e fluorination

The lower chamber i s surrounded by an

electrically-heated furnace while t h e upper is heated with electrical s t r i p heaters.

Five u n i t s are required, each having a capacity of 6 ft3.

The

corrosion rate is about 1 m i l per hour of fluorination time; hence t h e fluorinator must be inexpensive and accessible f o r frequent replacement, It w a s designed t o d i s s i p a t e decay heat and heat of reaction t o t h e atmos-

phere i n the c e l l through a w a l l 1/2 in. thick a t a temperature of 9OOOF. The preferred materials of construction are e i t h e r INOR-8 or Alloy 79-4 (70 per cent N i , 4 per cent Mo, 17 per cent Fe),

L-Nickel has been used

f o r f l u o r i n a t o r construction but is susceptible t o intergranular attack. Spent f u e l is fluorinated batchwise a t about 1000째Fo It takes about

a 6 f t 3 batch, In current p r a c t i c e t h e attack of fluorine on t h e vessel is severe, The high rate of 6 hours t o v o l a t i l i z e t h e uranium (99.9+%) from

corrosion is believed t o r e s u l t f r o m t h e combined action of l i q u i d salt and gaseous fluorine phases. under investigation.

However, several l i n e s of improvement are

These include t h e use of t h e "frozen wall" f l u o r i -

nator (35) wherein a l a y e r of s o l i d salt is maintained on t h e v e s s e l wall by proper control of the cooling, and t h i s l a y e r protects t h e w a l l . Another approach consists of spraying t h e molten s a l t i n t o a r e l a t i v e l y cool atmosphere of fluorine.

Uranium hexafluoride is formed i n and rapidly

removed from t h e microdroplets which then cool and freeze before they s t r i k e the w a l l .

Not only are t h e w a l l temperatures lower, but t h e r e is no

l i q u i d phase i n contact with t h e w a l l . The fluorides of some f i s s i o n products, notably Mo, Zr, Nb, Cs, Ru, and T e are v o l a t i l e and accompany t h e UF6.

These are separated from t h e

uranium i n t h e CRP t r a p and NaF absorber described below.

It is not ex-

pected t h a t t h e fluoride of Pa w i l l be v o l a t i l e under conditions specified,


97 ORNL-LR-DWG 39150-R2

THERMOCOUPLE WELL

FURNACE LINER

-

TREPANNED SECTION (4-in. OD) -FLUORINATION CHAMBER (16-in. OD ) DRAFT TUBES

WE INLET

..

.

. _ I _ .


98 Protactinium w i l l remain i n t h e barren s a l t and be l o s t t o waste.

The l o s s

amounts t o about 10 g/day i n t h e reference design reactor, and studies have shown t h a t it is not economical t o recycle t h e barren salt t o t h e f l u o r i nators after holding it t o allow t h e Pa t o decay,

The cost increase of

additional fluorinator capacity and interim storage vessels more than off-

sets t h e value of t h e 233U recovered, CRP Trap and NaF Absorbers

5.2.5

After leaving t h e fluorinator, UF6 and accompanying f i s s i o n product

fluorides pass i n t o a two-zoned NaF absorption system.

The first, called I

t h e complexible radioactive products (CRP) t r a p is operated a t about 400OC

ana removes fluorides of chromium, zirconium, niobium, cesium, stroxtium, and rare earths, as w e l l as entrained s a l t p a r t i c l e s .

Uranium hexafluoride

is not absorbed here, but i s absorbed i n t h e second zone operated a t 100째C, along with t h e fluorides of molybdenum and ruthenium, and traces of others. Some ruthenium carries through i n t o the fluorine r e c i r c u l a t i o n system. Uranium is recovered from t h e beds by desorption a t 4o0째c0

It is

collected i n cold t r a p s described below, The stationary bed absorber, shown i n Fig. 5,4, contains j u s t over one cubic foot of NaF.

Six unites are required.

Each is mounted i n a

lightweight, low-heat capacity electric furnace which opens on hinges for removal of t h e absorber.

A cooling-heating tube 2-1/2 inches OD carrying

coolant and containing electric heaters extends through t h e center of t h e bed.

An i n t e r i o r c y l i n d r i c a l b a f f l e forces t h e process stream t o follow a

U-shaped path through t h e bed, Design l i m i t a t i o n s arise i n t h e rate at which t h e bed temperature can be cycled and t h e bed thickness,

The granular bed i s a r a t h e r e f f e c t i v e

i n s u l a t o r and m u s t be made i n t h i n sections t o f a c i l i t a t e heating and cooling.

The absorbers therefore have l a r g e length-diameter r a t i o s , When t h e bed becomes saturated with f i s s i o n products, t h e absorber is

removed from t h e furnace, emptied, and recharged remotely i n a 4-5 day cycle

.


99

ORNL-LR- DWG 39257

COOLING AIR INLET rOUTLET L I" % !/,-in. NPSSCHED 40

ELECTRICAL RODTYPE HEATERS, 4000 - W TOTAL

$8 x 98-h. DIA

NaF PELLETS

SIL-0-CEL INSULATING POWDER

/

6-in. NPS SCHED 40-

MATER I A L: INCONEL

2'/2-in. OD x '4-in. WALL

'1'4 - in. PLATE'

THERMOCOUPLE WELLS Fig. 5.4. Sodium Fluoride Absorber.

.


, i

-i

w-t

100 Cold Traps

5,2.6

These are similar t o those used i n t h e ORNL Fluoride V o l a t i l i t y P i l o t Plant as shown i n Fig, 5 , 5 .

Two t r a p s are mounted i n series,

The first

is operated a t about -4OOC and t h e second a t -60째C, Adequate surface f o r rapid transfer of heat and collection of s o l i d

UF6 must be provided.

The components must have small thermal i n e r t i a so

... defrosting, t h e t r a p s t h a t t h e temperature may be changed quickly, During are heated t o 90째C a t 46 psia t o allow UF t o m e l t and drain t o collection

6

cylinders. Reduction Reactor

5,2.7

The reduction of UF6 t o UF4 is accomplished i n a r e a c t o r patterned

after t h a t described by Murray ( 7 0 ) and consists of a 4-in. d i m by 10-ft high column.

Uranium hexafluoride and flourine are mixed w i t h excess H

a nozzle a t t h e top of t h e reaction chamber.

2

in

The uranium is reduced i n t h e

H -F 2 flame, and f a l l s t o t h e bottom of t h e chamber where it is collected i n molten s a l t of s u i t a b l e composition, Gaseous materials are discharged

- 2

through a f i l t e r .

The reactor has a capacity of 10-15 kg of UF6 per hour.

Losses are very low and t y p i c a l l y are less than 0.1 per cent. Transfer Tanks

5.2.8

Stripped fuel is drained f r o m t h e fluorinators i n t o t r a n s f e r tanks (two each) from whence it is d i s t r i b u t e d t o interim waste storage tanks, The t r a n s f e r tanks have capacities of 60 f t 3 ,

Decay heat is radiated from

t h e surface of t h e tank through a 1/2 in. a i r gap t o a water-cooled jacket. While being held i n the t r a n s f e r tanks, t h e salt is t r e a t e d with He or other i n e r t gas t o remove traces of F2 of HF t h a t would increase corrosion i n the waste storage tanks.

Waste Storage Tanks

502.9

Waste s a l t is stored i n s t a i n l e s s s t e e l shipping cylinders 2 f t i n diam by 8 ft long.

These are placed a t t h e bottom of steel thimbles

2,75 f t i n diam and f i f t e e n feet long which d i p i n t o a water-filled

canal.

Heat is dissipated by radiation and convection across t h e 4-inch


101

ORNL-LR-WG 1909t

-

FILTER CARTRIDGES

. 1

ERANT TUBES (4)

12

INLE

0 12 --SCAM IN INCHES

PRIMARY COLD TRAP

Fig. 5.5.

Primary Cold Traps.

24


102 a i r gap, through the thimble wall and i n t o t h e water.

After cooling, t h e

tanks a r e shipped t o a s a l t mine f o r permanent storage,

5.2.10

Freeze Valves Flow stoppage

Conventional valves cannot be used with molten salts.

is achieved by freezing a plug of s a l t i n a section of a l i n e with a j e t of cooling a i r .

Electric heaters are used t o thaw t h e plug when flow is

A freeze valve for t h e MSRE is shown i n Fig, 5.6.

desired.

5 2 11 Samplers

The apparatus pictured i n Fig. 5,7 is being t e s t e d for use with t h e

MSRE (12).

Essential features are t h e h o i s t and capsule f o r removing t h e

sample from t h e vessel; a lead-shielded cubicle with manipulator, heating elements and service piping, and a transport cask for removing t h e sample

f r o m t h e process area.

The sampling cubicle is mounted on t h e c e l l biolo-

g i c a l s h i e l d i n an accessible area. 5,2,12

Biological Shield

Calculations were made using t h e Phoebe program f o r t h e IBM 704 com-

I n t h e study by Carter, Milford, and Stockdale (211, on which t h e present estimate i s based, spent f u e l was brought t o t h e processing plant puter,

immediately a f t e r removal f r o m t h e reactor, and t h e s h i e l d was accordingly made q u i t e thick.

I n t h e present instance, t h e f u e l is cooled a t t h e pro-

cessing plant f o r not less than 90 days so t h a t t h e shielding requirements

are not as extreme.

However, i n order t h a t t h e central plant have more

general u t i l i t y , no reduction i n s h i e l d thickness was made. 5.2,13

Process Equipment Layout

Process equipment is l a i d out according t o t h e major process operations: prefluorination storage, fluorination, t r a n s f e r , NaF absorption, cold t r a p s and product collection, UF6 storage,

+

UF,, reduction, and interim waste

Equipment is grouped i n cells according t o a c t i v i t y l e v e l and i n

an arrangement t h a t minimizes distances between vessels, of molten s a l t are required i n t h e processing sequence,

Three t r a n s f e r s

G-


103


f

i

-i I

i

Fig. 5.7.

Radioactive Material Sampler.

us

-


105 I n t e r i m waste s t o r a g e v e s s e l s are l o c a t e d a d j a c e n t t o t h e processing

area i n a l a r g e canal. T o f a c i l i t a t e remote maintenance, v e s s e l s are arranged so t h a t a l l equipment i s a c c e s s i b l e from above, and a l l process and s e r v i c e l i n e s can be connected remotely.

Over-all b u i l d i n g space i s d i c t a t e d by remote

maintenance c o n s i d e r a t i o n s r a t h e r t h a n by a c t u a l v e s s e l s i z e , 5,2,14

P l a n t Layout

I n o r d e r t o e s t a b l i s h uniformity i n c o s t e s t i m a t i o n of n u c l e a r power p l a n t s , t h e Atomic Energy Commission has s p e c i f i e d c e r t a i n ground r u l e s ( 5 2 ) covering topography, meteorology, climatology, geology, a v a i l a b i l i t y of l a b o r , accounting p r o c e d w e s , f i x e d charge r a t e s , etc.

These ground

r u l e s were used i n t h i s s t u d y , Advantage was taken of a design study and o p e r a t i n g experience w i t h a remotely maintained r a d i o a c t i v e chemical p l a n t r e p o r t e d by Farrow (32) t o o b t a i n o v e r - a l l p l a n t arrangements, as shown i n Fig, 5.8, The h y p o t h e t i c a l s i t e l o c a t i o n is 500 miles from t h e r e a c t o r s i t e , The p l a n t i s l o c a t e d on a stream t h a t i s navigable by b o a t s having up t o 6-ft d r a f t ,

There is convenient highway and r a i l r o a d access.

l o c a t e d on l e v e l t e r r a i n i n a grass-covered f i e l d .

The p l a n t is

The e a r t h overburden is

8 f t deep with bedrock below.

5,2,15

Capital Cost Estimate

The c o s t of t h e f u e l processing p l a n t was apportioned among t h r e e principal categories:

b u i l d i n g c o s t s , process equipment c o s t s , and a u x i l -

i a r y process equipment and s e r v i c e s c o s t s .

The b u i l d i n g c o s t s included

such items as s i t e p r e p a r a t i o n , s t r u c t u r a l materials and l a b o r , permanently i n s t a l l e d equipment, and m a t e r i a l and l a b o r f o r s e r v i c e f a c i l i t i e s ,

Proc-

ess equipment c o s t s were c a l c u l a t e d f o r t h o s e t a n k s , v e s s e l s , f u r n a c e s , and

similar items whose primary f u n c t i o n i s d i r e c t l y concerned with process operations,

Process s e r v i c e f a c i l i t i e s are items such as sampling f a c i l i -

t i e s , process p i p i n g and process i n s t r u m e n t a t i o n which are i n t i m a t e l y assoc i a t e d with process o p e r a t i o n s .


I

m

m

*

-

I

r


107 Accounting Procedure, -The

accounting procedure s e t f o r t h i n t h e

Guide t o Nuclear Power Cost Evaluation (52) was used as a guide i n t h i s estimate,

This handbook was written as a guide f o r cost-estimating reactor

plants, and t h e accounting breakdown is not s p e c i f i c for a chemical processing plant,

Where necessary the accounting procedures of the handbook

were augmented. Process Equipment. - A

large number of process vessels and auxiliary

equipment i n these plants is similar t o equipment previously purchased by

ORNL f o r t h e fluoride v o l a t i l i t y p i l o t plant f o r which cost records are available,

Extensive use was made of these records i n computing material,

fabrication and over-all equipment costs.

In some cases it was necessary

t o extrapolate the data t o obtain costs f o r l a r g e r vesselso Items t h a t

were estimated i n t h i s manner include t h e fluorinators, furnaces, NaF absorbers, and CRP traps.

The cost of t h e UF6-to-UF4 reduction unit was

based on a u n i t described by Murray (70).

The u n i t had a l a r g e r capacity

than was needed f o r these plants, but it was assumed t h a t t h e required unit would have about the same over-all cost,

Refrigeration equipment and

cold t r a p s were estimated from cost data on ORGDP and ORNL equipment, For vessels and tanks of conventional design, t h e cost was computed from the cost of material (INOR-8 f o r most vessels) plus an estimated fabr i c a t i o n charge, both charges being based on t h e weight of t h e vessel,

A

summary of values used i n estimating process vessels by weight is given i n Table 5.1.

Some items of process equipment were of s p e c i a l design and sig-

n i f i c a n t l y different from

essels f o r which cost d a t a were available.

nation storage tanks, t h e high fabrication For t h e s h e l l s of t h e pre by comparison with an available shop esticost values shown were obtain mate f o r a similar vessel, Auxiliary process items such as process piping, process e l e c t r i c a l service, instrumentation, sampling connections and t h e i r i n s t a l l a t i o n were not considered i n s u f f i c i e n t design d e t a i l t o permit d i r e c t estimation, cost was assigned t o these items which was based upon previous experience i n design and cost estimation of radiochemical processing plants,

In

assigning these costs cognizance was taken of the fact t h a t t h e plant is

h) E

remotely maintained,

A


*

108 4i

Table 5.1,

Vessel, Pipe, and Tubing Costs

-~ _ _

--

~~

~

~

Alloy 79-4

INOR-8 $/ft

s/lb

Stainless S t e e l 304 0.65

Metal Cost

Fabrication Cost: Prefluorination INOR t r a n s f e r tanks 3 and 4 Fluor inators Transfer vessels Waste storage vessel Waste storage thimbles UF dissolvers 1 / Y in. OD x 0,042 wall tube 1 in. IPS, Sch, 40 pipe 1-1/2 in. IPS, Sch. 40 pipe

Buildings. -The

6.06 30.05 41.67

3.50 26.40 16.04

13.71

building estimate shown i n T a b l e 5.2 included t h e

cost of land acquisition, s i t e preparation, concrete, s t r u c t u r a l steel, painting, heating, ventilation, a i r conditioning, elevators, cranes, service piping, laboratory and hot c e l l equipment, etc.

The individual

c o s t s were calculated using current data for materials and labor, and were based on t h e drawings prepared. Process Equipment Capital Cost. -Process

equipment c a p i t a l costs f o r

t h e two fluoride v o l a t i l i t y plants are presented i n Table 5.3,

These costs

are t h e totals of material, fabrication and i n s t a l l a t i o n charges. Building Capital Cost. -As mentioned above, process equipment and buildings were t h e only items considered i n s u f f i c i e n t design d e t a i l t o permit d i r e c t estimation.

The remainder of t h e c a p i t a l c o s t s were esti-

mated by extrapolation from previous s t u d i e s of radiochemical processing plants.

The fact t h a t t h e plant is remotely maintained was an important

factor i n estimating process instrumentation and electrical and sampling connections.

These items are expensive because of counterbalancing,

spacing, and a c c e s s i b i l i t y requirements. Construction overhead fees were taken a t 20 per cent of d i r e c t

materials and labor for a l l buildings,installed process equipment, piping,

G


109 Table 5.2.

Fluoride V o l a t i l i t y Processing Building Costs 30 ft3/day Plant Capacity Materials

Total

Labor

Receiving Area Excavating and b a c k f i l l Concrete, forms, etc. Structural s t e e l , etc. Roofing Services

$ 38,000

108,000 58,000 13,000 63,000

$ 18,000 $

56,000 241,000 107,000 29,000 102,000

133,000 49,000 16,000 39,000 $

535,000

$ 85,500 $

639,000 235,000 75,600

269,700 1,159,000 512,000 140,500

160,000 188,000 255,000 2,000

550,000 489,000 1,120,000 42,000

Processing Cells Excavating and b a c k f i l l Concrete, forms, etc. S t r u c t u r a l steel, etc. Crane area roofing, painting, etc. Crane bay doors Services Bldg. movable equipment Viewing windows

$184,200 520,000 277,000 64,900 390,000 301,000 865,000 40,000

$4,282,200 Waste Storage Excavation and b a c k f i l l Concrete, forms etc. S t r u c t u r a l steel, etc. Crane area roofing Painting Services Bldge movable equipment

,

$ 95,000

332,000 400,000 86,500 44,500 565,000 225,000

$ 44,400 $

496,000 404,000 100,800 * 44,500 289,000 30,000

139,400 828,000 804,000 187,300 89,000 854,000 255,000

$3,156,700 Operations and Laboratories $ 72,200 Excavation and b a c k f i l l 82,600 Concrete forms, etc. 204,700 S t r u c t u r a l steel, etc. 8,500 Roofing 79,100 Super structure Misc. s t r u c t u r a l material 31,900 Services 352,000 Misc. equipment 300,000

$ 33,400 $

115,100 43,200 4,300 27,100 35,100 276,000 43,500

105,600 197,700 247,900 12,800 106,200 67,000 628,000 343,500

$1,708,700


110 Table 5.2.

Continued

Materials

Tbtal

Labor

Outside U t i l i t i e s Cooling tower, motors, Pumps, piping Water resevoir, pumps, piping Fire protection (house 4 equipment 1 Yard l i g h t i n g Boiler house steam heating (4,000 kw a t $75/kw) A i r compressor system Steam d i s t r i b u t i o n 4 condensate return Cooling water supply & return Water supply connection Process drain l i n e s Sanitary sewer connections Radioactive hot drain connections C e l l v e n t i l a t i o n connect i o n s t o stack Off-gas connectiws t o stack Storm sewer system Electrical substation 4 l i n e s (3000 kw a t $60/kw) Stack (200 f t ) Guard house and p o r t a l s Autos, trucks, crane, b u l l dozer

$

70,000 300,000

35,000 5,000

300,000

10,000 3,500

40,000 1,700 3,500

3,700

12,000 9,000

10,000 16,000 180,000 50,000 5,000

so ,000 $1,104,400

Land and Land Improvements Land (160 acres a t $lOO/acre) Leveling & grading Topsoiling and seeding Fencing, ( 2 miles a t $4/ft) Railroad spur, 100 f t Asphalt roads 4 parking areas

16,000 50,000 20,000 44,000 20,000 200,000 $

350,000


111 Table 5.3.

I n s t a l l e d Cost of Fluoride V o l a t i l i t y Process Equipment 30 ft3/day Plant Capacity Number

Description

cost

Receiving Cooling jackets f o r shipping tanks

160

2,75 I D by 15-ft high, carbon s t e e l

$

120,000

Thermocouples, radiation monitors, etc.

480,000

2-ft x 2 - f t ; INOR-8; 0.375 in. s h e l l , 0.5 i n head

100,000

2

2,7 f t x 3 f t ; 50 kw

7,000

Fluorinators

5

1.75 f t d. by 9 f t high; 6 f t 3 s a l t ; Alloy 79-4; 0,s in. s h e l l , 0.5 i n . head

40,000

Furnaces

5

2.7 f t d. x 4 f t h; 75 kw

16,000

Movable bed absorber

5

6 in. d x 4 f t h

25,000

Instrumentation Prefluorination Storage Transfer tanks

2

.

Furnace Fluorination

Absorption 6 i n . sch. 40 pipe, 603 f t long

150,000

15

-4OOC u n i t s , copper

112,500

15

-75OC units, copper

37,500

Absorbers with furnaces

30

Cold t r a p s NaF chem, t r a p

5

6 in. sch, 40 pipe x 6 f t ; heated

3,000

Vacuum pump

1

50 cfm displacement

3,000

1

10 kg/day capacity; Inconel

Reduct ion and Compounding Reactor

66,150 3

Dissolver

1

2,7 f t d. x 2.7 f t h; 12 f t s a l t ; INOR-8; 0.5 in. s h e l l

5,500

Heater

1

3.4 f t d. x 3,7 f t h; 71 kw

6,000

S a l t make-up tank

2

3*4 f t d o x 6.7 f t h; INOR-8; 40 f t 3 capacity

26,000

Heater

2

4.1 f t d. x 7.7 f t h; 178 kw

34,000


. 112 T a b l e 5.3.

Continued

Number Transfer Tanks

2

Description

cost

3 4.5 f t d. x 4,s f t h; 60 f t of s a l t ; INOR-8 with heaters

$

100,000

Waste Storage Shipping and storage tanks

As Needed

2 f t I D by 8 f t h; s t a i n l e s s 3 s t e e l 304 L; 10 f t salt; 0.25 i n , s h e l l and head

Thimbles

1200

2.75 f t d o x 15 f t h; ordinary s t e e l 304 L; 0.1875 s h e l l

Included with operating expenses 600,000

Miscellaneous Equipment h f r i g e r a t i o n unit

2

50,000 Btu/hr a t -4OOC

10,000

Refrigeration u n i t

2

8,000 Btu/hr a t -75OC

10,000

HF disy:sal u n i t F2 su- :.!I system

1

2.8 f t d. x 5.3

5

Tank and t r a i l e r

500

ft h

35,000

Total Process Equipment

instrumentation, electrical and other d i r e c t charges.

$1,987,150

This rate is higher

than current charges f o r t h i s type of construction and estimates,

Archi-

tect engineering and inspection fees were taken as 15 per cent of a l l

charges including construction overhead. A summary of t h e c a p i t a l cost estimate

is presented i n T a b l e 5.4.

?

i


113 *-

Table 5 0 4 0 Summary of Capital Cost Estimate for Molten-Salt Reactor-Fuel Fluoride V o l a t i l i t y Processing Plant Capacity

-

30 ft3 of Salt/Day

Receiving Area Processing Cells Waste Storage Operations and Laboratories Outside U t i l i t i e s Land and Improvements Process Equipment Process Piping Process Instrumentation Process Electrical Connections Sampling Connections

$

30,000 ~~

Total, I n s t a l l e d Equipment and Buildings General Construction Overhead a t 20% s u b t o t a l Architectural Engineer, etc., a t 15% s u b t o t a l Contingency a t 20% s u b t o t a l I n t e r e s t During Construction, 9.3% of subtotal

~

$13,718,150 2,743,660 2,469,294 3,786,250 2,112,728

$24,830,082

Total

5.2.16

535,000 4,282,200 3,156,700 1,708,700 1,104,400 350,000 1,987,150 320,000 205,000 39,000

ODeratinrr and Maintenance Cost Estimates

The manpower requirements were estimated consistently with the procedures outlined i n t h e Guide (52); t h e r e s u l t s a r e l i s t e d i n Table 5.5.

Materials, u t i l i t i e s , maintenance materials, etc. were estimated by conside r a t i o n of the process s t e p s involved; t h e estimates are l i s t e d i n Table 5,6 together with a summary of t h e labor cost. The t o t a l operating cost was estimated t o be $4,040,850 annually, 5.2.17

MSCR I r r a d i a t e d Fuel Shipping Cost The shipping cask must accommodate a molten-salt shipping cylinder

having a volume of 10 f t

3

e

The cost was estimated from t h e weight which

was determined by t h e shielding requirements.

A u n i t cost of $l.OO/lb

fabricated was allowed, including charges f o r an INOR l i n e r and a condensing-water radiator.

Three casks were allowed so t h a t one might be a t t h e

r e a c t o r s i t e , a second a t t h e processing plant, an6 a t h i r d i n t r a n s i t . Results are l i s t e d i n Table 5 , 7 ,


-

1

i

114

-?

-Wâ&#x20AC;?! T a b l e 5.5. Operating Manpower Estimates for 30 ft3/day Fluoride V o l a t i l i t y Plants

-

cost ($/pear) Management Manager Assistant manager Secretary

2

18,000 15,000 10;000

4

43,000

1 4 12 12 2

12,000 30,000 66,000 60,000 9,600

31

177,600

1 2 12 3 8 1 2

10,000 16,000 69,600 18,000 46,400 4,350 8,700

29

173,050

1 6 10 6

8,000 39,000 52,000 28,800

23

127,800

1 1

-

Production Superintendent S h i f t supervisor Oper a t or Helper Secretary

-

Maintenance Superintendent Mechanical engineer Mechanic Machinist Instrument man Clerk Storeroom keeper *

-

Laboratory Supervisor Chemist Technician Helper

-

Health Physics Supervisor Monitor Clerk Records keeper

1 4 1 1

8,000 20,800

7

36,400

4,000

- 3,600

12,

--


115 Table 5.5.

Continued

No.

cost ($/year)

Account a b i l i t y Engineer Clerk

1 1

7,000

2

11,000

2 4

3 1

16,000 36,000 15,900 4,500

10

72,400

-

4,000

Engineer i n g 5

Mechanical engineer Chemical engineer Draftsman Secretary

-

General Office Manager Account ant Payroll c l e r k Purchasing agent Secretary

1 1 2 1

5,000 4,800 8,000

2

8,000

7

30,600

-

4,800

Miscellaneous Guard Fireman Receptionist Laundry worker

8

32,000

4 1 3

16,000

Nurse Janitor

1 3

Total

4,000

10,800 4,800

- 10,800

20

78,400

133

750,250


116 Table 5.6.

Shipping

Fluoride V o l a t i l i t y Plant Direct Annual Operating Cost

- Storage Tanks (50 a t

$2,500)

$ ,125,000

Chemical Consumption Fluorine ( a t $2 O O / l b KOH ( a t $O.lO/lb) Hydrogen (at $2 O O / l b ) NaF (at $O.lS/lb) Nitrogen ( a t $O.o5/ft HF (at $Oo20/lb) Graphite ( a t $0 1 5 / l b Miscellaneous

.

.

120,000 16,600 4,500 300 3,400 6,100 1,100 5,300 157,300

Utilities

.

E l e c t r i c i t y ( a t $0 O l / k w h r Water (at $0.015/1000 g a l ) Heating (based on steam a t $0.25/1000 lbs)

362,000 5,700 8,500

376,200

Labora Operating Laboratory Maintenance Supervision Overhead ( a t 20% of above)

406,400 127,800 173,050 43,000 150,000 900,250

Maintenance Materials Site b C e l l structures and buildings Service'and u t i l i t i e s Process equipment'

10,000 76,000 78,800 243,300

Total Direct Operating Cost

b82,lOO $2,040,850

a

Summarized from T a b l e 5 . 5 .

bBuilding services excluded. C

Includes process equipment, process instrumentation and sampling.

h

bi c


117 T a b l e 5,7.

MSCR Irradiated F u e l Shipping Cask Data and Shipping Cost*

Cask weight Cost of cask Number of casks S a l t volume i n shipment Age of salt a t shipment Days s a l t accumulation En shipment Round t r i p distance Round t r i p time Method of shipment Number of shipments per f u l l power year Freight rate

Unit shipping cost

100,000 lbs $100,000

3 10 f t 3 20 days 6 days 1000 miles 10 days Rail 50

$2.40/100 Ibs1000 miles $330/ft3 salt

*Data and cost adapted from Reference 20.

5.2.18

MSCR Unit Processing Cost

The various bases and contributions t o t h e u n i t processing cost are collected i n Table 5 , 8 for t h e fluoride v o l a t i l i t y c e n t r a l plant process-

For t h e reference design f u e l containing 32 kg of thorium per ft3, t h e u n i t cost was $36.90 per kg of thorium. ing of MSCR fuel.

i

Although the cost was expressed i n terms of $/kg of Th, it should be remembered t h a t only isotopes of uranium are recovered.

Stripped s a l t ,

containing valuable thorium, lithium-7, and beryllium, as w e l l as f i s s i o n products, is discarded. Additional o r a l t e r n a t e processing would be required t o recover any of these components.


118 Unit Processing Costs, Central Fluoride Table 5.8. V o l a t i l i t y Processing Plant f o r MSCR Fuel ~~

~

~~

~

Capacity of plant

30 ft3/day

Reference design f u e l

32 kg Th/ft3

Annual charges : Capital ($24.8 million a t 15%) Operation and maintenance

$3,7 million $2,0 million

Daily charge (80% plant factor)

$19,5OO/day

Batch s i z e

188 f t 3 or 6000 kg Th

Processing time

6,3 days

Tm-around time

2 days

Processing plant cost

$26.60/kg Th

Shipping cost

$10.30/kg Th

Total processing cost

$36,90/kg Th

5.3

Thorex Central Plant

The reference plant described i n the Guide (52) is a c e n t r a l f a c i l i t y capable of processing 1000 kg Th/day with thorium discard or 600 kg Th/day with thorium recovery.

The plant was designed s p e c i f i c a l l y f o r thorium

metal or thorium oxide fuels; however, since other types of thorium fuels

were not s p e c i f i c a l l y excluded, it was assumed t h a t t h e plant would a l s o accept a fluoride-salt f u e l ,

I t was f u r t h e r assumed t h a t the fluoride f u e l

would be processed a t t h e same base charge as t h e metal or oxide fuel. This assumption amounted t o assigning t h e same charge t o a fluoride headend treatment as t o dissolution and feed preparation s t e p s f o r t h e other fuels.

The tail-end treatment for t h e conversion of Thorex n i t r a t e product

t o fluoride feed material was assigned a cost t h a t was thought t o be repres e n t a t i v e of t h e processing steps. c

w-


119

5.3.1

Head-End Treatment The head-end treatment shown schematically i n Fig, 5.9 has not been

demonstrated,

However, t h e chemical principles have been established (40)

by laboratory investigations of the s t a b i l i t y of fluoride salt fuels. I t is known (40) t h a t t h e oxides of uranium, thorium and beryllium are

very stable compounds having t h e indicated order of s t a b i l i t y U02

Be0

Tho2

and t h a t oxygen o r oxygen-bearing compounds must be eliminated from fluor i d e s a l t f u e l s t o insure t h e i r s t a b i l i t y o In t h e proposed head-end treatment, t h e draw-off from t h e reactor a t 500

-

5SOOC would be contacted i n a

spray with steam o r high temperature water (%2OO0C) t o p r e c i p i t a t e the oxides of uranium, beryllium, thorium, protactinium and some of t h e f i s s i o n products.

I t is believed t h a t rare earth oxides can be precipitated i n

t h i s manner.

Lithium fluoride is a very s t a b l e compound and would probably

not e n t e r i n t o reaction with water,

It should remain i n t h e system as L i F

and be frozen i n t o small c r y s t a l l i n e p a r t i c l e s . The hydrolysis would form l a r g e q u a n t i t i e s of HF which i n aqueous media is r a t h e r corrosive,

Therefore t h e selection of materials of con-

s t r u c t i o n f o r t h e precipator w i l l be a problem,

Disposal of HF can be

accomplished by d i l u t i o n with large volumes of a i r and dispersion from a stack o r by neutralization with an inexpensive base, Some cleanup of t h e HF stream w i l l be required because of v o l a t i l e f i s s i o n products,

Reuse of

t h i s HF i n t h e subsequent hydrofluorination s t e p (see Fig. 5 . 8 ) may not be f e a s i b l e because of water vapor i n t h e gas, The second s t e p i n t h e head-end treatment is dissolution of a l l t h e hydrolyzed components t h a t are soluble i n n i t r i c acid,

The oxides of

uranium, thorium, protactinium and rare e a r t h products should dissolve q u i t e readily,

Since lithium fluoride and beryllium oxide are q u i t e in-

soluble i n aqueous media, negligible amounts of these compounds should be dissolved, Also, it is almost certain t h a t some of t h e f i s s i o n products

w i l l be insoluble and remain with t h e lithium and beryllium.

Dissolution

should proceed smoothly because it has been shown by P i t t (73) t h a t t h e p a r t i c l e s i z e s produced when a molten s a l t is sprayed i n t o water are i n t h e micron and submicron range,


120

w-

ORNL-LR-OWG

off -gas HF

Molten

HNO,

Precipitation

-

Dissolution

71093

n

Filtrotfon

Aqueous

Solvent Extraction

I

111, Product 200.c

Make -Up

U4(W* Th ( NO,), QP)C@JO,)l

100.C

Wotte FP's

Be 0 ( FPk) 1

Woste +Discord

Nit rogen

Denitrot ion Reconstitute(

Fuel UF

.

Be Ft ThR U F4

Fig. 5.9.

Proposed Aqueous Processing for MSCR.

I

Product

I

.


121 -2-

c)-

D i s s o l u t i o n is followed by s o l i d - l i q u i d s e p a r a t i o n e i t h e r by f i l t r a tion or centrifugation,

Aqueous n i t r a t e s o l u t i o n s are f e d i n t o a f e e d

adjustment s t e p preceding s o l v e n t e x t r a c t i o n by Thorex and t h e i n s o l u b l e

material is r o u t e d t o h y d r o f l u o r i n a t i o n ,

A t t h i s p o i n t a p o r t i o n of t h e

s o l i d s can be d i s c a r d e d as a purge of f i s s i o n products t h a t remain i n t h e precipitate, 5.3.2

Solvent E x t r a c t i o n Decontamination of thorium and uranium can be accomplished by w e l l -

e s t a b l i s h e d Thorex procedureso Aqueous n i t r a t e s o l u t i o n s are evaporated u n t i l about 0,15 N a c i d d e f i c i e n t and f e d t o an e x t r a c t i o n column, . . )

In t h e

e x t r a c t i o n column both thorium and uranium are e x t r a c t e d i n t o an o r g a n i c phase ( t r i b u t y l phosphate) l e a v i n g t h e bulk of t h e f i s s i o n products i n t h e 5

aqueous phase, w i t h decontamination factors up t o 1 0

Waste from t h e first e x t r a c t i o n c y c l e c o n t a i n s a l l of t h e p r o t a c t i n i u m t h a t was i n t h e f e e d stream.

However, t h e amount is i n s u f f i c i e n t t o war-

r a n t r e c y c l i n g t h e waste a f t e r an a d d i t i o n a l decay period.

In the interim

between d i s c h a r g e from t h e r e a c t o r and chemical processing, t h e f u e l ages about 120 days so t h a t only about 15% o f t h e p r o t a c t i n i u m remains undecayed. The waste is given permanent storage i n l a r g e underground tanks. I n a f u e l r e c y c l e system such as t h e MSCR it is n o t n e c e s s a r y t o decontaminate f u r t h e r by t h e use o f a d d i t i o n a l e x t r a c t i o n c y c l e s .

The pres-

ence of 232U and 228Th w i l l make r e c y c l e f u e l too r a d i o a c t i v e f o r d i r e c t handling, r e g a r d l e s s ,

After e x t r a c t i o n , t h e r e f o r e , it is s u f f i c i e n t t o

p a r t i t i o n uranium and thorium i n a s t r i p p i n g column by t h e p r o p e r a d j u s t ment of o r g a n i c and aqueous flow rates.

I n t h i s o p e r a t i o n , thorium is

s t r i p p e d f r o m t h e o r g a n i c phase i n t o an aqueous phase; uranium remains i n t h e o r g a n i c phase,

A subsequent s t r i p p i n g o p e r a t i o n r e t u r n s t h e uranium

t o t h e aqueous phase.

5.3,3

u

The produce streams are r e s p e c t i v e l y Th(f103),+and

Tail-End Treatment Fuel reconstitution begins w i t h acceptance of the n i t r a t e products

from t h e s o l v e n t e x t r a c t i o n p l a n t .

I t i s necessary t o convert t h e n i t r a t e s


122 t o t h e fluorides.

I n t h e case of thorium t h i s is accomplished by pre-

paring t h e oxide i n a d e n i t r a t i o n process followed by hydrofluorination i n

a molten s a l t mixture,

Th (NO3 l4

The s t e p s are as follows: HF Thu2 i n molten sal?

steam d e n i t r a t i o n

woooc

ThF4

Steam d e n i t r a t i o n is an established procedure i n t h e sol-gel process (33) for preparing highly f i r e d , dense Tho2 or Tho2

- U02 f u e l ,

In t h i s

case t h e aqueous n i t r a t e s o l u t i o n from t h e Thorex process would be evapor a t e d t o c r y s t a l l i z e Th(N03)4; t h e crystals i n t u r n would be contacted with superheated steam f o r t h e a c t u a l d e n i t r a t i o n ,

Final preparation of t h e

f l u o r i d e has t o be accomplished i n a second high temperature operation i n molten f l u o r i d e salts.

Thorium oxide i s q u i t e i n t r a c t a b l e t o attack by

hydrogen f l u o r i d e under most conditions; however, i n t h e presence of molten f l u o r i d e s t h e r e a c t i o n w i l l occur, pounds, e.g., dissolution

.

The presence of o t h e r high valence com-

other thorium or uranium f l u o r i d e s , i n t h e melt abets t h e

The conversion of uranyl n i t r a t e t o t h e t e t r a f l u o r i d e is not as straightforward as t h a t of thorium because of t h e required valence change.

In t h e Excer process developed a t O W L , uranyl n i t r a t e from t h e last Thorex s t r i p p i n g column can be fed d i r e c t l y onto a cation exchange r e s i n (Dowex 50 W ) which absorbs t h e uranyl ion. After loading, t h e r e s i n is

,.

e l u t e d with aqueous hydrofluoric acid t o produce U02F2, which is reduced i n an e l e c t r o l y t i c cell.

c

The aqueous s o l u t i o n is allowed t o flow i n t o a

mercury cathode i n which UF4

0.75 H20 p r e c i p i t a t e s from t h e aqueous

phase,

The p r e c i p i t a t e is separated by centrifugation or f i l t r a t i o n and

dried,

Water of hydration is not tenaciously held, and moderate drying

conditions are s u f f i c i e n t t o expel it, A second method of converting U02(N03) t o UF4 i s t o reduce U(V1) t o U ( 1 V ) i n t h e presence of f l u o r i d e at 500

-

600째C0

In t h i s tail-end treat-

ment t h e two n i t r a t e s o l u t i o n s of thorium and uranium are mixed and cod e n i t r a t e d using superheated steam t o y i e l d Tho2 and U03,

are converted t o t h e f l u o r i d e s i n a 500

The mixed oxides

- 6OOOC molten f l u o r i d e s a l t bath

T

e-


123

+

by contacting with a gaseous mixture of H2

HF.

Uranium i s reduced by

t h e hydrogen and hydrogen f l u o r i d e t o t h e t e t r a v a l e n t form and dissolves

as UF4; thorium i s metathesized t o ThF4 which a l s o dissolves i n t h e melt. The processing cycle i s completed when t h e uranium f l u o r i d e is mixed with f l u o r i d e s of thorium, lithium, and beryllium oxide plus make-up feed and hydrofluorinated,

Beryllium recycled as Be0 dissolves as BeF2 when

hydrofluorinated i n molten f l u o r i d e solutions, 5,3.4

Processing Costs The u n i t processing c o s t s were computed according to t h e p r e s c r i p t i o n

given i n Guide t o Nuclear Power Cost Evaluation, Vol. 4, Section 460,

The

escalated d a i l y charge f o r operation of t h e Thorex plant w a s taken as $17,500 i n l a t e 1962. 600 kg/day.

Since t h e thorium was recycled, t h e capacity was

As described i n a previous section, t h e d a i l y increments of

f u e l withdrawn from t h e r e a c t o r were accumulated f o r 120 days and combined i n t o a s i n g l e processing batch. If t h e withdrawal r a t e exceeded 3

1.25 f t /day (40 kg Th/day), t h e batch s i z e exceeded 4800 kg of Th and reFor t h i s range of process times,

quired e i g h t days o r longer t o process. t h e "turn-around-time"

i s specified t o be eight days i n t h e Guide.

t h e t o t a l processing time is 120r/600

+

Thus

8 days (where r is t h e withdrawal

r a t e i n kg Th/day) and t h e t o t a l amount processed i s 1 2 0 r kg of Th.

The

cost f o r operating t h e Thorex plant i s thus 120r

+

Separation Cost = 120r

= (29.2

+

1167/r) $/kg Th

The c o s t of reducing s o l i d f u e l elements i n t o a form s u i t a b l e f o r processing (aqueous s o l u t i o n of uranyl and thorium n i t r a t e s ) is included i n t h e d a i l y operating cost (lob),

It seems l i k e l y t h a t t h e c o s t of convert-

ing t h e MSCR f u e l by t h e method proposed i n Sec. 5.2 would be less c o s t l y than t h e reduction of s o l i d f u e l elements by Darex o r Zircex,

However, an

investment of $500,000 f o r MSCR f u e l head-end treatment was allowed, based


I24 on estimates extracted from reference 20.

This r e s u l t e d i n a head-end

treatment charge of $0.53/kg Th processed. The c o s t of converting recovered thorium n i t r a t e t o ThF4 was not given i n t h e Guide.

However, t h e c o s t of converting low enrichment uranium t o

UF6 was only $5.60/kgO

Since t h e proposed thorium conversion process is

simpler and s h o r t e r than t h e uranium process, it seemed adequate and conservative t o assign a c o s t of $5.00/kg Th f o r t h e conversion of ThF4. Summing these charges, one has Processing Plant Cost,= (34.7

+

1167/r) $/kg

where r = kg Th/day removed from t h e reactor. Shipping c o s t s are presented i n Section 465 of t h e Guide (521, and include freight, handling, cask r e n t a l , and property insurance.

The c o s t

is $16/kg of uranium (or thorium) for spent f u e l elements. MSCR f u e l must be packaged before shipment. I n t h e scheme proposed, 3 t h e d a i l y productions of 1.67 ft3/day are accumulated i n 35 f t batches i n t h e spent f u e l f a c i l i t y (Sec, 408elO) which contains 5 tanks each having a 3 capacity of 35 f t and provided with s u f f i c i e n t cooling t o remove a f t e r h e a t 3 f r o m freshly i r r a d i a t e d fuel. A 10 f t batch is drained i n t o an INOR "bottle" (cylinder) one foot i n diameter and 1 3 feet long.

This is in-

s e r t e d i n t o an individual shipping cask and transported t o t h e Thorex plant.

If t h e b o t t l e s c o s t as much as $10,000 apiece, but are re-usable

f o r a t least 10 years, they w i l l add t o t h e shipping cost only $1.74/kg Th

a t t h e lowest rate of processing (40 kg/day), The charge f o r shipping decontaminated thorium back t o t h e plant is represented t o be $l.OO/kg i n t h e Guide. Since t h e processing plant operating charge was a l l levied against t h e thorium, t h e only charges on t h e uranium are shipping a t $17/kg per round t r i p and reconversion, for which operation t h e Handbook gives $32/kg for converting t h e n i t r a t e of highly enriched 235U t o UF6,

Although

t h e cost of conversion t o UF should be less, t h e given c o s t was assumed. 4 The t o t a l charge f o r t h e uranium was thus $49/kg, and was charged against

a l l isotopes of uranium, including 238U and t h e precursor, 233Pa.


125 5.4

Comparison of Processing Cost Estimates

For t h e purposes of comparison, t h e r e s u l t s of t h r e e other processing c o s t s t u d i e s are c i t e d .

The first, by Carter (20), d e a l t with a central

Fluoride V o l a t i l i t y plant designed e s p e c i a l l y f o r processing MSCR fuel a t

a rate of 31.5 ft3/day (1000 kg Th/day),

The estimates were based on t h e

same design study (21) used t o prepare t h e estimates reported i n Sec. 5.2. The s c a l i n g , however, was performed on t h e p l a n t as a whole, r a t h e r than with individual items, and t h e s c a l i n g f a c t o r was 0.31.

The estimated cap-

i t a l c o s t was $31.5 million, and t h e annual operating expense was $3.0 million. This study is r e f e r r e d t o below as t h e "CMS" study.

~

The second study, a l s o by Carter (201, d e a l t with a c e n t r a l Thorex

-

f a c i l i t y adapted t o process f l u o r i d e f u e l s (1000 kg Th/day) and was based on a d e t a i l e d analysis of t h e head-end and tail-end processes (described i n t h e previous s e c t i o n ) as w e l l as t h e Thorex separation process, and

w a s based on an unpublished design and c o s t study performed by Carter, Harrington, and Stockdale a t ORNL i n which flow sheets, equipment s p e c i f i cations, p l a n t arrangement drawings, building drawings, etc, were prepared and t h e items were costed.

as t h e "CHS" study.

For brevity, t h i s study i s r e f e r r e d t o below

It is, e s s e n t i a l l y , an independent estimate of t h e

f a c i l i t y assumed i n t h e Guide t o Nuclear Power Cost Evaluation (521, adapted f o r f l u o r i d e fuels.

The estimated c a p i t a l c o s t , scaled t o a capac-

i t y of 1000 kg Th/day, was $36.5 m i l l i o n , and t h e annual operating expense

was $3.6 million. The t h i r d study, performed by W. H e Farrow, Jr, (321, d e a l t with sev-

eral radiochemical separation p l a n t s f o r s e v e r a l d i f f e r e n t s o l i d f u e l s and clads and with both d i r e c t and remote maintenance.

The purex process, which

is similar t o Thorex, was employed f o r the separations,

The most applicable

case was t h a t of a remote

aintained plant capable of t r e a t i n g one tonne

per day of n a t u r a l urani

Although t h e process was described i n consid-

e r a b l e d e t a i l , a c o s t breakdown was not given by Farrow

The c a p i t a l c o s t

was $43 million, and t h e annual d i r e c t operating c o s t was$3.7 million.

One

might reasonably assume t h a t a Thorex plant of t h e same capacity would have

b, *

approximately t h e same c o s t s , and t h a t t h e head and tail-end treatments f o r f l u o r i d e f u e l s would be no more expensive, and perhaps less, than those of the solid fuels,


I26 It i s seen t h a t t h e estimates of t h e t o t a l c o s t , including shipping, vary by a f a c t o r of two, ranging from $37 t o $75/kg of thorium processed.

It seems plausible t h a t processing MSCR f u e l ' i n a .central Fluoride Vdlatili t y f a c i l i t y w i l l cost less than $50/kg, and possibly less than $QO, and t h a t processing i n a Thorex plant w i l l cost less than $70, and possibly

less than $50 per kg of thorium. Comparison of Estimates of Processing T a b l e 5.9. Costs f o r t h e MSCR Reference Designa This Work

CMS

AEC

CHS

Farrow c

Type of Plant

F1. Vol. F1. Vol. Thorex

Thorex

Location

Central Central

Central

Central Central

Capacity, kg Th/day

1000

1000

600

1000

(1000 )b

Batch Size, kg Th

6000

6000

6000

6000

6000

2

2

8

2

(2F

Capital Cost, $106

25.2

31.2

0

36.5

43

Annual Operating E maintenance Cost, $106 Process Plant Cost, d

2.0

3.0

r

3.6

3.7

26.6

34.5

56.5

40.5

46.4

Shipping Cost, $/kg Th

10.3f

10.3

18.7

10.3

(20P

Processing Cost, $/kg Th

36.9

44.8

7502e

50.8

(66.4

Turn-Around-Time

, Days

(Purex)

$/kg Th

IC

a Spent f u e l is withdrawn from t h e r e a c t o r (1.67 ft3/day; 53.3 kg Th/day)and shipped t o a c e n t r a l processing plant i n 10 f t 3 batches. These are accumulated i n 6000 kg Th batches, cooled for an average of 90 days, and processed i n s i x days (10 days i n AEC p l a n t ) . bPurex p l a n t capacity of one tonne of n a t u r a l uranium per day assumed equivalent t o one tonne thorium per day. C

These items w e r e not estimated by Farrow; values chosen were thought t o be consistent with h i s general treatment, dFixed charges were 14.46%, except f o r AEC plant where d a i l y opera t i n g charge (escalated t o 1962) was given i n reference 52. e Turn-around-time was 8 days. With a more r e a l i s t i c time of 2 days, and shipping c o s t s of $10.30/kg Th, t h e t o t a l c o s t s would be only about $ 5 0 , which is comparable t o t h e o t h e r estimates l i s t e d , except Farmw'S. \

fTable 5.7.

b-


127 6.

6.1

FUEL CYCLE ANALYSIS

Analvsis of Nuclear Svstem

The nuclear calculations for t h e MSCR were performed by means of t h e MERC-1 program f o r t h e IBM-7090,

This program, described by Kerlin e t al.,

i n Appendix D, uses t h e multigroup neutron diffusion code Modric and t h e isotope code ERC-10 as chain links. 6.1.1

Computer Programs

-

Modric. - T h i s

program was employed i n a 34-group version using t h e

group energies and cross sections of t h e various elements given i n refer-

ence 54. The cross sections were adopted f o r t h e most p a r t from Nestor's tabulation (721, with some minor modifications described below. MaxwellBoltzman averaged thermal cross sections and resonance i n t e g r a l s of import a n t materials are l i s t e d i n Table 6.1.

Thermal spectrum hardening was

ignored , Although t h e treatment of downscattering i n Modric provides f o r t h e t r a n s f e r of neutrons from any group i q t o any of t h e t e n next lower groups, t h e required s c a t t e r i n g matrices were not available when t h e MSCR calculat i o n s were begun.

Subsequently, t h e matrices were computed taking i n t o

account fast f i s s i o n s and i n e l a s t i c s c a t t e r i n g i n thorium and t h e fact t h a t t h e elastic s c a t t e r i n g lethargy decrements were i n some ranges l a r g e r than t h e group widths.

A s i n g l e calculation was made t o determine t h e impor-

tance of t r e a t i n g t h e downscattering i n t h i s more precise mannero The

effect was found t o be i n s i g n i f i c a n t ,

-

ERC-10, -The b a s i c isotope equation i n ERC-10 computes t h e concentra-

t i o n t h a t is i n e q u i l i b r i

i t h t h e sources (make-up, recycle, transmuta-

t i o n , decay, f i s s i o n ) and

s i n k s (transmutation, decay, waste, sales,

ly 233U o r 235U) may be s e l e c t e d t o satisfy a c r i t i c a l i t y equation; t h e feed r a t e or sales rate required to maintain t h e

recycle),

One isotope (u

c r i t i c a l concentration i s t h Three isotopes, 234U9 236U9 and 238U, approach equilibrium with periods long compared t o t h e assumed f u e l l i f e (30 years),

The e q u i l i b r i u n


128 Table 6 . 1 .

Cross Sections and Resonance I n t e g r a l s Used i n MSCR Multigroup Neutron Calculations (Values i n barns per atom)

Thermal cross Sectiona Material Fission (vu,) Absorption (aa) 232111

Resonance I n t e g r a l s b Fission (vuf) 8.286 x 10-1

Absorption (ua) 9.684 x 10

lo2

8.67221 x

233~a 23313

2.01972 x

2 34u

1,0602 x 10

2 3 5 ~

7,54945 x

236~

5.487 1.0983

37Np 238~

lo3

lo2

2 9,60119 x 10

lo2

6.89453 x

2 4.66792 x 10

lo2

2*87528 x X

10

3,225

5.70047 x 10

2

lo2

2.77241 x

6Li

4,58811 X lo2

'~i

1.5774 x

9Be

4.748 x

9F

1.975 x

loo2

1%

1.897 x '.01

INOR-8

6.004

13 5 ~ e

4.5756 x lo4

%axwell-Boltzman averaged a t 1200OF

.

bCut-off a t 0,437 ev. ; i n f i n i t e d i l u t i o n values. concentrations of these were not computed; instead, first approximations of t h e i r time-mean concentrations ( s t a r t i n g with a r e a c t o r i n i t i a l l y inven-

t o r i e d with 235U 95% enriched) were computed as described i n Appendix Ha The t r a n s i e n t behavior of all other isotopes, including 233Pa9 233U9 and f i s s i o n products, was ignored i n the optimization s t u d i e s . The appSoxmations involved i n t h i s approach are examined i n Appendix H,

MERC-1.

-Input

( 5 4 ) c o n s i s t s of s p e c i f i c a t i o n s of t h e geometry and

dimensions of t h e r e a c t o r regions, i n i t i a l guesses a t t h e composition, information on power, f u e l volume, processing modes and rates, composition of make-up materials, u n i t values, and processing costs.

The output c o n s i s t s

/-,

W


129 of nuclear d a t a (breeding r a t i o , mean eta, neutron balance, etc.), processing d a t a (feed and production rates), and a f u e l cycle cost (inventory, make-up processing, etc.).

An examination of these d a t a d i s c l o s e s t h e

p r i n c i p a l items of cost and suggests changes i n s p e c i f i c a t i o n s which might reduce these.

I n some cases, t h e effect of changing an input parameter is

not r e a d i l y predictable; e,g,, increasing t h e concentration of thorium i n t h e f u e l stream usually increases t h e conversion r a t i o and so reduces t h e 23%J

tory.

feed requirements but on t h e other hand increases t h e f i s s i l e inven-

I n cases such as t h i s , t h e input parameters are varied systemati-

c a l l y i n a "factored" set of calculations yielding t h e maximum information from a minimum number of cases,

The f u e l cycle cost is thus optimized

with respect t o s e v e r a l variables simultaneously. 6.1.2

(See Section 6,4.1.)

Reactor Physics Model The r e a c t o r was computed as a homogeneous mixture i n equivelant spher-

ical geometry ( i o e o , t h e input diameter was 1.09 times t h e diameter of t h e c y l i n d r i c a l core),

Thus t h e heterogeneity of t h e core, which is appreci-

able, was ignored,

But t h i s treatment is conservative i n t h a t t h e reso-

nance escape is underestimated, r e s u l t i n g i n a pessimistic estimate of t h e mean eta of t h e system,

I n regard t o t h e estimated captures i n thorium,

these can always be matched a t some neighboring concentration,

Aside from

a minor effect on t h e spectrum of neutrons, t h e chief e r r o r introduced is a s l i g h t underestimate of t h e thorium inventory. But t h i s is not important, f o r t h e inventory charges contribute only a small portion of t h e t o t a l f u e l c o s t s (less than 5%). The equivalent s p h e r i c a l core comprised t h r e e zones; an inner zone consisting of a homogeneous mixture of f u e l s a l t and graphite, a s p h e r i c a l annulus about one inch t h i c k f i l l e d with f u e l s a l t , and a s p h e r i c a l r e a c t o r v e s s e l one inch t h i c k ,

The

assumed t o be 1200째F, and

e f f e c t i v e temperature of t h e f u e l was e r a t u r e was a l s o assigned t o t h e graph-

o r t h r e e hundred degrees warmer. 6.1.3

Cross Section Data Thorium-232,

- Saturation of resonances

(self-shielding) was found t o

be important i n f i v e of t h e neutron energy groups,

c

The e f f e c t i v e cross


130 sections of thorium i n these groups were calculated by means of a correlat i o n developed by J. H. Miller (Appendix B) and based on t h e t h e o r e t i c a l analysis of e f f e c t i v e resonance i n t e g r a l s made by L a Dresner (29). 2200 m l s cross s e c t i o n of 39 barns, as recom-

Protactinium-233. - A

~

mended by Eastwood and Werner (301, was assigned.

This was d i s t r i b u t e d as shown i n Appendix A.

900 barns was adopted, Uranium-233. - A

A resonance i n t e g r a l of

I

value of 2.29 was adopted f o r eta a t thermal energies,

based on t h e recent measurements of Gwin and Magnuson (43),

For energies

above thermal, Nestor's estimates (72) were used, as tabulated i n Appen-. d i x A.

Resonance s a t u r a t i o n effects were ignored.

Beryllium. -The nored.

(n,2n) r e a c t i o n of energetic neutrons i n 9Be was ig-

It is of small importance i n t h i s graphite moderated r e a c t o r , and

is moreover o f f s e t by t h e disadvantageous (n,a) r e a c t i o n which uses up a neutron and leads t o t h e formation of 6 L i , Fission Products. -These,

excepting xenon and samarium, gere handled

c o l l e c t i v e l y i n t h e Modric c a l c u l a t i o n and individually i n t h e ERC calculation

, as described

below.

An "effective" concentration of an "aggregate"

f i s s i o n product was computed f r o m ERC r e s u l t s and used i n t h e multigroup calculation i n conjunction with an a r b i t r a r y set of group c r o s s s e c t i o n s composed of a hypothetical standard absorber having a thermal c r o s s s e c t i o n

of one barn and an epithermal cross s e c t i o n corresponding t o a l / v variat i o n above thermal. Thermal cross s e c t i o n s and resonance i n t e g r a l s f o r f i s s i o n products

were mostly taken from t h e compilation- of Garrison and Roos (371, supplemented by estimates from Bloemeke (91, Nephew (711, and Pattenden (90). The d a t a used, including f i s s i o n y i e l d s and decay constants, are tabulated

i n Appendix E. 6.2"Analysis

of Thermal and Mechanical System

The analysis included consideration of f l u i d flow i n and mechanical arrangement of t h e r e a c t o r and equipment associated with t h e e x t r a c t i o n of heat f r o m t h e f u e l stream, as they affect f u e l cycle costs.

,-\

&J

311


131 Maximum Fuel Temperature

6.2.1

It is believed t h a t t h e rate of corrosion of INOR w i l l be very small

and t h a t t h e heat exchangers w i l l have a very long lifetime i f t h e tempera t u r e of t h e f u e l s o l u t i o n does not exceed 130OOF (121, The maximum allowable temperature may be higher; i f so, f u t u r e generations of molten-

salt r e a c t o r s w i l l be a b l e t o achieve higher s p e c i f i c powers and higher thermal e f f i c i e n c i e s

Minimum Fuel Temperature

6,2.2

For t h i s , llOO째F was selected.

Earlier work ( 3 ) had shown t h i s t o be

a reasonable value, and calculations summarized i n Table 6,2 confirmed i t s optimal q u a l i t y f o r t h e MSCR,

Perturbations of the affected c a p i t a l c o s t s

were calculated i n Table 6.2, and it was concluded t h a t the optimum tempera t u r e i s very near llOO째F (see a l s o Sec. 3.1.12). 6.2 3

Velocity

Erosion does not appear t o be a problem i n s a l t - I N O R systems, nor does t h e r e appear t o be any dependence of t h e corrosion rate on velocity. Rather, t h e v e l o c i t y appears t o be limited by considerations of pumping power and stresses induced by pressure gradients.

Pressure drop across t h e

heat exchangers is limited by t h e rapid increase i n c o s t of INOR s h e l l s with increasing wall thickness.

Likewise, maximum allowable pressure drop

across t h e r e a c t o r core is determined by l i m i t a t i o n s on s t r e n g t h and t h i c k -

ness of t h e r e a c t o r v e s s e l and i n t e r n a l support members. Velocities, pressures, and wall thicknesses a t various points i n t h e f u e l circuit of t h e rence design are l i s t e d I n Table 6,3. The pumping power requ e r heat exchanger loop was calculated t o margin of 500 horsepower for pump and be about 1500 horsepower; w i en losses. This pumping power requirement results i n a pumping c o s t (with electric power a t 4 mills/kwhr) of

motor i n e f f i c i e n c i e s and un 0.004 m i l l s / b h r e .


Table 6.2.

MSCR Minimum Fuel Temperature Optimizationfc

Case 1

Case 3

Case 2

Minimum f u e l temp., OF

1050

1100

1150

Fuel stream temp. r i s e , OF

250

200

150

Fuel s a l t flow rate, l b / h r x

89

111

148

Heat exchanger log-mean temp,

144

174

200

53,000

46,000 (-$306

62 101

80 84

123 73

696 (+$3,025) 584 (+$125)

575 488

504 (-$1,775) 424 (-$83)

133 255 (-$1,625 15,400 (-$31)

150 320 18,500

195 410 (+$2,250) 21,300 (+$28)

9,100 (-$736) 17,300 (+$256)

12,800 16,000

22,200 ( 6 1 , 8 7 2 ) 15,200 (-$l60)

$+1.49

Reference Condition

$+1,83

difference,

OF

2

Heat exchanger area, f t Heat exchanger pressure drop, p s i Fuel s i d e Coolant s i d e

Heat exchanger volume, f t 3

Fuel s i d e Coolant s i d e

P

w

h,

Fuel c i r c u i t piping Pressure drop, gsi Pipe volume, f t Pipe weight, Ibs. Pumps Fuel c i r c u i t , hp Coolant c i r c u i t , hp

Net c a p i t a l c o s t increment $million

*Capital c o s t increments with r e s p e c t t o reference condition (1100OF) associated with each item are given i n p a r e n t h e s i s i n $thousands. In t h i s c a l c u l a t i o n , r e a c t o r o u t l e t temperature, configuration, h e a t exchanger c r o s s s e c t i o n , f u e l v e l o c i t y i n piping were held constant, as was flow r a t e of coolant s a l t and i t s temperature extremes. The power of t h e r e a c t o r was constant, and t h e thermal e f f i c i e n c y was not a f f e c t e d ,

gi

C)


2

133 -%â&#x20AC;?

c4

T a b l e 6.3.

C h a r a c t e r i s t i c s of Fuel C i r c u i t of 1000-Mwe Molten S a l t Converter Reference Design Reactor

Locat ion

Velocity (ft/sec 1

Pump discharge Top of heat exchanger

190 185

35

-

Heat exchanger tubing

Pressure (psia)

8A

Bottom of heat exchanger Bottom of reactor

35

-

Fuel channels i n core

4.3

Top of r e a c t o r (Pump suction)

-

-

0.035

95

0 250

22.5

20

0e 406

80

7,4

Minimum Wall Thickness (inches )

-

-

0 312

*Shell-side

6.2.4

Fuel Volume Contributions t o t h e volume of t h e f u e l system are l i s t e d i n Table 4.1

f o r the reference design.

The volume of f u e l i n t h e e x t e r n a l system de-

pends on t h e power l e v e l and various l i m i t a t i o n s such as those on s a l t temperature, pressure, velocity, thermal stress i n and minimum thickness of heat exchanger tubing, etc. The reference system was designed with considerable conservatism. The f u e l volume could be reduced appreciably w i t h an increase i n s p e c i f i c power; pumping power c o s t s and c a p i t a l investment i n pumps would increase.

Cost of heat exchangers might also increase.

The

design of t h e system should be optimized with respect t o t h e sum of a l l t h e c o s t s affected; but t h i s l a y o

i d e t h e scope of t h e present study.

ernal portion of t h e f u e l system of t h e 3 This is very much smaller reference design is approximately 2.8 M w t / f t 3 than t h e 7.6 M w t / f t used i n a - p r i o r study of a molten s a l t breeder (3) The power density i n t h e

.

which was based on a study by Spiewak and Parsly (991, who estimated a s p e c i f i c power of 4,9 M w t / f t 3 f o r a first generation p l a n t and 7.6 M w t / f t 3

u

for subsequent p l a n t s ,


134 Since only about one fourth of t h e t o t a l is contributed by t h e f u e l i n t h e a c t i v e core, t h e t o t a l f u e l volume is not very s e n s i t i v e t o changes i n f u e l volume f r a c t i o n i n t h e core.

The t o t a l volume is r a t h e r more sen-

s i t i v e t o changes i n core diameter.

The volume of neutron-active f u e l in-

creases as t h e cube of t h e core diameter (height equal t o diameter), while volume i n t h e r a d i a l annulus and i n t o p and bottom plenums increases as t h e square. 2000

About 40% of t h e t o t a l volume i s a f f e c t e d ,

I n a f u l l y optimized system, t h e f u e l volume might plausibly be f t 3 , and perhaps be as low as 1800 ft', provided some of t h e holdup i n

end-plenums, etc. can be eliminated.

6.3

-

Analysis of Chemical System

The composition of t h e f u e l stream as a function of its chemical in-

t e r a c t i o n with t h e r e a c t o r environment and with t h e processing p l a n t is considered.

Behavior of xenon-135 is ImporYtant and is d i s c u s s e d , i n d e t a i l

i n Section 6.8.

Stagnation of f u e l i n crevices between moderator elements

may be important; however, it does not seem possible a t t h i s time t o evaluate t h e effect except t o say t h a t p a r a s i t i c captures of neutrons i n f i s s i o n products immobilized i n such places w i l l take place.

This uncer-

t a i n t y was lumped with t h a t associated with corrosion products as discussed below

.

So far as the composition of t h e f u e l

stream is concerned, t h e chem-

ical effects of t h e two proposed processing methods (Thorex vs f l u o r i d e

The recycled uranium is radioactive and must be handled a t least semi-remotely whether t h e thorium is recycled or not. v o l a t i l i t y ) are t h e same.

Both methods r e s u l t i n recycle feed containing only n e g l i g i b l e amounts of nuclear poisons (other than isotopes of uranium) and both r e t u r n 233Pa removed from t h e reactor as 233U with losses t h a t depend only on t h e holdi n g time p r i o r t o chemical treatment, 6.3.1

Thorium-232

.

Thorium may be recycled i f Thorex processing is used; however, according t o t h e Guide, t h e capacity of a multi-purpose processing p l a n t would be

h

Wc


135 -=(J

reduced by 40% (52).

(One could design t h e plant t o handle t h e thorium

without loss of capacity, however, and recover t h e thorium a t no extra cost.)

It t u r n s out t h a t t h i s reduction i n capacity almost exactly o f f s e t s

t h e value of t h e thorium saved,

With t h e f l u o r i d e v o l a t i l i t y processing,

thorium is not recovered except by means of an a d d i t i o n a l s t e p presently not a v a i l a b l e 6.3.2

.

Protactinium-233 With mean residence times of t h e order of 1000 days, protactinium f o r

t h e most p a r t decays while still i n t h e c i r c u l a t i n g f u e l system.

However,

t h e process stream carries 60-80 grams per day out of t h e r e a c t o r system. I n t h e proposed reference design processing scheme (Section 5.21, t h e processing stream i s accumulated f o r 120 days and then held as a batch f o r an a d d i t i o n a l 30 days, giving a t o t a l hold-up time of 150 days and an average decay t i m e of 90 days.

A t t h e time of processing, 85% of t h e 233Pa i n t h e

mixed batch w i l l have decayed t o 233U, so t h a t t h e maximum loss of 233Pa

w i l l be only 9 t o 1 2 grams per day. 6.3.3

Uranium-233 The loss per pass through t h e processing p l a n t was assumed t o be 0.3%. I n t h e reference design r e a c t o r , t h e product stream from t h e process

plant was recycled t o t h e r e a c t o r ; however, t h e economics of sale of t h e product stream w a s also examined. 6.3.4

Uranium-234

i g h t per cent 238U, Oo72 per cent 235U, and Natural uranium is 99.27 0.0055 per cent 234U. If t h e r e were no enrichment in-234U r e l a t i v e t o

a diffusion p l a n t product containing 95 weight per cent 235U would a l s o contain 0,726 per cent 234U, with t h e b a l e being 238U. In o r 235U,

allow f o r some enrichment, t h e composition of t h e feed was taken t o be 95 per cent 235U, 1 per cent 234U, and 4 per cent 238U, A s with 233U, t h e processing losses were 0.3 per cent/passo I n i t i a l l y t h e r e a c t o r contains l i t t l e 234U.

rium

The approach t o equilib-

is slow, and, as seen i n Appendix H, t h e average concentration over


136

a period of t h i r t y years is only 65 per cent of t h e equilibrium concentrat i o n . The average value, r a t h e r than t h e equilibrium value, was therefore used i n t h e nuclear calculations. 6.3,s

Uranium-235 An enrichment of 95% was s e l e c t e d f o r t h e make-up,

Losses i n process-

ing were assumed t o be 0.3% p e r pass. 6.3.6

Uranium-236

~

The concentration of t h i s isotope a l s o approaches equilibrium slowly

with respect t o a f u e l lifetime of 30 years, t h e 30-year l i f e was used. 6.3.7

A concentration averaged over

Losses i n processing were assumed t o be 0.3%.

Neptunium-237 This was assumed t o be removed completely i n t h e processing plant.

6.3.8-

Uranium-238

For reasons given above, t h e fuel make-up stream was assumed t o cont a i n 4% 238U by weight. Losses i n t h e process plant were assumed t o be 0.38. The 30-year average concentration was used i n t h e nuclear calculat i o n s instead of t h e equilibrium concentration,

.

6 3.9

Neptunim'-239 and Plutonium Isotopes Only small amounts of these w i l l be formed, and they are l o s t i n t h e

waste.

Accordingly, t h e i r formation i n t h e f u e l stream was ignored and no

breeding c r e d i t was taken for absorptions i n 238U0 6.3.10

Salt

The fuel carrier -.i n t h e reference design consisted of 68 mole per cent LiF, 23 per cent BeF2, -and 9 per cent ThF,,. Lithium i n t h e make-up s a l t Captures i n 'Li, Be, and F were lumped under an equiva__ The mean r e a c t o r concentration of 6 L i and l e n t isotope "Carrier-1."

was 99,995% 7Li.

neutron captures t h e r e i n were computed separately.

Lithium and beryllium

i n t h e process stream are l o s t t o t h e waste, and no value was assigned t o


137 t h e waste.

The make-up rate was made equal t o t h e discard rate f o r 7 L i and

Be; t h e 6 L i feed rate was proportioned t o t h e 7 L i make-up rate.

6,3,11

Xenon-135 and Related Isotopes

I n t h e reference design r e a c t o r , it was assumed t h e graphite has a diffusion c o e f f i c i e n t no g r e a t e r than loF6 cm2/sec ( D = 10" cm2 /sec f o r MSRE graphite), a porosity no greater than 0.01,

and t h a t 1 0 per cent of

t h e f u e l stream is r e c i r c u l a t e d t o t h e dome of t h e expansion chamber and

Here Xe is desorbed and swept away i n a stream of helium gas with an e f f i c i e n c y of 100 per cent per pass, With these assumptions, t h e loss i n breeding r a t i o due t o absorptions i n 135Xe is 0.017 as shown t h e pump bowls.

i n Appendix G, where t h e l o s s e s corresponding t o o t h e r assumptions are a l s o given e

Noble Metal'Fission'Products

6.3.12

It was assumed t h a t t h i s group of isotopes, comprising Mo, Rh, Ru, Pd,

and In, "plate out" on INOR surfaces i n t h e f u e l c i r c u i t with an e f f i c i e n c y of 1.0 per cent per pass,

6.3.13

Other Fission Products

These are removed 100 per cent i n t h e Fluoride V o l a t i l i t y process, and only n e g l i g i b l e amounts w i l l be present i n t h e recycle stream from a Thorex plant e ;

-

6,3,14

1I

Corrosion Products

Data are meager from which t h e concentration of corrosion products i n t h e c i r c u l a t i n g f u e l stream could be estimated, I n an in-pile loop operated for 15,000 hours, t h e concentrations of iron, nickel, and chromium appeared t o f l u c t u a t e about equilibrium values (84, p a 791, On the b a s i s of t h e s e d a t a one might expect t h e f u e l t o contain 50 ppm of nickel, 500 ppm of chromik, and about 250 ppm of iron, I n t h e reference design r e a c t o r , , these concentrations would r e s u l t i n a poison f r a c t i o n (loss i n breeding r a t i o ) of 0,006 u n i t s .

A poison f r a c t i o n of 0.008 u n i t s w a s arbi-

t r a r i l y assigned t o corrosion products i n t h e calculations, making some

L


138 allowance f o r f i s s i o n products immobilized i n Cracks and Crevices i n t h e moderator, etc.

6,C"Fuel'Cycle

Optimization

The designer has l i t t l e c o n t r o l over some of t h e independent v a r i a b l e s

For instance, t h e maximum allowable f u e l temperature is f i x e d by necessity of l i m i t i n g corrosion rates. For some t h a t affect t h e f u e l cycle cost.

variables t h e f u e l c o s t may decrease monotonically as t h e v a r i a b l e tends toward an extreme value, but o t h e r c o s t s may increase,

For example, de-

of f u e l s a l t decreases inventory charges but creasipg . . t h e e x t e r n a l volume _.

increases pumping c o s t s o A p l a u s i b l e and conservative e x t e r n a l volume was s e l e c t e d for t h e reference design optimization; however, effect on f u e l cycle c o s t of decreasing t h e e x t e r n a l volume is e a s i l y estimated.

After t h e values of such fixed or l i m i t i n g v a r i a b l e s were established, t h e r e remained s e v e r a l which required optimization simultaneously with respect t o t h e fuel costs. variables

.

These variables were designated t h e "key"

The key independent v a r i a b l e s were found t o be t h e diameter (D) of t h e

core, t h e volume f r a c t i o n (F) of f u e l i n t h e core, t h e concentration (t4) of thorium i n t h e f u e l s a l t , and t h e processing rate (R).

The second and

t h i r d combine t o f i x an important subsidiary variable, t h e C/Th atom r a t i o (an i n d i c a t i o n of t h e degree of moderation of t h e system).

Fixing a l l four

and then s a t i s f y i n g t h e c r i t i c a l i t y equation together with t h e equilibrium isotope equation results i n f i x i n g t h e Th/U ratio, breeding r a t i o , f u e l c o s t , etc. Exploratory c a l c u l a t i o n s showed t h a t t h e f u e l cycle c o s t is a r a t h e r s e n s i t i v e function of t h e C/Th ratio, but is i n s e n s i t i v e t o t h e diameter of t h e core i n t h e range from 15 t o 20 feet.

Relative breeding ratios and

f u e l cycle c o s t s for a series of c a l c u l a t i o n s are shown i n Table 6.4. On t h e b a s i s of these results, t h e 20-foot core having a f u e l volume

f r a c t i o n of 10 per cent and using a f u e l s a l t containing 9 mole per cent ThF4 was s e l e c t e d for f u r t h e r study.

The optimum processing rate f o r t h i s


139 Conversion Ratios and Fuel Cycle Costs 1000 Mwe T a b l e 6.4. Molten S a l t Reference Design Reactor Processing a t Rate o f Two Cubic Feet p e r Day

Case

Core D i a m f t o

Vol. Frac. Fuel

Mole % Th F4

C/Th

Th/U

Conv. Ratio

FCC m/kwhr

1

15

0,18

13

107

1066

0091

1.09

2

17,7

0.18

13

107

12.4

0.96

1.02

3

20

0.18

13

107

13.5

0.99

1.04

4

15

0.10

9

293

21.0

0.84

0.73

5

17.7

0,lO

9

293

23,l

0,87

0.69

6*

20

0010

9

293

2400

0.90

0.70

7

15

0,107

5

468

20,O

0.68

0.82

8

17.7

0.107

5

468

21.7

0,72

0.78

9

20

0 107

5

468

22.9

0.74

0.77

.

*Reference case, preferred over Case 5 because o f lower power d e n s i t y , lower v e l o c i t i e s , etc,

combination o f key v a r i a b l e s was then determined from r e s u l t s l i s t e d i n

Table 6.5. The numbers given i n Table 6.5 are p l o t t e d i n Fig, 6.1.

The f u e l c o s t

has a minimum somewhat below 0.7 mills/kwhre a t conversion r a t i o s l y i n g between 0 . 8 5 and 0.9. i

-

S l i g h t changes i n c o s t assumptions, etc. could s h i f t

t h e l o c a t i o n o f t h i s minimum over a wide range, To t h e l e f t of t h e cost minimum, loss of neutrons t o f i s s i o n products i n c r e a s e s burnup c o s t s more r a p i d l y than processing c o s t s decline; t o t h e r i g h t t h e processing l o s s e s outweigh t h e gain i n conversion.

The extreme

i n t h e conversion r a t i o r e s u l t s from t h e fact t h a t processing l o s s e s increase l i n e a r l y with processing rate whereas loss o f neutrons t o f i s s i o n products decreases only i n v e r s e l y , The processing c o s t s used i n computing T a b l e 6.5 were t h o s e a s s o c i a t e d with a central Fluoride V o l a t i l i t y Plant (Sec, 5,2) but followed, where

-

a p p l i c a b l e , t h e p r e s c r i p t i o n given i n t h e Guide (52).

time was 2 days f o r a l l of t h e cases l i s t e d ,

The "turn-around"


140

1.1

1.0

0.9

0.6

0 Central Fluoride Volatility Processing Reference Design Case 0-5

0.4 0

5

I 10

CONVERSION RATIO

Fig. 6.1. MSCR Fuel Cycle Cost Versus Conversion Ratio.


141 Table 6.5.

Processing Variables

Case

Cycle Time days

Volume Rate f t /day

Weight Rate kg Th/day

Conv. Ratio

Fuel Cycle Cost mills/kwhre

1

3000

0.83

27

0 835

0.71

2

2000

1.25

40

0 868

0.68

3*

1500

1.67

53

0.895

0.68

4

1250

2.00

64

0 e 904

5

1000

2.50

80

.

0.70

0 917

0.74

6

500

5.0

160

0.930

1.01

7

250

10.0

320

0.908

1.64

.

5CReference Design Case

6.5

Reference Design Reactor

The MSCR is capable of producing power a t a f u e l cycle c o s t , including

s a l t charges, of 0.7 mills/kwhre a t a conversion r a t i o of 0.9, The most important u n c e r t a i n t i e s i n t h i s calculation arise i n connect i o n with ( a ) t h e behavior of xenon i n t h e core, (b) c o s t s estimated f o r t h e Fluoride V o l a t i l i t y processing p l a n t , ( c ) v a l i d i t y of t h e base charges assigned t o t h e materials and ( d ) c o s t of packaging t h e spent f u e l f o r shipment. The influence of xenon behavior is examined i n Sec. 606, and t h e c o s t assumptions i n Sec. 6.7,

If t h e l o s s e s t o xenon assumed f o r t h e

reference design case are a t t a i n a b l e (and it should be possible t o achieve t h e assumed performance by improving t h e graphite and t h e sparging process), then a l l t h e u n c e r t a i n t y \ r e s i d e s i n t h e processing cost.

Since processing

c o s t s are only about 0.08 mills/kwhre (Table 6,101, even a l a r g e e r r o r would not s i g n i f i c a n t l y influence t h e t o t a l fuel cycle cost. Uncertainties i n regard t o t e c h n i c a l f e a s i b i l i t y arise i n connection with compatibility and s t a b i l i t y of t h e graphite moderator (Sec, 4,2.2), t h e r e a c t o r vessel and i t s i n t e r n a l s t r u c t u r e (Sec. 4.2.2Ie

Also, t h e

hazards associated with t h e proposed methods of c o n t r o l (Sec. 4 , 2 , 3 ) have not been evaluated.

L

of


142 6.5.1 Specifications

Case 3 was selected for the reference design before the complete curve was generated; its fuel cost is not significantly greater than the minimum. The reactor characteristics and operating data for the reference design are given in Section 4, nuclear data in Table 6.6. Table 6.6. Nuclear Characteristics of 1000 Mwe Molten Salt Converter Reactor Case No. Carbon/thorium atom ratio

-300

Thorium/fissile uranium atom ratio

-27

Fraction of fissions in thermal neutron group Fraction of fissions in 235u Fraction of fissions in 2 3 3 ~ Ratio of total fissions to fissions in 233U and 235U Mean eta of 233U Mean eta of 235U Mean eta of all fissions*

3

â&#x20AC;&#x2DC; 0.82

,0.15 0.85

1.0018 2.253 1 979 2.219

Effective resonance integral of 232Th mermal cross section of 233~a

66 barns

Effective resonance integral of 233Pa Average power density in core

900 barns

Ratio peak-to-radial average power density Maximum graphite exposure rate, nvt/yr

2.1

Neutron energy X.1 MeV Neutron energy >1.0 MeV Average thermal flk Average fast flux Fissions per 235U atom added

39 barns 14.1 kw/liter

c

5.0

X

1021

1.8 X 1021 3.7 x 1013 4.2 x 1013 9.2

Fissions per fissile atom processed

1.5

Exposure, Mwd/tonne of thorium Specific power, Mwt/kg fissile

47,000 0.9

~~

*Neutrons produced from all sources per absorption in 233u and 235U. P


143 6.5.2

Neutron Ecdnomy From Table 6.7,

it is seen t h a t f i s s i o n s i n 232Th contribute very

l i t t l e t o neutron production.

The fast f i s s i o n cross s e c t i o n s of thorium

are appreciably less than those of 238U3 moreover, t h e fast f l u x i n t h e

MSCR is not p a r t i c u l a r l y high (Table 6.6).

Also, t h e thorium is r a t h e r

d i l u t e compared t o concentrations customarily proposed f o r blankets, The nuclear loss r e s u l t i n g from absorptions i n Pa is appreciable but not serious.

It could be reduced by reducing f u r t h e r t h e volume f r a c t i o n

of f u e l i n t h e core, but t h i s is already about as low (0,l) as seems techn i c a l l y f e a s i b l e ; a f u r t h e r decrease would probably add more t o t h e f u e l cost i n terms of increased inventory charges (since concentration of tho-

rium and uranium i n t h e f u e l stream would increase, while t h e e x t e r n a l volume would remain t h e same) than would be saved i n terms of f u e l replacement costs.

Increasing t h e e x t e r n a l volume i n order t o d i l u t e t h e Pa would

not be economical f o r t h e same reason. The nuclear loss t o Pa could be reduced by removing t h e Pa rapidly from t h e c i r c u l a t i n g stream and holding it u n t i l it decays t o 233U. I n order t o be e f f e c t i v e , such a process would have t o t r e a t t h e e n t i r e f u e l stream for Pa removal i n a period not g r e a t e r than t e n days (mean l i â&#x201A;Ź e of Pa is about 40 days).

Thus an extremely simple and e f f i c i e n t process is

required, as f o r example, t h e passage of t h e f u e l stream through beds where Pa is s e l e c t i v e l y absorbed and retained u n t i l 233U i s formed, which then desorbs i n t h e presence of l a r g e amounts of uranium i n solution. N o such

process is presently known, although s q e work has been reported (19 p. 117,

74) on t h e p r e c i p i t a t i o n of protactinium oxide from molten salt s o l u t i o n s containing up t o 2000 ppm of ur ium by contacting t h e melt with BeO, Thop,

or U O i " Development of such a process might reduce t h e l o s s e s t o Pa appreciably and add W.01 u n i t s t o t h e conversion r a t i o , \

For reasons indicated i n Sec, 6.1, t h e concentration of 234U used i n t h e equilibrium calculation was av

aged over a period of 30 years,

Although t h e reactor was assumed t o be i n i t i a l l y fueled with enriched uranium containing 1.0% 234U and t o be supplied with make-up f u e l of t h e same :composition, ninety-nine per cent of t h e 234U i s formed by transmutation of

233U.

It disappears from t h e r e a c t o r by transmutation t o 235U and l o s s i n


. 144 Table 6.7.

.

Neutron Economya i n t h e 1000 Mwe Molten S a l t Converter Reference Design Reactor

Item

Captures

Fissions

Absorptions

.

. .

2321111

0,8535

0 0011

0 8546

2 33 ~ a 2 3 3uC

0.0084

0.0000

0 0084

0.0888

0 7488

0.8376

2340

0.0572

0.0002

0 0574

23 5u

0 0301

0 1323

0 1624

236~

0 0184

0 0001

0 OA85

.

. . .

237Np

0 0074

2 3811

0 0029

Carrier s a l t

0.0387

Graphite

0 0564

1 35 ~ e

0 0170’

Other f i s s i o n products

0 0867

Corrosion products Delayed neutrons

0.0082 0.0046

Leakage

0 0513

.

.

.

. . -.

0 0000

--

--

--

-_-

_--

. .

. . .

0 0074

0.0029 0 0387 0.0564 0 O m c

0.0867

.

0 0082

. .

0 0046

0 0513

Neutron y i e l d a

2 2121

Processing l o s s e sb

0.005

Net conversion r a t i od

0.90

0

a

Neutrons p e r neutron absorbed ‘in 233U and 235U0

b

Processing loss o f O03%/pass f o r 233U and 235U, undecayed 233Pa.

and

C

Loss corresponding t o graphi e having gas p o r o s i t y of 1% and d i f f u s i o n c o e f f i c i e n t of IOo‘ compared t o current graphi t e p r o p e r t i e s of 10% and lo”. dExcluding captures i n 238U and c o r r e c t i n g for f i s s i o n s

of thorium.


I

145 t h e processing cycle (0.3% per pass).

Since 23% is i n f e r i o r with respect

t o 233U as a nuclear f u e l , it is advantageous t o keep t h e concentration of â&#x20AC;&#x2122;

However, t h e designer has l i t t l e c o n t r o l over

234U as low as possible.

t h i s , inasmuch as t h e only menas of removing it is t o s e l l spent f u e l t o t h e AEC, and t h i s , as shown i n Sec. 6.8,3,

i s not advantageous.

The concentration of 2 3 6 U used was a l s o averaged over a period of In t h i s t h i r t y years, s t a r t i n g with a clean r e a c t o r containing no 236U.

case, t h e only source is capture of neutrons i n 235U; t h e s i n k s are transmutation t o 237U (which decays promptly t o 237Np) and l o s s e s (0.3% per

pass) i n t h e processing over 236U is by varying 235U fed t o t h e system, It i s assumed t h a t processing cycle,

cycle.

The only e f f e c t i v e c o n t r o l t h e designer has

t h e conversion r a t i o thus varying t h e amount of and by sale of spent f u e l . 237Np was removed 100 per cent per pass through t h e

Parasitic captures are appreciable (0.0074 u n i t s on t h e

conversion r a t i o ) ; however, s p e c i a l processing f o r t h i s reason does not appear t o be worthwhile, although it might be for other reasons.

Parasitic captures i n carrier s a l t r e s u l t e d i n a conversion loss of 0,039 units. Of t h i s , captures i n 6 L i contributed 0,014 u n i t s . The grade

s a l t used (99.995% 7 L i ) appears t o be about t h e b e s t a v a i l a b l e at attractive prices. On t h e o t h e r hand, use of i n f e r i o r grades would not Of

result i n lower f u e l costs. The b e s t way t o c o n t r o l l o s s e s t o 6 L i is t o recycle carrier s a l t from processing instead of discarding it as w a s assumed i n t h e reference design. The p o s s i b i l i t i e s are examined i n Sec. 6.8, where it is shown t h a t about 0.01 u n i t s might plausibly be saved on t h e conversion r a t i o , and about

0,08 m i l l s i n replacement

The f u e l cost was opt

t h respect t o p a r a s i t i c captures i n

moderator and neutron leakage simultaneously as described i n Sec. 6a4.

Losses t o graphite might be decreased by decreasing the C/Th r a t i o , but t h e gain would be more than offset by l o s s e s i n eta of 233U and,increased leakage.

Leakage decreases slowly with increasing diameter, but f u e l cost

is i n s e n s i t i v e , as shown i n Table 6.4.

Efforts t o reduce t h e leakage by

use of a graphite reflector were not successful (Appendix M ) , l a r g e l y because of t h e necessary presence of a f u e l annulus a t least one inch t h i c k


146 a t t h e periphery of t h e core as a r e s u l t of tolerance allowance for diff e r e n t i a l thermal expansion. The estimate of t h e l o s s of neutrons t o xenon (0.017 u n i t s on C.R.)

cm2/seco, a porosity was based on an assumed d i f f u s i o n c o e f f i c i e n t of accessible t o gas of 1 , O p e r cent, and a sparging rate of 10 per cent (16-ft3/sec) of the c i r c u l a t i n g f u e l stream with 100 per cent removal of The prospects are good f o r t h e development of , grades of graphite t h a t would reduce t h e l o s s e s t o xenon i n t h e WSCR to no

xenon p e r pass (Sec. 6.8,2),

more than 0,005 units on t h e conversion r a t i o . Captures i n samarium (149Sm and 151Sm) r e s u l t i n a loss of 0,013 u n i t s ~

i n conversion r a t i o .

This loss is independent of t h e processing rate ex-

cept at very s h o r t cycle times of t h e order of days.

Thus, t h e r e is not

much prospect of reducing t h i s l o s s except by t h e application of some simple, rapid process similar t o t h a t suggested for 233Pa above.

Captures i n o t h e r f i s s i o n products can be controlled by varying t h e processing rate.

The savings from g r e a t e r f u e l conversion must be balanced

against increased processing cost.

For t h e p a r t i c u l a r p r i c e s t r u c t u r e and

nuclear properties assumed, t h e optimum r a t e of processing is 9 . 7 ft3/day (53 kg of Th/day).

The corresponding l o s s i n conversion r a t i o due t o

captures i n o t h e r fission products i s 0.074 units.

In t h i s calculation,

it was assumed t h a t xenon i s sparged as described above, noble metals are reduced by chromium i n metal structures t o t h e zero valence state and "plated out," and a l l o t h e r f i s s i o n products are removed by passage thraugh t h e processing cycle which, a t t h e c o s t s estimated i n Sec. 5.1, optimized

a t a cycle time of %1500 days. If t h e processing c o s t schedule (Sec, 3,2 and 5.1) were t o change i n

t h e d i r e c t i o n of lower costs, t h e optimum processing rate would increase, and t h e f i s s i o n product captures could be decreased.

Although improve-

ments and economies i n t h e Fluoride V o l a t i l i t y process are t o be expected, t h e remote operations and maintenance c o s t s of t h i s process are l i k e l y t o

remain high; therefore t h e process eventually used should be as simple as possible. The s o l u t i o n may l i e i n t h e d i r e c t i o n of d i s t i l l a t i o n or fract i o n a l c r y s t a l l i z a t i o n (80, p, 80), or perhaps extraction with l i q u i d

metals

.


I

f

147

The estimated loss i n conversion r a t i o t o corrosion products was 0.006 u n i t s ; a l o s s of 0.008 u n i t s was allowed i n t h e calculations (Sec. 6.3). Since t h e concentration of corrosion products (Fe, N i , and C r ) i n t h e melt appears t o reach an equilibrium i n times t h a t are s h o r t compared t o t h e processing cycle time, it seems unlikely t h a t t h e processing w i l l have much influence on t h e concentrations of corrosion products, and t h e associated

loss of neutrons seems unavoidable, 6.5.3

Inventories and Processing Rates

It may be i n f e r r e d from Table 6.8, Column 1, t h a t t h e s p e c i f i c power of t h e MSCR is 0,35 Mwe/kg f i s s i l e , which i s comparable with many of t h e advanced systems currently being put forward,

The exposure (about 40,000

Mwdt/tonne of thorium) is of the same order as t h a t of "competitive" reactors, ,

Table 6.8. Inventories of Nuclear Materials -1000 Molten S a l t Converter Reference Design Reactor

Mwe

Inventories, kg

Material

232Th 3 ~ a

-

2 3 3 ~ 234"

-

23 2 3 6 ~ 237 NP 2 3 8 ~

5

Salt Fission products Corrosion products ~~

~~~

Reactor Plant

Processing

Total

80,000 95.5 2110 484 420 682 49 109 109,000 4,400 82

6500 7.8 172 39.6 45* 56 5 11 8,900 360 7

86,500 1 03 2,282 524 465 738 54 120 117,900 4,760 89

~

*Including 10.6 kg i n reserve t o keep r e a c t o r i n operation f o r 30 days after unscheduled i n t e r r u p t i o n of recycle from processing plant.


148 The processing rates given i n Table 6.9 are t h e amounts removed d a i l y from t h e c i r c u l a t i n g f u e l stream. These d a i l y increments are accumulated f o r 116 days t o form a processing batch.

Is,

The average f i s s i o n pro-

duct a c t i v i t y i n t h e material as processed corresponds t o an e f f e c t i v e holding time of about 90 days (based on the Way-Wigner c o r r e l a t i o n as reported i n reference 55, p. 81). Thus, a process batch contains 6500 kg of thorium; t h i s is processed a t t h e rate of 1000 kg/day f o r uranium recovery, taking 1\17 days t o process and two days for "turn-around."

Process and Make-up Rates -1000 Mwe Table 6.9. Molten S a l t Converter Reference Design Reactor \

Rate, kg/day

Material

2 32Th

*233"3 ~ a 234u 23 5" 236u

237 NP 238u

Salt Fission products Corrosion products

6.5.4

To Processing

Make-up

53.3 0,064 1 400 0 322 0.280 0.463 0 033 0.073 72.7 2.94

.. .

3.66

.

0 003 0 324

0.055

Fuel Cycle Cost

The f u e l cycle c o s t for t h e reference design r e a c t o r , which is near optimum on t h e bases chosen, comes t o 0.68 mills/kwhre. A breakdown is given i n T a b l e 6.10. Inventory of fissile materials costs about 1/11 mills/kwhre, when optimized with respect t o t h e processing rate. It could possibly be reduced a t a given processing rate by improving t h e graphite t o reduce xenon poisoning and by reducing t h e volume of f u e l i n t h e external heat transfer c i r c u i t a t t h e expense of g r e a t e r pumping power costs.

P

Wc


149 Fuel Cycle Cost -1000 Mwe Molten S a l t Converter Table 6.10. Reference Design Reactor with S a l t Discard Charges,* mills/kwhre

Material Inventory 2 32m

Replacement

0.03 0,Ol 0.18 0,04

233~a 233u 23 5u

Total Salt costs Total charges

Processing

Total

0.04

0.26

-

0.06

0.08

0.16 0.20

[0.08]

0.08

0.54

0,14 0.68

*Cost bases are given i n Sec. 3

Replacement c o s t s f o r f i s s i l e material a r e most s e n s i t i v e t o conversion r a t i o .

Increasing t h i s from 0.90, as i n t h e reference design r e a c t o r ,

t o 0.95 would cut t h e cost i n h a l f ,

The increase i n conversion r a t i o might

be achieved by means discussed i n Sec. 6.7.3. The inventory charge f o r s a l t is s t r i c t l y a function of t h e f u e l s a l t volume.

The replacement cost depends on t h e processing r a t e , and was

optimized.

A major improvement here would c o n s i s t of adding equipment i n

t h e processing p l a n t f o r recovering lithium and beryllium.

The f u e l cycle c o s t of 0.68 mills/kwhre estimated for t h e MSCR is cons e r v a t i v e l y based on t h e scale-up of proven technology (including t h e processing) with t h e following minor exceptions:

1.

Technology of r e a c t o r vessel design.

2.

Graphite technology,

3.

Xenon sparging technology

These, however, are considered within t h e reach of current technology, i.e., no developmental break-throughs are required. Although d i f f i c u l t problems i n gamma heating may be encountered i n t h e design of t h e r e a c t o r v e s s e l and its i n t e r n a l members, it appears t h a t

u

these can be solved.

The problem i n r e l a t i o n t o graphite is t o produce


150 pieces of t h e size required t h a t are chemically compatible with f u e l s a l t and have porosity and permeability suitably.low with respect t o xenon absorption (Sec. 6.8.2).

6,6

Parameter Studies

I n t h i s section, t h e effect on t h e fuel cycle c o s t of various assumpt i o n s concerning t h e processing c o s t , and of s e v e r a l modes of processing

are considered. 6.6.1

Processing Cost.asâ&#x20AC;&#x2122;Parameter The f u e l cycle c o s t reported i n Sec. 6.7 (Table 6.10) was based on t h e

assumed use of a central Fluoride V o l a t i l i t y f a c i l i t y requiring approximately $20 million i n c a p i t a l investment and having an annual operating ex-

I n Sec. 5.4, t h i s estimate was compared with o t h e r current estimates f o r s i m i l a r p l a n t s and f o r Thorex p l a n t s of pense of $2 m i l l i o n (Table 5.7). comparable capacity.

I n t h i s s e c t i o n , t h e processing p l a n t c a p i t a l invest-

ment and t h e operating cost are considered as parameters, without regard t o t h e kind of plant. The effect on t h e f u e l cycle c o s t of t h e MSCR a t various processing rates of varying processing c o s t s f o r a p l a n t having a nominal capacity of 1000 kg Th/day (30 f t 3/day of reference design s a l t )

was calculated.

I n all cases, t h e output of t h e r e a c t o r was accumulated

u n t i l a processing batch of about 6000 kg of Th was collected; t h i s was then shipped t o t h e processing p l a n t and processed a t a rate of 1000 kg Thlday. The turn-around time was assumed t o be 2 days, The r e s u l t s of t h e c a l c u l a t i o n are presented i n Fig, 6.2. corresponding t o d a i l y charges of -$20,000 â&#x20AC;&#x2DC; t o $60,000.

The curves

It is seen t h a t , i n

t h e conversion r a t i o range from 0.8 t o 0,9, t h e f u e l cycle c o s t is not very s e n s i t i v e t o t h e processing cost; i n t h e reference design, t h e f u e l cycle c o s t does not exceed 0.85 mills/kuhre a t a d a i l y charge of $40 thousand, and only s l i g h t l y exceeds 1.0 mills/kwhre for a charge of $60 thousand.

At

$40 thousand, t h e minimum f u e l cycle cost is less than 0.8 mills/kwhre a t a

conversion ratio of about 0 . 8 5 , and t h e cost remains below 1,O mills/kwhre f o r conversion ratios up t o about 0.92,


151

Om-DWG.

2.0

65-791

1.8

1.6

1.4

1.2

1.0

P X) $60,000 daily

0.8

I

$40,000

0.6 0.80

0.85

0.90

CONVERSION W T I O

Fig. 6.2.

MSCR Processing Cost Versus Conversion Rat io.

0.95


. 152

,w -

It w a s concluded t h a t , on any reasonable cost b a s i s , t h e optimum f u e l

cycle cost f o r t h e MSCR w i l l not exceed 0.8 mills/kwhre, and t h a t conversion r a t i o s up t o 0.92 can be obtained a t f u e l cycle c o s t s not exceeding 1.0 mills/kwhre. 6.6.2

Effect of Xenon Removal

The s o l u b i l i t y of xenon i n LiF-BeF2 is very low (107); i n t h e reference design reactor t h e equilibrium pressure is about 0.06 atmospheres. Xenon thus tends t o leave t h e s a l t a t any phase boundary,

It can be re-

moved rapidly by spraying a portion of t h e c i r c u l a t i n g stream i n t o a space f i l l e d with helium o r by subsurface sparging.

It may form microbubbles

clinging t o t h e surface of t h e graphite moderator, and it w i l l tend t o d i f f u s e i n t o pores i n t h e graphite, including pores inaccessible t o t h e salt.

Xenon is also removed from t h e system by decay t o 135Cs and by

r e a c t i o n with neutrons t o form Ij6Xe, which is s t a b l e and has a l o w neutron capture cross section. Xenon poisoning i n t h e MSCR was calculated by t h e method of Watson and Evans (1071, as shown i n Appendix G.

The important physical p r o p e r t i e s are

t h e porosity, e, of t h e graphite ( f r a c t i o n of graphite volume accessible t o xenon) and t h e d i f f u s i v i t y of xenon, D (cm2/sec), The key v a r i a b l e s are t h e diameter of t h e graphite logs, t h e f u e l c i r c u l a t i o n rate, and t h e sparging f r a c t i o n , r, ( f r a c t i o n of c i r c u l a t i n g stream sparged o r sprayed t o removed xenon).

I n t h e reference design, the logs are s i x inches i n diam-

eter, and t h e c i r c u l a t i o n rate is 160 ft3/sec, I n the reference design r e a c t o r , a gas-accessible porosity of 0,Ol and with a sparging f r a c t i o n of 0,1 (16 ft3/sec), a d i f f u s i o n c o e f f i c i e n t of would r e s u l t i n a tolerable xenon poison f r a c t i o n of 0,017 neutrons per neutron absorbed by fissile atoms.

These physical property values are an

order of magnitude smaller than those of c u r r e n t l y a v a i l a b l e graphite where e

= 0.1 and

I)

= 10â&#x20AC;?; however, t h e assumed values are both presently a t t a i n -

able i n small pieces of graphite (107) and t h e c o n t r o l of xenon poisoning i n t h e MSCR appears t o l i e within t h e reach of developing technology. The conversion r a t i o increases and t h e f u e l cycle c o s t decreases with decreasing xenon poisoning,

T a b l e 6.11 compares t h e calculated r e s u l t s f o r

I I

U

I


-L

153

u

Effect of Graphite Properties and Sparge Rate T a b l e 6.11. on Performance of 1000 Mwe Molten S a l t Converter Reactor . ... _ _

-~~~

Case

Conversion Ratio

Xe P,F.

Fuel Cycle Cost mills/kwhre

a

0.054

0.84

b

0 045

(O.86)*

C

0.017

0.90

0.68

d

0 001

0,92

0.66

. 0

0.79 (0.77 )*

*Interpolated

t h r e e cases:

(a)

case, with a very porous graphite (say AGOT) and

with no sparging; ( b ) available graphite with eD = Reference Design Reactor with eD = * 'ol

ant? r-=O-.l;

and r = 0.1;

(c)

(d) "impermeable"

graphite with eD = 0 and r = 0.1. It is seen t h a t while available graphite (Case b) is not s i g n i f i c a n t l y b e t t e r than t h e llworstll graphite (Case a), nevertheless, t h e f u e l cycle c o s t is only 0.1 mill/kwhre higher than f o r "impermeable" graphite (Case d). It is concluded t h a t t h e f u e l cost is not very s e n s i t i v e t o xenon poisoning, t h a t it w i l l be less than 0 , 8 mills/kwhre, with a v a i l a b l e graphite, and t h a t with modest improvements over a v a i l a b l e graphite (to e of 0.01 and D of lO"), 6.6.3

t h e fuel cost w i l l be not more t

Effect of Product Sale Without Recycle A t least two b e n e f i t s accrue from t h e sale of recovered f i s s i l e iso-

topes (as UF6) t o t h e AEC:

(a) Make-up f u e l (235U) would then be nonradioactive and could be compounded with f r e s h s a l t i n a d i r e c t l y maint a i n e d and operated f a c i l i t y ; (b) t h e reactor would tend t o be purged of 236U.

The second b e n e f i t is r e a l l y i l l u s o r y , inasmuch as t h e 236U prdduced

w i l l eventually capture a neutron i n some r e a c t o r somewhere, and t h e r e f o r e reduces t h e value of t h e recovered isotopes by an appropriate amount.

W

On t h e o t h e r hand a penalty is incurred i n t h a t 233U, a superior f u e l i n thermal r e a c t o r s , i s a l s o l o s t from t h e system. A s seen i n Table,6.12, t h e p e n a l t i e s outweigh t h e advantages considerably.


154 Effect of Sale of Spent Fuel on MSCR Performance

Table 6;12.

Case

A

B

Recycled

Sold

Absorptions i n 2 3 % ~

0,0185

0 0105

Mean Eta

2.21

2011

Conversion Ratio

0090

0.82

Inventory Charges

0.32

0036

Replacement Charges

0028

1.38

Processing Charges Production Credit

0.08

0.08

.

Spent f u e l is

Fuel Cycle Cost, mills/kwhre

N e t Fuel Cost, mills/kwhre

6.7

e-

0.68

0.53

1.29

Alternative Design and Cost Eases

Thorex Processing Cost Estimates

6.7.1

The preferred method of processing MSCR f u e l is by f l u o r i n a t i o n

(Sec. 5.2)

, mainly

because t h e processing can conveniently be integrated

with t h e r e a c t o r and thus achieve very l o w f u e l cycle c o s t s ,

However, f o r

one reason or another, it may be d e s i r a b l e t o process t h e spent f u e l from t h e first MSCR i n s t a l l a t i o n s i n a central Thorex f a c i l i t y .

Accordingly,

t h e f a c i l i t y s p e c i f i e d i n t h e Guide t o Nuclear Power Cost Evaluation (52,

104) was modified appropriately as described i n Sec. 5.3 t o handle MSCR fuel. An allowance of $500,000 a d d i t i o n a l c a p i t a l c o s t f o r t h e head-end treatment was made. Costs w e r e calculated on t h e bases given i n t h e Guide. The turn-around time was e i g h t days and t h e shipping charge was $17/kg for round t r i p .

The assumed loss of f i s s i l e material was 1.3 per cent/pass.

The f u e l cycle c o s t s for t h e reference design r e a c t o r (Tables 6,6 through 6.10) were recomputed using a cost schedule estimated f r o m t h e

-

Guide (521, as shown i n Table 6,13, where they &e compared t o correspondi n g r e s u l t s f o r c e n t r a l Floride V o l a t i l i t y processing (Sec. 5.2).


155 Effect of Processing Method

Table 6.13.

'on Fuel Cycle Cost Central Thorex Processing c o s t , $/kg Th Processing l o s s e s , per cent/pass Conversion Ratio

Central Fluoride Volatility

75,2

44.8

1.3

003

0.89

0.90

Inventory charges

0033

0032

Replacement charges

Oa24

0.28

Processing charges

0.17

-

0,74

0.68

Fuel cycle c o s t , mills/kwh

Total, mills/kwh

0008

Thus, even though processed through t h e AEC reference p l a n t , t h e c o s t s of which appear t o be conservatively high, t h e f u e l cycle cost f o r t h e r e f erence design MSCR w i l l not exceed 0 , 7 5 mills/kwh, . .. .. .. .. . 6 . 7 . 2 .Redctd~-Integrated.FluorideV o l a t i l i t y Processing Several advantages can be r e a l i z e d by i n t e g r a t i n g t h e pmcessing with ". The p r i n c i p a l saving r e s u l t s from sharing r e a c t o r s h i e l d i n g and remote maintenance equipment. Savings i n laboratory f a c i l -

r e a c t o r operations.

i t i e s and personnel are also-important.

Shipping c o s t s and associated re-

ceiving f a c i l i t i e s are eliminated, 3 A 30 f t /day central f l u o r i d e v o l a t i l i t y plant requires a c a p i t a l outl a y of about $25 million and would need t o s e r v i c e f i f t e e n t o twenty

1000 Mwe MSCR's i n order t o achieve t h e u n i t processing c o s t s estimated i n

Section 5.2.

An integrated p l a n t r e q u i r e s a much smaller investment and

t h e u n i t processing cost w i l l be, of course, independent of t h e number of r e a c t o r s i n useo An integrated f a c i l i t y should be designed for continuous flow process-

ing, a t least i n t h e f l u o r h a t o r and t h e UF6 reduction r e a c t o r , i n order t h a t t h e equipment and t h e volume of f u e l - s a l t held up might both be small.


f

156

Reactor-1ntegrated.Pecipitation Process

6.7.3

3 A t a processing rate of 1.7 f t /day, t h e concentration of rare e a r t h

f l u o r i d e s i n t h e f u e l salt w i l l be approximately 0.5 mole per cent, which

is perhaps s l i g h t l y i n excess of t h e s o l u b i l i t y l i m i t a t t h e minimum f u e l temperature of llOO째Fo This suggests t h a t it may be possible t o remove rare e a r t h s f i s s i o n products from t h e MSCR f u e l stream by f r a c t i o n a l cryst a l l i z a t i o n -an extremely a t t r a c t i v e p o s s i b i l i t y , for such a process could conveniently be c a r r i e d out i n t h e r e a c t o r c e l l and c l o s e l y integrated with t h e reactor system,

The s t e p s involved are exceedingly simple, involving

only t h e t r a n s f e r of l i q u i d s and heat, and is t h e r e f o r e inherently safe and economical. The f i s s i o n product neutron poisoning estimated for t h e reference des i g n can be approximately matched by charging every day f i v e cubic feet of f u e l s a l t containing 0.5 mole per cent rare e a r t h t h o r i d e s t o a c r y s t a l -

lizer.

The salt is cooled t o 900째F, which is 13OF above t h e temperature a t

which s o l i d solutions containing Th or U separate.

A t t h i s temperature,

t h e s o l u b i l i t y of t h e rare e a r t h f l u o r i d e s is 0.2 mole per cent or less, judging from t h e data of Ward e t a l , , (108, 106).

Thus about 60 per cent

of t h e rare e a r t h s w i l l p r e c i p i t a t e or l'freeze'l on t h e walls of t h e crystallizer.

The t o t a l mass of t h e s o l i d s w i l l be only 6-8 kg.

After t h e

f u e l s a l t is returned t o t h e reactor system, t h e f i s s i o n products are dissolved i n f l u s h salt The process described should e f f e c t i v e l y remove rare e a r t h s from t h e f u e l salt.

It w i l l not remove alkali metals, a l k a l i n e e a r t h s , and m i s -

cellaneous o t h e r metals,

These w i l l accumulate i n t h e fuel, but t h e i r in-

. growth can be p a r t i a l l y compensated by operating t h e freeze-process a t a s l i g h t l y more r a p i d rate and maintaining t h e rare e a r t h concentration a t say 0.45 mole per cent or by discarding barren f u e l - s a l t i n a 1500 day cycle

.

608 Evolution of a Self-sustaining MSCR r

Although t h e MSCR concept may not have t h e c a p a b i l i t y of evolving i n t o a breeder r e a c t o r (see Sec. 7.1) having a doubling t i m e less than 25 years,

LJ


157 t h e p o s s i b i l i t y exists, however, t h a t it could, without increase i n power c o s t s , achieve a net conversion r a t i o s l i g h t l y g r e a t e r than unity, and thus become self-sustaining and independent of outside supplies of f i s s i l e isotopes,

Conditions under which t h i s might be achieved are l i s t e d and

discussed below,, 6.8.1

Reduction of Leakage With t h e advent of separated

92M0,

it becomes f e a s i b l e t o surround t h e

core of t h e MSCR with a t h i n blanket of high-density thorium s a l t ( 2 5 % ThF,,, 75% LiF).

This s a l t , having a liquidus temperature below llOO째F, could be

c i r c u l a t e d slowly through 6-inch diameter molybdenum tubes replacing t h e o u t e r two l a y e r s of graphite moderator logs, and replacing 1-ft end sect i o n s of t h e central logs.

The isotope

2200 m / s cross s e c t i o n of 6 millibarns.

is a magic nuclide, having a The epithermal cross s e c t i o n is

92M0

c u r r e n t l y being measured a t OWL, and preliminary r e s u l t s i n d i c a t e t h a t t h e resonance i n t e g r a l is a l s o very small. Thus, s t r u c t u r a l molybdenum should capture only a n e g l i g i b l e f r a c t i o n of neutrons. It would be necessary t o remove t h e bred 233U r a p i d l y from t h e f e r t i l e

stream f o r two reasons:

( a ) f i s s i o n s i n t h e f e r t i l e stream would tend t o

increase t h e leakage, (b) f i s s i o n products i n t h e f e r t i l e stream would capture neutrons.

If t h e f e r t i l e stream were processed r a p i d l y by f l u o r i -

nation, t h e concentration of 233U could be kept very low, f i s s i o n s would be suppressed, t h e inventory charge f o r 233U would be l a r g e l y avoided. The f e r t i l e stream carrier s a l t could be recycled without f u r t h e r treatment. , The r a p i d i t y of t h e processing, however, implies t h e use of an on-site,

reactor-integrated f a c i l i t y such as t h a t described i n Sec. 6.9.2, *Out

By t h i s means perhaps half t h e leakage neutrons could be saved, adding 0 . 0 2 5 u n i t s t o t h e conversion ratio. i

6.8,2

Reduction 'of Xenon 'Captures I n t h e reference design, graphite p r o p e r t i e s an order of magnitude

b e t t e r than those c h a r a c t e r i s t i c of current graphites were assumed, r e s u l t ing i n a loss i n conversion r a t i o due t o captures i n 135Xe absorbed i n t h e

LiJ

graphite of only 0.017 units.

By f u r t h e r improvements (e.g.,

by spraying a

t h i n coating of 92M0 on t h e moderator logs and carburizing t o prevent xenon

- -

I


.

,

158

c.

from penetrating t h e moderator) perhaps another 0.01 u n i t s on t h e conversion r a t i o could be saved. Reduction of Fissfon.Product Poisoning

6.8.3

Somewhat over 0.085 u n i t s were l o s t from the conversion r a t i o as a r e s u l t of captures i n f i s s i o n products i n t h e reference design ( T a b l e 6.7). The processing c o s t i n a c e n t r a l Fluoride V o l a t i l i t y f a c i l i t y was about 0.08 mills/kwhr,

If an on-site, reactor-integrated Fluoride V o l a t i l i t y and

HF-Solution f a c i l i t y were used (Sec. 2.2,14),

t h e rate of removal of rare

e a r t h s could be increased perhaps by a f a c t o r of 10, Solubles ( C s , Ba, etc.) could be purged by discarding s a l t i n a 1500 day cycle, as i n t h e reference design. If t h i s were done, t h e loss of conversion r a t i o t o f i s s i o n products could be reduced t o about 0,010 u n i t s . Improvement of Mean 'Eta and Reduction of 236U Captures

6.8.4

Other than by varying t h e C/Th r a t i o , t h e designer has no d i r e c t con-

t r o l over these. Nevertheless, t h e above improvements i n neutron economy have an effect on t h e conversion r a t i o t h a t is g r e a t e r than t h e i r cumulat i v e sum, f o r 233U is superior t o 235U i n respect t o neutron production, and moreover y i e l d s a f e r t i l e isotope (234U) upon capture of a neutron. An increase i n r e l a t i v e concentration of 233U by any means increases t h e

number of neutrons a v a i l a b l e for breeding, and reduction of 235U feed rate r e s u l t s i n a decrease i n 236U concentration, The above l i s t e d improvements should r e s u l t i n an increase i n eta of 0.01 u n i t s and a reduction of captures i n 236U by 0.01 u n i t s , a t least. . .. ... . . . g ' P o t e n t i a 1 of MSCR f

Ultimate Bree

6,8.5

The conversion r a t i o _. i n t h e reference design is about 0,900

With improvements l i s t e d above, conversions s l i g h t l y i n excess of 1,O may be .- . achieved, as shown i n Table 60140 Thus, t h e MSCR may be capable of evolving stepwise i n t o an economidal, self-sustaining breeder r e a c t o r with f u e l cycle c o s t s probably i n t h e range of 0.7

- 1,O

mills/kwhre.

It should be emphasized t h a t t h e l i m i t i n g conversion r a t i o estimated

f o r - t h e MSCR does not apply t o molten salt breeder r e a c t o r s ,

It has been

w-


159

i

Table 6.14.

Ultimate Breeding P o t e n t i a l of Molten S a l t Converter Concept

Conversion r a t i o i n reference design

0.90

Savings due to: Reduct ion i n leakage

0,025

Reduction i n xenon captures

0,010

Reduction of f i s s i o n product captures

0 075

Improvement i n eta and reduction of captures i n 236U

.

0.020 7

Ultimate conversion r a t i o

1.03

shown ( 3 1 t h a t two-region, two-fluid, thermal r e a c t o r s optimized with respect t o breeding are capable of achieving doubling times of 25 years or less, (See also Sec. 1.7.) I n addition, t h e advent af s t r u c t u r a l 92M0

makes possible, i n principle,.the design of two-region, two-fluid, fast molten-salt reactors which may have doubling times as s h o r t as t e n years, which w i l l not need t o use separated 6 L i i n t h e carrier salt, and which t h e r e f o r e may be processed economically by f l u o r i n a t i o n only.


,

. 160

h

W8 7.

MSCR CAPITAL INVESTMENT, FIXED CHARGES, AND OPERATING EXPENSE 7.1

Introduction

The equipment, a u x i l i a r i e s , and a u x i l i a r y s e r v i c e s described i n Sec, 4

were costed f o r ORNL by Sargent and Lundy Engineers of Chicago, I l l i n o i s (95, 96). Equipment arrangement drawings s u f f i c i e n t for piping take-offs and building cost estimates were made.

Details of t h e s e s t u d i e s are given

i n t h e referenced r e p o r t s , .... .

7.2

.Summar?y'ofMSCR Capital Investment

I n T a b l e 7.1 are l i s t e d t h e p r i n c i p a l i t e m s of c o s t i n t h e MSCR reference design.

A d e t a i l e d breakdown i s given i n Appendix N o The account

numbers correspond, where applicable, t o t h e AEC systems of accounts given i n The Guide t o Nuclear'Power Cost Evaluation (52).

Table 7.1. 1000 Mwe Molten S a l t Converter Reactor Capital Investment 2500 Mwt 1038 Mwe 1083 Mwe 41,5% 8220 Btu/kwhr

Fission energy release rate, Mwt Net station power Gross s t a t i o n power Station efficiency Heat rate Plant factor Total c a p i t a l investment

0.8

$143/kwe

Direct Construction Costs 2 1 Structures and Improvements

2U. Improvements t o s i t e 212 Buildings 218 Stacks Reactor Container Total Account 2 1

$

501,500 5,465,450 31,000

(Included i n 212) (5,997,950)

.Lid*

c


161 Table 7.1, 22

Continued

Reactor P l a n t ' 221 222 223 224 225 226 227 228 229

Reactor equipment Heat transfer system F u e l f a b r i c a t i o n and handling system F u e l p r o c e s s i n g system waste d i s p o s a l Low-level r a d i o a c t i v e waste d i s p o s a l I n s t r u m e n t a t i o n and controls Feed water supply Steam, condensate, and water p i p i n g Other r e a c t o r equipment

Total Account 22 23

Turbo-generator u n i t C i r c u l a t i n g water system Condensers and auxiliaries Central lubrication system Turbine p l a n t instruments E c o n t r o l s Turbine plant p i p i n g A u x i l i a r y equipment f o r g e n e r a t o r s Other equipment

Total Account 23

(51,324,350) 21,495,000 1,644,200 3,104,900 36,000 426,000 ( I n c l u d e d i n 228) 137,000 ( I n c l u d e d i n 228) 26,843,700

Accessory Electrical Equipment 241 242 243 244 245 246 247

25

8,823,300 23,609,700 1,517,200 (Not i n c l u d e d ) 361,150 1,100,000 4,939,500 7,925,000 3,048,500

Energy Conversion System 231 232 233 234 235 236 237 238

24

$

Switchgear Switchboards P r o t e c t i o n equipment Conduit Power and control w i r i n g S t a t i o n service equipment

637,400 286,000 131,600 213,200 210,200 2,281,900 615,000

Total Account 24

4,375,300

Electrical structures

Miscellaneous P l a n t Equipment 251 Cranes and h o i s t s 252 A i r compressors and vacuum pumps 253

Other

Total Account 25 Total Direct Construction Cost

195,000 64,900 540,000 799,900 89,341,200

I n d i r e c t C o n s t r u c t i o n Costs

i

Construction overhead (20% of d i r e c t l a b o r )

V ' i

General and a d m i n i s t r a t i o n (2,5% of d i r e c t c o s t s ) Subtotal

2,333,300 5,775,500 (97,450,000)


-.

-

9.

162 Table 7.1. .

......

.

n

w*

Continued

Miscellaneous c o s t s (1.2% of s u b t o t a l ) Subt o t a1 Engineering Dssign and Inspection Architect-engineers (11.1% of s u b t o t a l ) Subtotal Nuclear engineers (3.8% of subtotal) Start-up expense (35% annual O&M expense) Land and land r i g h t s Subtotal Contingency (10% of s u b t o t a l ) Subt o t a l I n t e r e s t during construction (9.4% of s u b t o t a l ) Total I n d i r e c t Construction Costs Total Construction Costs

$

,

974 500 (98,424,500) 10,925,000

, ,

(109 349 500 ) 4,155,300 746,900 360,000 (114,611,700) 11,461,200 (126,O72,900) 11,850,900 48,582,600 137,923,800

Intermediate Coolant- Salt Inventory* TOTAL MSCR CAPITAL INVESTMENT

10,951,800

,

$ $48 875,600

*Including i n t e r e s t during s t a r t u p

7.3

MSCR Fixed Charges c

The fixed charges were computed i n accordance with t h e i n s t r u c t i o n s

i n The Guide t o Nuclear Power Cost Evaluation, Vol. 5, Production Costs, page 510-2 (52), for an investor-owned public u t i l i t y , and are shown i n Tables 7.2 and 7.3.


163 Table 7.2.

Nation-Wide .Approximated Fixed Charge Rates Percent P e r Year Depreciating

Non-Depreciating

@ o f i t on investment

6,75

6.75

Depreciation (30-yr sinking fuqd)

1.11

Interim replacements

0.35

--

Property insurance

0.40

0.40

Federal income taxes

3.40

3,40

S t a t e and l o c a l taxes

2.45

2.45

14 46

13,OO

Total

Table 7,3.

--

1000 Mwe Molten S a l t Converter Reactor Fixed Charges

Item

Depreciating c a p i t a l Non-depreciating c a p i t a l Land, etc. Coolant Working c a p i t a l Nuclear insurance Annual fixed charges

Investment 137,600,000 360,000

Rate Annual Expense %/yr S/yr 14.46 13.0

19,950,000

,

Power Cost Mills/kwhre 2.74

1 530,000

0.21

340,000

0.05

450,000

..--

21,820,000

3.0


164 ...

r . . . . .

MSCR Operating'and-MaintenanceCost Estimate

7.4

The MSCR is a s i n g l e reactor, s i n g l e t u r b i n e p l a n t , and, although i t s I

power generation capacity is high, i t s manpower requirements are r e l a t i v e l y low because of t h e s i n g l e u n i t operation, a c o s t of $872,000 p e r year,

Plant personnel t o t a l s 1 0 1 with

Materials cost $220,000,

cluding provision f o r periodic equipment overhaul

Maintenance, in-

, s p e c i a l s e r v i c e s pro-

vided by off-site personnel and organizations, t o t a l s $800,000,

With an

allowance of 14 percent for c e n t r a l office expense, t h e t o t a l cost is $2,154,000 or 0,30 m/kwh,

'The Guide t o Nuclear Power Cost Evaluation (52)

criteria were followed where applicable i n determining p l a n t organization. 7,4,1

.

Labor and Materlals I

The manpower requirements of t h e 1000-Mwe MSCR plant are shown i n supervision. Routine operation of t h e Table 7.4 for operations and, general . plant may r e q u i r e fewer people, p a r t i c u l a r l y on t h e t e c h n i c a l s t a f f ,

Re-

view of reactor p l a n t personnel requirements developed by Sargent and Lundy (941, and Kaiser Engineers (51) f o r 300 Mwe ( n e t ) p l a n t s are compared i n T a b l e 7.4 f o r s i n g l e u n i t systems.

I n p r a c t i c e , t h e actual d i s t r i b u t i o n of manpower may s h i f t , but t h e

For example, one storekeeper may be i n s u f f i c i e n t , i n which case an engineering a s s i s t a n t or

t o t a l l a b o r c o s t should remain approximately as shown.

maintenance mechanic helper might be replaced by a s t o r e s c l e r k , I n general, t h e turbine room operation includes, i n addition t o t h e

. )

,

turbo-generator proper, (a) water supply and disposal systems, e o g o sani-

.

t a r y s e r v i c e , t r e a t e d and c i r c u l a t i n g water systems, ( b ) b o i l e r feed systems which include b o i l e r feed pumps, steam c i r c u l a t o r s , and Loeffler b o i l e r s , and (c) t h e turbine a u x i l i a r i e s , e , g o , l u b r i c a t i n g o i l systems, l i q u i d and gas coolant systems, and instrument and compressed a i r systems. The reactor operation includes (a) salt charging systems, (b) salt withdrawal systems, (c) salt shipping f a c i l i t i e s , (d) l i q u i d and gaseous waste d i s p o s a l systems, and (e) pressurized gas supply and treatment systems, These f a c i l i t i e s are s t a f f e d on a semi-automated b a s i s ; f o r f u l l y automated operation, t h e manpower requirements would be less.

.-.

uf


165 Table 7.4.

Personnel Requirement Estimates

Sargent & Lundy 150 MW 350 MW

-

Kaiser 300 MW

AEC Guide 300 Mw(PWR)

MSCR 1000 MW

9

4,

4

7

P l a n t Management, Office & S t o r e s Operating Dspt

47

32-39

36

49

Technical S t a f f

26

6

6

16

Maintenance Dept e

13 95

Total

7.4.2.

-

-

20

-

62-69

59

29

13

101

Operation and Maintenance Cost The estimate o f $2,154,000 shown i n Table 7.5 does not t a k e i n t o

account any unforeseen d i f f i c u l t i e s and expenses t h a t may be encountered i n t h e MSCR.

It may well be t h a t requirements f o r personnel and equipment

for maintaining a r a d i o a c t i v e molten s a l t system are greater than estimated

-

p o s s i b l y by as much as a f a c t o r of t h r e e ,

This cannot be a c c u r a t e l y

determined u n t i l maintenance procedures have been more c l e a r l y defined and t h e r e a c t o r p l a n t designed i n greater d e t a i l than was p o s s i b l e i n t h e present study.

Based on 1038 Mwe net and a p l a n t f a c t o r of 0.8, t h e c o n t r i b u t i o n t o the power cost is 0,3 mills/kwhre.

T a b l e 7.5.

1000 Mwe MSCR Annual Operating and Maintenance Expense Salary or Wage Rate

Annual

Personnel Required

Expense

Wages E S a l a r i e s P l a n t Management S t a t i o n Supt ,.

Ass 't Supt , Clerk-Steno Clerk-Typist Clerk-Steno

$ 15,00O/yr 12,OOO/yr 2,50/hr 2 31/hr 2 50/hr

1

2

$

15,000

1 2 1

24,000 5,200 9,600 5,200

7

59,000

-


t

166

A

1

Table 7.5. . . . .. . . .

Continued Salary or Wage Rate

Personnel Required

Annual Expense

Technical S t a f f Supv. Eng. (L)*

Nuclear Eng. (L) Engineer Health Physics Supv. Eng. A s s ' t . Lab Technician Radiation Protection

$ 11,00O/yr 9,60O/yr 8,400/yr 8,40O/yr 6,00O/yr 2.85/hr 3 25/hr

$

5 2

11,000 9,600 25,200 8,400 18,000 29,700 13,500

16

115,400

5 6 4

54,000 46,800 30,400

10

72,800

8 9 2

50,000 65,500 9,400 23,400

1 1 3 1 3

-

Operating S t a f f S h i f t Supto (L) Senior Control oper. (L) Control Qer. (L) Turbine Oper. Equipment Attendant S p e c i a l Operator (L)

Janitor Watchman

Maintenance S t a f f Maintenance Supt. Foreman

Instrument Mechanic Electrician Pipe Fitter-Welder Machinist Mechanic Helper

10,80O/yr 3 75 /hr 3.65/hr 3 SO/hr 3 OO/hr 3.50/hr 2,25/hr 2 25/hr

.

.

.

10,80O/yr 7 ,SOO/yr 3.25/hr 3.25/hr 3.25/hr 3 25/hr 3 25/hr 2 65/hr

.. .

Total Labor

49 5

352,300

2

10,800 22,500 40,600 33,800 13,500 13,500 54,100 11,100

29

200,000

101

725,700

1 3 6 5

2 2 8

-

P

Fringe Benefits a t 20%

145,300

T o t a l Wages & S a l a r i e s

872,000

f

Materials for Routine Operations O i l Supply Gas Supply Treated Water Coolant S a l t Make-up Office Supplies Laboratory Supplies & Chem.

84,000 4,000 2,000 40,000 15,000 5,000

-.

W--

+(L) Denotes l i c e n s e d r e a c t o r operator. 0


167

Continued

T a b l e 7.5.

~

Salary o r Wage Rate

Personnel Required

Miscellaneous (e.g., r a d i a t i o n p r o t e c t i o n , c l o t h i n g E equip. Consulting Services Subtotal

~~

Annual Expense $

50,000 10,000

210,000

Contingency

10,000

T o t a l Materials

220,000

Maintenance Turbine & t u r b i n e a u x i l i a r i e s r o u t i n e maintenance materials Reactor E r e a c t o r a u x i l i a r i e s r o u t i n e maintenance materials Turbine 3-yr overhaul ( p r o r a t a ) Turbine system auxiliaries overhaul Reactor system overhaul Reactor auxiliaries overhaul

150,000 300,000

50,000 50,000 100,000 50,000

Subtotal

700,000

Contingency

100,000

T o t a l Maintenance

800,000

C e n t r a l Office, General E Admin, Expenses a t 14 Percent Grand T o t a l Operating E Maintenance Expense

,

262,000 $2,154,000


168 8.

RESULTS AND CONCLUSIONS

The c o s t of electric power i s commonly resolved i n t o t h r e e components: The f u e l c o s t , t h e fixed charges, and operation and maintenance expense.

8,1

Fuel Cost

As shown i n Sec. 6, t h e f u e l cycle cost i n t h e MSCR ranges from 2 mills/kwhr (electrical) down t o 0.7 depending on t h e conversion r a t i o

desired and t h e method and/or c o s t of processing assumed.

For present pur-

poses, t h e minimum c o s t associated with t h e reference design r e a c t o r summar i z e d i n Table 6.6 was s e l e c t e d as representative,

This r e a c t o r is "near-

term" and predicated on t h e scale-up of current technology with t h e exception of t h e moderator graphite, which was an order of magnitude b e t t e r than c u r r e n t l y available. graphite i n respect t o porosity and permeability. The f u e l cycle c o s t , using Fluoride V o l a t i l i t y processing i n a central

p l a n t with discard of carrier salt and contained thorium and recycle o f isotopes of uranium, was reported i n Sec. 6.7,l.

Table 8.1 1000 Mwe Molten S a l t Converter ,.Reactor Fuel Cycle Cost

Item

Mills/kwhr ~~

Inventories Fertile Fissile Salt

0.03 0.23 0.06

Replacement Fertile F i s s i l e (95% 235U) Salt Reprocessing Total, mills/kwhr

0,04 0.16 O,O8 0.08

0.32

-

0.28

0.08

0.68


169 8.2

. Fixed

'

Charges

The c a p i t a l investment f o r t h e 1000 Mwe (1038 Mwe n e t ) s t a t i o n was e s t i m a t e d by Sargent and Lundy, Engineers from information s u p p l i e d by O W L , as r e p o r t e d i n Sec, 7, and summarized i n Table 7.1. The investment comprised $137,564,000 for d e p r e c i a t i n g c a p i t a l items, and $11,311,800 f o r Fuel s a l t f i x e d charges

non-depreciating items ( c o o l a n t s a l t and l a n d ) .

are included i n f u e l c y c l e cost, Working c a p i t a l was estimated according t o t h e p r e s c r i p t i o n given i n

-

t h e Guide (521, and i s shown i n Table 8.2.

1000 Mwe MSCR Working C a p i t a l

Table 8,2, 1.

2.7% of annual o p e r a t i n g labor and f u e l c o s t s :

2.

25% of annual maintenance and materials:

Total Working C a p i t a l

$ 200,000

250,000 $ 450,000

S i m i l a r l y , t h e n u c l e a r hazard insurance premium was e s t i m a t e d by p r e s c r i p t i o n a t $340,000 p e r year.

The f i x e d charges are c o l l e c t e d i n

Table 8.3.

1000 Mwe Molten S a l t Converter

Table 8.3,

Reactor Fixed Charges Investment

Item

Depreciating c a p i t a l

Rate Annual Expense Power Cost Mills/kwhr %/yr S/yr

137,600,000

14.46

19,950,000

2.74

11,760,000

13,O

1,530,000

0.21

340,000 21,820,000

0.05 3,O

Non-depreciating c a p i t a l

Nuclear i n s u r a n c e Annual f i x e d charges


170 8.3

Operation .and 'Maintenance Expense

This expense was estimated i n Sec, 7,2 and amounted t o $2,154,000 per year, of which $1,020,000 was f o r maintenance and materials and t h e rest

was f o r labor, supervision, and management, A t 1038 Mwe n e t , t h e contribution t o t h e power cost is 0.3 mills/kwhr (electrical),

8.4 .Cost of Power The t h r e e components of t h e power cost are assembled i n T a b l e 8.4.

T a b l e 8*&. Cost of Power i n a 1000 Mwe Molten S a l t Converter Reactor

Item

Mills/Kwhr

Fuel cycle cost

0.7

Fixed charges

3.0

Operation and maintenance Cost of power, mills/Kwhr

0.3

400

I n regard t o t h e f u e l cycle c o s t , it was shown i n Sec. 608 t h a t t h i s would not exceed 1.0 mill/kwhr even though t h e f u e l were processed i n a 1.0 tonne/day Thorex plant costing up t o $60 thousand per day t o operate.

Such a general purpose plant could provide processing of f u e l from thorium

reactors having a t o t a l power c a p a b i l i t y of about 20,000 Mwe a t an exposure of 50,000 Mwd/tonne , The energy conversion system, accessory electrical equipment, and miscellaneous plant equipment (Items 23, 24, and 25 i n Table 7.1) are conventional items, and t h e estimation of t h e i r c o s t appears t o be r e l a t i v e l y unambiguous. tainty.

Items 2 1 and 22 are perhaps subject t o considerable uncer-

But t h e i r t o t a l is only about $57 million, whereas t h e contingency

i t e m is $11 million,

However, t h e p o s s i b i l i t y e x i s t s t h a t Item 21 was

--

W

.


171 grossly underestimated, or t h a t t h e effect of r a d i a t i o n shielding requirements on building c o s t s was underestimated,

Supposing Item 2 1 t o be

$18 million ( f a c t o r of 31, and allowing another $5 million f o r t h e r e a c t o r plant

(lo%),

t h e investment comes t o $166/kw, and the fixed charges t o

3.5 mills/kwhr, as an upper l i m i t . Provision f o r labor i n Sec. 7.2 seems adequate,

Materials and supplies

accounted for about h a l f t h e operation and maintenance c o s t s ; i f t h i s were doubled, t h e contribution t o t h e cost of power would be about 0.5 mills/kwhr. Collecting these probable upper l i m i t s on t h e cost of power i n t h e MSCR, t h e sum is 5.0 mills/kwhr.

8.5

Breeding P o t e n t i a l of t h e MSCR

The nuclear c a p a b i l i t y of t h e reference design MSCR is summarized i n Fig, 8.1.

With c e n t r a l Fluoride V o l a t i l i t y processing, t h e minimum f u e l

cycle c o s t is about 0.7 mill/kwhr and t h e conversion r a t i o about 0.9 a t a processing cycle time of 1500 days,

Increasing t h e rate increases t h e con-

version r a t i o t o perhaps as high as 0.93, but t h e f u e l c o s t rises steeply. The use of an on-site reactor-integrated ( i n s i d e t h e r e a c t o r c e l l ) Fluoride V o l a t i l i t y f a c i l i t y would increase t h e f u e l cost by only 0.1 mill/kwhr i n t h e 1000 Mwe s t a t i o n , The use of an on-site reactor-integrated p r e c i p i t a t i o n process f o r removing rare e a r t h s might reduce t h e f u e l cost below 0.5 mills/kwhr, By taking advantage of p o t e n t i a l improvements i n neutron economy (but r e t a i n i n g t h e e s s e n t i a l f e a t u r e s of t h e MSCR) an upper l i m i t on t h e conversion r a t i o of 1,03 was estimated.

This was i n t e r p r e t e d t o mean t h a t

t h e MSCR is capable of evolving i n t o a self-sustaining r e a c t o r requiring I

only thorium feed,

Outside source of f i s s i l e isotopes would not be needed.

This l i m i t a t i o n does not apply t o two-region breeders, which were shown previously t o b

capable of doubling times as s h o r t as 25 years (1).


172

we *-

ORNL- LR- DWG

1. I

76712

I I

Q: 0

le0

0.9

0.0

0.7

#

b I

0.6

-

0.5

I

Symbols 0 Central Fluoride V o l a t j l i Processing ~ CI Full Xenon Poisoning 0 Xenon â&#x201A;Źxclusion Q Reference Design Case

I

I I 014

0.0

9

0 1

I10

I, I

CONVERSION RATIO

Fig. 8.1. Nuclear Capability of 1000 Mwe Molten Salt Converter Reactor.


173 8.6

'Conclusions

The 1000 Mwe MSCR requires a c a p i t a l investment of S143Jkwhr.

The

fixed charges are 3.0 mills/kwhr, t h e f u e l cost is 0,7 mills/kwhr, t h e maintenance and operation expense of 0 0 3 m i l l s / k w h r , and t h e n e t power cost i s 4.0 mills/kwhr. S u b s t i t u t i o n of sodium f o r LiF-BeF2 as t h e intermediate coolant and replacement of t h e Loeffler boiler-superheater complex with a more conventional sodium-heated b o i l e r would reduce t h e c a p i t a l investment t o about $125/kwhr,

Development of a l t e r n a t i v e methods of processing (e .g.

, preci-

p i t a t i o n of rare e a r t h f l u o r i d e s ) provides p o t e n t i a l f o r reducing f u e l cycle c o s t ,

An examination of t h e neutron economy i n d i c a t e s t h a t t h e MSCR

should be capable of evolving i n t o a self-sustaining breeder r e a c t o r

(BR

% l o o ) not

dependent on outside sources of f i s s i l e isotopes.

8 . 7 ' .Recommendations

The comprehensive program of research and development f o r molten s a l t r e a c t o r s i n progress a t ORNL i s concerned a t present with t h e construction and operation of t h e MSRE,

This MSCR evaluation has disclosed c e r t a i n

a d d i t i o n a l areas of study, research, and development important t o t h e r e a l i z a t i o n of t h e nuclear and economic p o t e n t i a l of molten salt reactors. These areas are l i s t e d below, together with s p e c i f i c examples i n each area. 8,7.1

T i t l e 1 Design Study of MSCR

This should be performed f o r two plant c a p a c i t i e s (100 and 1000 Mwe) i n order t o d e t e c t unrecognized development problems and t o v e r i f y t h e economic predictions made i n t h i s report. conducted by an organizatio and cooperation i n spe chemical processing) e

The study preferably should be

s i d e t h e Laboratory, but with close l i a s o n

a1 s t u d i e s (cog., nuclear design, reactor-integrated

study of 100 Mwe i n s t a l l a t i o n is d e s i r a b l e t o

bridge t h e gap between t h e MSRE (10 M w t ) and a f u l l - s c a l e prototype.


174 8.7.2

Conceptual Design Studies'of'Advanced Breeder Reactors P r i o r s t u d i e s have established t h e nuclear p o t e n t i a l of a thermal

molten salt-thorium breeder,

However, t h i s concept should be re-examined

and developed i n g r e a t e r d e t a i l i n t h e l i g h t of t h e technology accumulated i n recent years.

I n addition, t h e p o t e n t i a l of fast molten-salt breeders,

including those breeding plutonium, or possibly both plutonium and 233" should be evaluated.

9

A number of concepts have been proposed, and it is

not clear, a t present, which of these o f f e r s t h e g r e a t e s t ultimate poten-

t i a l coupled with t h e least d i f f i c u l t y of development, 8.7.3

Fundamental Studies of Alternative Chemical Processes While t h e f l u o r i d e v o l a t i l i t y process is s u i t a b l e f o r t h e recovery of

isotopes of uranium f r o m short-cooled f u e l , i t s use alone does e n t a i l t h e discard of t h e carrier s a l t t o r i d t h e system f i s s i o n products. T h i s fact

l i m i t s t h e processing rate and conversion r a t i o i n thermal reactors using valuable isotopes of lithium and beryllium, although fast r e a c t o r s using f l u o r i d e s of sodium and potassium, etc,, are not so limited,

For t h e

thermal r e a c t o r s a t least, an a l t e r n a t i v e process is needed i n which separ a t i o n of valuable components (thorium, uranium, lithium, beryllium) from f i s s i o n product isotopes is e f f e c t e d by t r a n s f e r between f l u i d phases, e.og., by extraction of molten salt f u e l with a l i q u i d metal.

Driving

forces for t h e t r a n s f e r can be provided by t h e use of a c t i v e metals and e a s i l y reduced f l u o r i d e s or perhaps by t h e application of e l e c t r i c potentials.

.. .... .. . Engineering Laboratory .Study_of.P r e c i p i t a t i o n Processing ..

8.7.4

This process, which has great p o t e n t i a l f o r reducing t h e f u e l cycle c o s t i n t h e MSCR and which, by necessity, must be integrated with t h e operation of t h e r e a c t o r and placed within t h e r e a c t o r c e l l , could make t h e introduction of HSCR power r e a c t o r p l a n t s independent of t h e availab i l i t y of c e n t r a l processing f a c i l i t i e s f o r molten salts,

This is an

enormous advantage f o r t h e i n i t i a l i n s t a l l a t i o n s , p a r t i c u l a r l y i f these are widely s c a t t e r e d i n remote locations. Also, t h e competitive position of t h e smaller i n s t a l l a t i o n s (100 Mwe) would be improved,


175 8.7.5

Pilot Plant Studv of'HF Dissolution Process

This process, though not e s s e n t i a l t o the r e a l i z a t i o n of the economic p o t e n t i a l of the MSCR would, if developed, make possible the evolution of the MSCR i n t o a self-sustaining system, and would c o n s t i t u t e a large s t e p i n the development of two-region, two-fluid breederso


177

APPEND1CES

Introduction I n t h i s portion of the report a r e collected reference materials, preliminary studies, and det a i l e d discussions t h a t support the assumptions used or conclusions drawn i n the main body of the report. Literature references i n the Appendices do not r e f e r t o t h e Bibliography which follows, but t o separate l i s t s of references given a t the end of each sub-appendix.

. .


.

,


179 Appendix A MULTIGROUP CROSS SECTIONS FOR MSCR CALCULATIONS C. W. Nestor

Table A . l .

Group Structure ~

Group 1 2 3 4 5 6 7

8 9 10 11

12 13 14 15 16 17

18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 Ep. Thermal 34 Thermal Group

-

au

0.91629 0.69315 0 69315 0 20400 1.09860 1 20400 1 09860 11.20400 1.09860 0 91629 0.98083 0 40547 0 10536 0.11778 0.20764 0.26236 0 10536 0 19574 0 11441 0.09531 0.18232 0 22314 0.16252 0.23052 0.30010 0 28768 0 31015 0 31845 0 47000 0 57982 0 55962 0.28768 0.31705 (1200째F)

U

0.916 1.609 2.302 3.506 4 605 5.808 6.907 8.111 9.210 10.126 11.107 11 512 11 617 11 735 11 942 12.204 12.309 12.505 12.619 12.714 12.896 13 119 13.282 13 572 13.813 14 101 14 411 14 729 15 199 15 779 16 339 16 627 16 944

Energy ( e ~ ) 4 x io6 + 107 2 x lo6 + 4 x 106 1 4 2 x 106 3 x 105 + 106 1 x io5 -+ 3 x io5 3 x io4 -+ 1 x 105 1 x io4 -+ 3 x 104 3 x io3 1 x io4 1 x io3 -+ 3 x io3 400 -+ io3 150 4 400 100 + 150 90 -+ 100 80 3 90 65 -+ 80 50 -+ 65 45 3 50 37 + 4 5 33 + 37 30 -+ 33 25 4 30 20 3 25 17 3 20 13.5 - 1 7 10 13.5 7.5 3 1 0 5.5 3 7 . 5 4 -5.5 2.5 4 4 1.4 -+ 2.5 0.8 j l . 4 0.6 4 0 9 8 0.437 3 0 . 6 0.07940


180 Table A.2.

Lithium-6 1

R

Group 1

2 3 4 5 6 7

fa,

6.16 X 10-1 4.19 X 10-1 4.36 X 10-1 2.02 x 10-1 4.99 x 10-1 4.96 X 10-1 7.38 x 10-1

8

1.11

9 10 11

1.75 2.62 4.04 4.81 5.05 5.32 5.83 6.52 6.85 7.48 7.92 8.28 8.91 9.93 1.07 1.20 1.38 1.57 1.81 2.12 2.69 4.35 5.61 6.39 7.49 7.49

32

13 14 15

16 17 18 19 20 21 22 23 24 25 26 27

28 29 30 31 32 33 34

X 10 x 10 X 10 X 10 x 10 x 10 X 10 x 10 X 10 X 10 x 10 x 10

a

6.74 X 1.618 X 10-1 2.530 X 10-1 7.142 X 10-1 1.463 X 10-' 7.680 X 10-1 1.163 2 057 3.650 6.020 9.689 1.359 X 10 1.542 X 10 1.631 X 10 1.770 X 10 1.991 x 10 2.181 x 10 2.353 X 10 2.542 X 10 2.678 X 10 2.872 x 10 3.179 X 10 3.500 X 10 3.863 X 10 4.413 X 10 5.110 X 10 5.940 X 10 6.950 X 10 8.470 X 10 1.100 x lo2 1.470 X lo2 1.810 X lo2 2.21 x 1 0 2 4.720 X lo2

f'Y 0 0 0 0

0 0 0

0 0 0 0

0 0

0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0

5.04 4.76 3.78 7.59 1.22 x 4.55 5.92 9*87 1.67 X 2.64 X 4.07 X 5.55 x 6.24 X 6.56 X 7.07 X 7.88 X 8.56 X 9.18 X 9.85 X 1.01 x 1.10 x 1.21 x 1.32 X 1.44 X 1.63 X 1.87 X 2.14 X 2.47 X 2.97 X 3.78 X 4.90 X 5.94 x 6.70 X

1.42 x

10

10 10 10 10 10 10 10 10 10 10 10 102

102 102

lo2 lo2 lo2 lo2 lo2 lo2 lo2 lo2 lo2 102

lo2 lo3


181

Table A.3.

Lithium-?

CJ

Group

20 21 22 23 24 25 26 27

4.45 x 4.06 X 3.28 X 3.28 X 7.86 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75, X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X 2.75 X

2%

29 30 31 32 33 34

1 2 3 4 5 6 7 8

i i 1

-

-

9 10 11 12

13 14 15 16 17 18 19

3’tr

a

2.11 x 10-

10-3 10’~

0 0

4.915 3.502 3 451 4.947 6 286 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2 850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850 2.850

2.75 x 10-1

2.55 x 10’~

0

2.850

2.75 2.75 2.75 2.75 2.75 2.75

3.36 3.36 5.12 6.17 7.24

x x x x x

0 0 0 0 0

1.66

X

2.850 2.850 2 850 2.850 2.850 2 850

X X X X X X

10” 10-1 10’’ 10’’ 10’’ 10‘’ 10-1 10’’ 10-1 10” 10-1 10-1 10’’ 10‘’ 10’’ 10’’ 10’’

lo-’ 10’’ lo-’ 10-1 10’’ lo-’ 10-1 lo-’ 10’’ 10‘’

10-1 lo-’ 10-l 10” 10’’ 10’’

3.17 X 5.76 X 7.37 x 1.34 x 2.66 x 4.08 x 7.31 x 1.31 x 2.03 x 3.44 x 4.57 x 5.75 x 5.75 x 5.75 x 7.31 x 7.73 x 7.73 x 8.73 x 1.11 x 1.11 x 1.11 x 1.11 x 1.30 x 1.50 x 2.00 x 2.00 x

lom6 lom6 10-6 10‘~ 10’~ 10-~

10’~ 10’~ 10’~ 10-4 10-4 10’~

10’~

log4 10-4 10-3 10-3

low3

10’~ 10-3

10’~ 10’~ 10’~

0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0

0


i

182

Table A.4.

1 2 3 4 5 6 7

8 9 10

ll 12 13 14 15 16

17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

0 4608 0 3240 0 6813 0.8700 1 1913 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.2122 1.24 1.24

6.3425 9.4485 1.3362 2.1688 3.8471 6.8585 1.2165 2.1688 2.8471 6.3425 1.0204 1.4312

Beryllium-9

X

0

x 10-7

0 0

X X X X X X X X X

lom5

0

x

1.6236

X

1.7167 1.8628 2.0956 2.2960 2.4762 2.6752 2.8190 3.0224 3.3454 3.6832 4.0646 4.6430 5.3775 6.2450 7.3085 8.1920 1.1601 1.5142 1.9013 2.15 x

X X X X X X X X

los4

lou4

x 10'~ X

x x 10-4 x 10-4 X X X X X X

0 0 0 0 0

lom4

5.048 x 10-3

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

5.2490 6 6160 6.6914 1.0649 X 10 1.3588 X 10 1.6068 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6ll2 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.6112 X 10 1.611 X 10 1.611 X 10


183

Table A.5.

Carbon-12

Croup

1

.1 . .-

1

2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

"tr

0 303676 0.2686 0.4108 0 6241 0.694568 0.74181 0.7584 0 7584 0.7584 0 7584 0 7584 0.7584 0.7584 0.7584 0.7584 0.7584 0.7584 0 7584 0 7584 0 7584 0 7584 0.7584 0.7584 0 7584 0 7584 0.7584 0 7584 0 7584 0 7584 0.7584 0 7584 0.7584 7.585 X 10-1 7.585 X 10-1 0

2.1565 X 3 213 4 544 7.380 1.308 X 2.332 X 4.187 X 7.375 x 1.308 X 2 157 3.470 4.866 5 520 5.830 6.335 7 125 7.805 8.420 9.095 9.585 1.028 X 1.138 1.253 1.382 1.579 1.829 2 124 2.485 3.030 3.944 5 245 6 465

4

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0

0

0 0

1.85 x

0

5.048 x

0

3.7354 5 4204 0.0604 9 2352 1.19114 1.28686 1.35370 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.35936 1.359 X 1.359 X

X X X X X X X X X X X X X

X X X X X X X X X X X X

10 10

10 10 10 10

10 10 10

10 10 10 10 10 10 10 10

10 10

10 10

10 10 10 10 X 10 X X

10

10 10 10


184

Table A.6.

Fluorine-19

U

Group 1

2 3 4 5 6 7 8 9 10 11

l2 I3 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

a

1.62 2.00 3.45 3.50 7.70 2.55 4.35 3.70 3.70 3.70 3.70 3.70 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.50 3.59 3.59

X

x x X X

x

x X X X X X X X X X X X X X X X X X X X X X X X X X

10-1 10-1 10-1 10-l10-1 10-1 10-1 10-1

10-1 10-1 10-l10-l10-1 10” 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10-1 10” 10-1 10-1 x 10-1 x 10-1

2.00 x 10-1 0 0

LOO x LOO x 10-~ 3.00 x 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0

0 0 0 0

0

0

0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

1.80 X . 4.5 x

0 0 0

5.40 7.50 9.60 1.35 2.40 1.50 1.35

X X X X

1.u. x

1.11x 1.11x 1.11 x 1.11x

1.05

X

10 10 10 10 10 10 10 10 10 10

1.05 X 10 1.05 X 10

1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05 1.05

X X X X X X X X X X X X X X X X X X X

10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10


!

185

Table 4.7.

15 16 17 18 19 20 21 22 23 24 25 26

5.98 x 10-1 5.89 X 10-1 6.98 x 10-1 1.010 2 375 3.158 3.165 3 175 3.010 2.939 2.930 2.941 2 * 967 2 967 2 967 2 967 2 967 2 967 2.977 3.018 3.018 3.018 3.018 2 995 2.978 3 077

1.868 x 1.904 x 1.936 X 1.972 x 2.006 x 2.031 x 2.057 X 2.096 x 2.264 x 2.544 x 3.531 x 5.778 x 1.159 1.159 1 159 1.159 1.160 1 161 9,800 x 2.009 x 2.016 X 2.027 X 2.043 x 2.092 x 2.291 X 2.652 x

27

3 077

28 29 30 31 32 33 34

3.156 3.270 3.270 3.141 3.152 3.285 3.285

3.554 4.215 5.099 6.740 8.809 1.258 2 874

1 2 3 4 5 6 7

8 9 10 11

12 13 14

j

INOR-8

-

-

0

0 0 0

10.347 10.693 10.521 10.760 13.270 14.48 20.793 42.011 48.153 54.372 59.116 57 442 56 977 57.488 57.58 57 62 58.23 57.30 56.00 57.070 56.140 57 349 58 698 57 628 58.698 58.837

3.038 x 10-1

0

58.930

x 10’1 x 10-1 x 10-l X 10-1 x 10-1

0 0 0 0

59.116 58.279 58.279 57.349 57 74 58 695 58.693

10-1 10-1 10-1 10-1 10-1 10-1 10-l 10-1 10-1 10-l

10-1 10-1

0 0 0 0 0 0 0 0 0 0 0 0 0

0

10-l

10-1 10-1 10-1 10-l 10-l 10-1 10-1

0 0 0 0 0 0 0

0

0 0

0

-

e


186

Table A. 8.

Xenon-135

Group

EST

1 2 3

0 0 0 0 0 0 0

4 5 6 7 8 9 10 11 I2

c

0 0 0

0

13

0

14

0 0 0 0

15

16 17 18 14 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0

3%

'a

7.43 x 10'2 3.908 x 10-1 7.242 X 10-1 9.843 x 10-1 9.546 X 10-1 3.960 X 10-1 1.515 x 10 3.045 x 10 5.133 X 10 e.580 x 10 1.047 X lo2 1.870 X lo2 2.198 X lo2 2.198 X lo2 2.198 x lo2 3.050 X lo2 3.282 X lo2 3.282 X lo2 4.222 X lo2 8.254 X lo2 8.254 X lo2 8.254 X lo2 8.254 x 10' 7.035 X lo2 6.781 X lo2 2.439 X lo3 2.439 X lo3 8.561 X lo3 1.720 X lo4 1.720 X lo4 2.100 x lo4 5.697 X lo4 4.20 x lo4 1.60 X lo6

0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0

0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0

O 0 0 0 0 0

0 0 0 0

0 0 0 0 0 0

J


187

Table A. 9.

Group

Sa,

1

0

3.000

10-1

0

2 3 4 5 6 7

0

3.000 x 10-1

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

3.262 x 3.600 x 1.027 1.570 1.570 2.240 1.641 X 4.857 X 4.857 X 8.829 x 1.911 x 1.911 x 1.911 x 1.756 X 1.714 x 1.714 X 1.586 X 1.037 x 1.037 X 1.037 X 1.037 X 1.463 X 1.887 x 5.991 X 5.991 X 5.991 X

0 0 0 0 0 0 0

8 9 10 11 12 13 14 15 16 17

18

,

Samarium-149

19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

X

10-1 10-1

10 10 10 10 102 102 102

lo2 lo2 lo2 lo2 lo2

lo2 lo2 lo2 lo2 lo2 lo2 lo2 lo2

1.000 x 102 1.000 x 1 0 2

1.391 X lo3 4.774 x 103 7.95 x 102

4.20 x

lo4

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0

0 0 0 0 0 0 0

0 0

0 0 0 0 0 0 0 0 0 0

0 0 0 0 0

0 0 0 0 0 0

0 0 0 0


--

188

w-I

Table A.10.

Samarium-151 --

..-L

a:

Group

rt'3

a

0-26 27 28 29 30 31 32 33 34

0 0

5.50 x lo2

0 0 0 0 0 0 0

0 1.30 x 7.20 X 2.28 x 7.25 x 9.80 x 4.91 x

Table A.11.

0

lo2 lo2

io3

lo2 102

io3

Thorium-232 (Infinite Dilution)*

Group 1 2 3 4

5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25

6.100 5.669 5.841 8.160 1.039 1.119 1.130 1.156 1.168 1.180 1.339 1.824 1.202 1.204 2.301 1.728 1.213 1.217 1.219 1.221 1.224 5.117 1.232 1.236 1.242

X X

x lom2 x loe2 X 10-1 x 10-1 X 10-1 X 10-1 x 1O-I X 10-1 x 10-1 x 10-1 x 10-1 x 10-1 X 10-l X 10-1 x 10-1 x 10-1 x 10-1 x 10-1 X 10-1 x 10-1 x 10-1 x 10-1 x 10-1

0.2408 0.1684 0.1450 0.1622 0.2708 0 4627 0.5680 0.7958 1.029 1.168 7 234 20.802 1.481 1.502 82.755 16 568 1.615 1 646 1 678 1 701 1.731 198.0 1.819 1.864 1.927

5.642 3.190 1.306

X X X

0 0 0 0 0

0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0

10-1 10-1 10-1

1.883 1.998 1.966 2.366 3.132 3.718 3.739 3.739 3.739 3.739 4.658 6.349 3.739 3.739 8.040 6.014 3.739 3.739 3.739 3.739 3.739 1.781 3.739 3.739 3.739

*Groups ll, 12, 15, 16, 22 computed in Appendix B.

x 10 x 10 X 10 x 10 x 10 X 10 x 10 x 10 x 10 x 10 x 10 X 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10

10 x 10 x 10 x 10 X

2


189 Table A. 11 (continued)

Group 26 27 28 29 30 31 32 33 34

Ea, 1.249 1.256 1.263 1.274 1.290 1.305 1.314 1.324 1.324

10-1 10-1 10-1 10-1 10-1 X 10-1 x 10-1 x 10-1 x 10-1 X

x x x x

c

7

0a

"Of

3.739 x 10 3.739 x 10 3.739 x 10 3.739 x 10 3.739 x 10 3.739 x 10 3.739 x l o 4.591 x 10 4.87 X 10

2.000 2 076 2.160 2.269 2.424 2.604 2 745 2.848 3.77

Table A.12.

F'rotactinium-233 c

Group 1

2 3 4 5 6 7 8 9 10 11

12 13 14 15 16 17

$8 19 20 21 22 23 24 25 26 27

U

a

SOT

4.285 4.286 8.569 8.577 8.589 8.613 8.652 8.728 8.851 9.020 9.312 9.480 9.530 9.589 9.702 9.862 9.932 1.007 1.016 1.024 1.040 1.062 1.079 1.106 1.147 1.192 1.248

x x

x X

loe2

x x x x x x X

x IOv2 X

x X

x x X X

loT1

10-1 x 10-1 x 10-1 x 10-1 X 10-1 x 10-1 X 10-1 x 10-1 X 10-1

3'tr

0.002403 0.00358 0.00506 0.00822 0.01457 0.02599 0.4609 0.08218 0 1457 0 2403 0 3866 0 5422 0 6151 0.6502 0 7059 0.7939 0 8697 1.1608 1.0130 1 0680 1.145 1.267 1.396 1.540 1.759 2.037 2 366

VU

f

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0

3'tr 1.496 X 1.496 x 1.496 X 2.991 X 2.991 X 2.991 X 2.991 X 2.991 x 2.991 x 2.991 x 2.991 X 2.991 X 2.991 x 2.991 X 2.991 x 2,991 X 2.991 X 2.991 X 2.991 x 2.991 x 2.991 X 2.991 X 2.991 X 2.991 x 2.991 X 2.991 X 2.991 x

10 10 10 10 10 10 10 10 10 10 10 10

10 10 10 10 10 10 10 10 10 10 10 10 10 10 10


190

G-

t

Table A.12 (continued)

-

ea,

Group

1.316 x 10’” 1.438 X 10’” 1.630 x 10-1 1.885 x 10-1 2.044 x lo-” 2.25 X 2.25 X 10-1

28 29 30 31 32 33 34

5 6 7

8 9 10 11 I2

13 14 15 16 17

18

\

19 20 21 22 23 24 25 26 27 28 29 30

7.400 7.110 6.950 9.320 1.140 1.263 1.371 2.366 1.838 2.427 3.309 4.451 4.324 3.673 5.273 3.253 2.351 6.391 6.420 5.564 7.434 5.650 7.289 1.019 2.284 1.050 6.010 6.350 6.010 8.750

X X X X X X X X

x X X X X X X X X X X

x x X X

x x X X X X X

10-1 10-1 10-1 10”10-1 10-1 10’” 10-1 10-1 10‘” 10-1 10-1 10’” 10-1 10-1 10’1 10-1 10-1 10-1 10-1 10-1 10”10-1 10-1 10-1 10-1

1.940 1.810 1.950 2.160 2.610 3.560 5.864 1.698 1.008 1.411 2.131 3.832 5.031 4.089 4.352 4.614 3.679 3.923 7.249 6.879 6.837 1.525 2.148 1.158 2.059 5.000

X

10

x 10 X X X

x x X X X

x X

X X X X

-

10 10 10 10 10 10 10 10 10 10 10

la

lo2 lo2 lo2 lo2

x x x 10 1.900 x 1 0 2 8.38 x 10 9.26 x 10 5.29 X lo2

%r

Vaf

2 769 3.376 737.0 764.0 7-204 8 374 18.943

Table A.13.

1 2 3 4

--

’Ua

0 0 0 0 0 0 0

2.991 X 10 2.991 X 10 2.991 X 10 2.991 X PO 2.991 X 10 7.486 X 10 1.32 X lo2

Uranium-233

6.570 5.190 5.240 5.430 6.180 8.120 1.319 X 10 3.801 X 10 2.268 X 10 3.175 X 10 4.795 x 10 8.622 X 10 1.132 X lo2 9.201 X 10 9.799 x 10 1.038 X lo2 8.277 X 10 8.828 x 10 1.631 X lo2 1.548 X lo2 1.480 X lo2 3.30 x lo2 4.65 X lo2 2.51 X lo2 4.46 x lo2 1.14 X lo2 3.73 x 1 0 2 1.83 x lo2 1.90 x 1 0 2 1.01 x 103

1.230 X 10 1.442 1.191 1 461 2 271 2 930 2 991 2.390 3.739 3.739 3.739 3.739 3.739 3.739 3.739 3.739 3.739 3 739 3.739 3.739 3.739 3.739 3.739 3.739 3.739 1.80 X lo2 6.00 2.80 3.00 1.50 x lo3

3.39 2.87 2.69 2.51 2.37 2.28 2-25 2.24 2.25 2.25 2.25 2.25 2-25 2.25 2.25 2.25 2.25 2.25 2.25 2.25 2.16 2.16 2.16 2.17 2-17 2 e.28 2.07 2.18 2.05 1.91

3.39 2.87 2.69 2.51 2.50 2.50 2.50 2.50 2.50 2.50 2.50 2.50 2.50 2.50 2.50 2-50 2.50 2.50 2.50 2.50 2.50 2.50 2-50 2.50 2.50 2-50 2-50 2.50 2.50 2.50

--

W


191 T a b l e A.13 (continued)

2 740

31 32 33 34

1-79 X lo2 1.405 x lo2 1.27 x lo2 3.40 X 10

1 430

1.545 1 545

3.990 X lo2 3.217 x lo2 2.9 x 10’ 6.54 x lo2 ~

Group

2 3 4 5 6 7 8 9 10 11 12 13 14 15

‘a

7.298 X 6.865 X 6.718 X 8.020 x 1.007 X 10-1 1.122 1.188 1 167 1.234 1 258 1.044 1.068 1.075 1.083 1.099

1.556 1.529 1-353 0-673 0.083 0.067 0 118 2.104 0.373 0 615 9.054 2.077 X 10 5.957 x 10 5.486 1.833 X 10

16

1.122

2.109

17 18 19 20 21

1.132 1.152 1.165 1.176 1.199 1.229 1.254 1.293 1.350 1.244 1-324 1.420 1.593 1.871

7.901 X 10 2 402 2 595 2.690 X lo2 2 932 3.245 3 573 3 943 4.504 5 216 6.509 X lo2 1.158 X lo3 8 645 1.125 X 10 1.496 X 10 1.844 X 10 2.144 x 10 5.54 x 10

22 23 24 25 26

27 28 29 30 31 32 33 34

2.50 2.50 2.50 2.50

Uranium-234

~-

1

2.23 2.29 2.29 2.29

-~

~~~

T a b l e A.14.

5.40 X I O 2 3.739 x 10 5.40 x lo2 8.47 x io2

2.228 2 454 2.749 2.749

X X

10-1 10-’-

X

10

3 875

~

~~

0 0 0 0 0 0 0 0 0

1.754 X 2.019 X 1.990 x 2.347 X 3.176 X 3.709 X 4.011 x 4.104 x 4.187 X 4.187 X 3.290 x 3.290 x 3.290 X 3.290 X 3.290 X

0

3.290 x 10

0 0 0 0 0 0 0

3.290 X 10 3.290 X 10 3,290 X 10 3.290 X 10 3.290 X 10 3.290 x 10 3.290 X 10 3.290 X 10 3.290 X 10 2.692 X 10 2.692 X 10 2.692 X 10 2.692 X 10 2.692 X 10 2.692 x 10 2.692 X 10 9.104 x 10 2.20 x 102

3.800 3.350 1.630 0.121 ‘ 0

0 0 0 0 0 0 0 0 0 0 0

10 10 10 10 10 10 10 10 10 10 10 10 10 10 10


192

Table A.15.

1 2 3

4 5 6 7 8 9 10 11 12 13

14 15 16 17 18 19

20 21 22 23 24 25 26 27

28 29 30 31 32 33

34

6.62 X 6.11 X 5.71 X 7.73 x 1.00 x 1.14 X 1.23 x 2.03 x 2.46 x 2.38 x 3.44 x 2.86 x 3.82 x 3.24 X 3.44 x 6.79 X 5.94 x 4.04 x 3.82 x 2.12 x 5.40 X 3.99 x 4.42 x 4.03 x 2.80 X 1.81 x 2.39 x 2.29 x 2.97 X 3.27 X 6.45 X 7.89 X 1.12 1.12

loe2 10-2 1 0 ' 1

10-1 10-1 10-1 1O-I 10-1 1 0 ' 1 10-1 10-1

10-1 10-1 10-1 10-1 10-1 10-1

10-1 10" 10-1 10-1 10-1 10-1 1 0 ' 10-1 10-1 10"10-1 10-l 10-1

2.80 2.80 2.80 2.80 2.80 2.80 4 173 7.44 1.36 X 1.97 X 2.66 X 2.78 X 2.78 X 3.76 X 2.85 X 5.12 x 5.40 X 4.52 x 1.58 X 5.97 x 4.91 X 1.01 x 1.47 X 6.19 X 1.28 x 1.14 X 7.02 X 2.95 x 4.21 X 3.39 x 7.20 x 7.17 X 9.22 X 3.05 X

10 10 10 10 10

10 10 10 10 10

lo2

10 10 102

lo2 10 lo2 lo2 10 10 10 10 10 10 10

lo2

---

b -I Uranium-235

3.547 2 195 3 185 3.124 3.852 5-36 8.143 1.14 X 10 1.74 X 10 2.92 X 10 4.44 x 10 5-63 X 10 4.73 x 10 7.08 X 10 6.95 X 10 1.09 x 1 0 2 9.14 X 10 9.42 X 10 2.00 x 1 0 2 1.04 x 1 0 2 8.99 x 10 1.29 X lo2 2.07 X lo2 8.46 X 10 1.31 X lo2 2.38 X lo2 7-60X 10 2.40 X 10 7.53 x 10 4.09 X 10 1.42 X lo2 1.45 X lo2 1.99 x 102 6.29 X lo2

1.23 1.44 1.19 1.46 2.27 2.92 2.99 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.74 3.25 9.62

10 10 x 10 X 10 x 10 X 10 x 10 X X

x x x x x

10 10 10 10 10

x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 X X

lo2 lo2

3.31 2.76 2.50 2.26 2.09 1.92 1.76 1.54 1.28 1.48 1.67 2.02 1.70 1.88 2.43 2.14 1.69 2.08 1.27 1.75 1.83 1.28 1.41 1.37 1.03 2.09 1.08 0.81 1.79 1.21 1.98 2.03 2.16 2.06


I

D

193 &

c$ Table A.16.

Uranium-236

Group

Ea,

O,.

v'f

1 196 9.344 x 10-1 8.993 x 10-1 3.88 x io-* 2.69 x 10-3 4.80 x 10-3 8.50 x 10-3 1.52 x 2.69 x lo'* 4.44 x 10'2 7-14X 1.944 x 10 1.136 x 10-1 8.035 x 10 1.304 x 10-1 1.467 X 10-1 9.507 X 10 7.022 X 10 7.142 X 10 1.405 x 10 7 452 2.342 x 2.578 x 10-1 2.845 x 10-1 3.250 x 10-1 3.764 x 10-1 3.712 x lo2 3.616 X lo2 6.238 x 10-1 8.120 x 10'1 1.079 1.331 1-54 2.10

2 965 2.242 1.592 9.325 x 10 0

3'tr f

1 2 3

4 5 6

7 8 9 10 11

12

13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

6.759 6.167 5.992 7.831 9.924 1.106 1.166 1.136 1.186 1.188 9.371 9.388 9.394 9.400 9.411 9.427 9.434 9.449 9.458 9.466 9.482 9.504 9.522 9.550 9.591 7.946

8.004 8.073 8.196 8.395

7.710

X X

x x x x 10-1 x 10-1 x 10-1 x 10-1 x 10-1 x x x 1 0 ' 2 x x x x x 10'2 x x x x x x x x 10'2 x x lo'* x lom2 x x io-2

7.726 X 9.85 x 9.85 x

0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

1.754

X X

2.019 1.990 x 2.347 X 3.176 X 3.709 X 4.011 X

4.104 4.187 4.187

X X X X

10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10

3.290 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 x 3.290 X 3.290 x 3.290 x 2.692 x 2.692 X 10 2.692 X 10 2.692 x 10 2.692 X 10 2.692 x 10 2.692 X 10 3.45 x 10 3.05 x 10


194 Table A.17.

Group

1 2 3 4 5 6-28 29 30 31 32 33 34

w-

Neptunium-237

-

Ea,

(T

0 0 0 0

-

1.970 1.450 1.340 0.382 0.025 0

0 0 0 0 0 0 0 0

%tr

*f

a

8.52 X 10 2.08 x lo2 1.72 X lo2 3.74 x 102 6.35 X lo2 1.07 X lo2

5.110 3 780 3.480 0.994 0.065

6.000 4.350 4.020 1.100 0.075

0

0 2.85 X 6.50 X 5.40 x 1.50 X

:o 0 0 0 0 0

lo2 lo2 lo2 lo2

0 0

Table A. 18. Uranium-238

-CJ

Group 1

2 3 4 5 6 7

8 9 10 11

I2 13 14 15 16 17 18 19 20 21 22 23

24 25

%I-

a

6.37 X 5.95 x 5.78 X 7.88 x 1.01 x 1.13 X 1.21 x 1.22 x 1.27 X 1.28 x 1.03 X 1.03 x 1.03 X 1.03 X 1.03 X 1.03 x 1.03 X 1.03 x 1.03 x 1.03 x 1.03 X 1.03 X 1.03 X 1.03 X 1.03 X

10’2

10-1 10” 10’’ 10” 10” 10’’ 10” 10-1 10” 10’’ 10-1 10-1 10” 10’’ 10’’ 10’’ 10” 10” 10” 10” 10”

7.74 x 6.04 X 3.48 X 1.34 X 1.98 x 3.37 x 5.23 X 7.79 x 1.19 2.30 8.51 2.61 X 1.44 1.16 X 4.92 X 1.26 1.26 1.26 3.85 x 1.26 1.27 2.76 X 1.27 1.28 1.46

10” 10” 10” 10-1 10‘’ 10” 10” 10”

10 10 10

lo2 lo2

1.92 1.45 6.64

X

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

1o-I

1.75 X 10 2.02 x 10 1.99 x 10 2.35 X 10 3.18 X 10 3.71 X 10 4.01 X 10 4.10 X 10 4.19 X 10 4.19 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 x 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10 3.29 X 10

6

--

a


195

Table A. 18 (continued) -

-a

~-

Sa,

Group 26 27 28 29

30 31 32

33 34

8.62 8.62 8.63 8,63 8.64 8.65 8.65 7.98 7.98

X X X X

lom2 x 10" X lom2

X

x 10-2 x 10'2

?JE f

a

1.28

4.14

X

lo2

1.29 1.30

1.31 1.32 1.32 6.01 X 10" 1.36

3Ftr 2.69 2.69 2.69 2.69 2.69 2.69 2.69 2.85 3.47

10 10 10 10 10 10 10 10 x 10

X X X X X X X X


196

Appendix B

EFFECTIVE THORIUM RESONANCE 1NTEG.RAT.S J. W. Miller"

I n t r o d w tion The "effective" resonance i n t e g r a l t o lethargy um of a particular isotope i n a mixture of isotopes i s defined by the equation

The value of (uao)eff i s a complex function of a number of variables. A most important item i s the scattering power of the medium, measured by u macroscopic scattering cross section per t h o r i m atom, P'

u = C N ak/Nth p t L k s

The greater the scattering power, the greater the probability t h a t a neutron w i l l be slowed through a resonance without absorption. The thermal

j

, I

i

i !

!

motion of absorber nuclei, on the other hand, increases t h e probability of absorption due t o Doppler broadening of the resonances. This e f f e c t i s small a t 2OoC, the standard temperature for reporting measured values, but increases at higher temperatures. Dresner has developed a method f o r computing the temperature e f f e c t f o r resolved resonances.' Group averaged cross sections f o r thorium a r e r e l a t e d t o the effective resonance i n t e g r a l and were calculated by means of the equation

i

where i i s the group number. ~

~~~

~

*Adapted from ORNL-CF 61-1-26.

!

ii


197

Analysis By assuming a Maxwell-Boltzmann speed distribution for the moderating medium and the Breit-Wigner formula for a single resonance, Dresner2 has arrived at the following equation for the resonance integral, Ierfrfor a single absorption resonance:

where

k = In f3

+ 5 In 10 In 2

Other terms are defined in the nomenclature. sult The function J ( E , k ) is tabula%ed in ref. 1. The tabulated have been plotted as the family of curves in Fig. B.l. The reduced mass of the neutron p is equal to neutron mass (-1.0)

multiplied by M/(M+l), where M is the mass of the absorbing atoms. Since the mass of thorium is 232, p was taken as unity for these calculations. Table B.l lists the resonance parameters for--eachof the 13 resolved thorium resonances (2). These parameters were used in the equation for Ieff for computing the resonance integral for each resonance. (Since these calculations were performed, additional thorium resonances have been resolved, and improved parameter values for previously resolved resonances have been obtained. The effect of these new values has not been evaluatdd.) The total resonance integral for each energy group is then obtained by adding together the separate integrals for each resonance. Table B.l also


198

ORNL-LR-DWG.

IO2

84604

~

8 6

4

2

6

4

2

4 0

0.I

0.2

0.3

0.4

0.5

E Fig. B . l .

J(k,k) Versus E for k = 6.16.

0.6

0.7


199

L c,

I

Table B.l.

Resonance Parameters of Thorium-232

,

GNU Group No.

11

AU

0.98083

Infinite Dilution Resonance Integral

Eo(ev)

rn(ev)

ry(ev)

235 234 212 201

195 172

0.0017 0.027 0.0012 0.016 0.022 0.067

0 034 0.034 0 034 0 034 0.034

130 122 114

0.009 0.023 0.013

0 034 0 034 0.034

8.4

7.10

0.034

12

0.40547

15

0.20763

69.7

0.039

0.043

17.2

16

0 26236

59.6

0.0046

0.021

4.4

22

0.22314

23.6 21.9

0.004

0.039 0.034

4.4.2

, ,

-

-

0.0022

Unresolved Resonances Group 1 through Group 10 Nonresonance Contribution Groups 13, 14, 17-21, and 2 3 3 3 Total- Cut-off at 0.0795 ev

-

5.3 10.3 96.9

lists the lethargy width f o r each goup. The group-averaged cross section is simply the group total resonance integral divided by the group lethargy width The infinite-dilutionresonance integral (Irn) f o r a particular

resonance may be obtained from


. 200

Sample Calculation Example: Compute the effective resonance integral (Ieff)for the thorium resonance at 23.6 ev with a carbon-to-thorium ratio of 200 and an absorber temperature of 649째C. The nonresonance scatkering cross section, uS of carbon is 4.8 barns, and the thorium nonresonance scattering cross

',

section, u m, is 12.5 barns. S

1.

USTh+(k)(JP

972.5 barns 2

2.

0.877 X

cm2

3.

10,250 barns

4.

2 uO

0.0949

.

0.1797

.

5.

6.

r -

A

0.2393

.


201

7.

8-

k =

In 8

+ 5 I n 10 In 2

J(5,k) = J(0.2393, 13.2) E

7.6 (from Fig. B . 1 )

= 12.2 barns

.

.

The inf i n i t e - d i l u t i o n resonance i n t e g r a l (I,) obtained i s 26.6 barns f o r the case considered.

The MERC-1 program used i n the MSCR study i s a 34-group diffusiontheory code.

The group structure i s such t h a t the resolved resonances of

thorium f a l l i n t o f i v e grbups:

11, 12, 15, 16, and 22.

The group-averaged

absorption cross section for each of these f i v e groups as a function of u

a t 649OC I s plotted i n Fig. 8-2.

P The t o t a l resolved resonance i n t e g r a l i s

plotted i n Fig. B-3 for three reactor temperatures. Symbols r'

Y

= Radiative capture width (ev)

I'n = Neutron width (ev) F = ,Total width (ev)

3c=

Wavelength of the neutron 2

Eo = Resonance energy (ev)

(d


202

ORNL-LR-DWG.

io2

2

4

6

8 io3

2

64606

4

up (barns)

Fig. B . 2 .

Group Averaged Absorption Cross Section Versus up 649째C.


c

C

a 6 W 0

I I I I I I

I

CURVE

T E M P E R A T U R E , "C

A B

925 649 280

4

Z h

I

C

-

a: Zb

>W

-I+ 02

wW [r:

10

a 6

~

6

' 402

4

2

cr

P

(barns)

io3

Fig. B . 3 . Total Resolved Resonance Integral Versus op for Moderator Temperatures of 280째C, 649"C, and 925OC.


204 u

= Nonresonance scattering cross section (macroscopic scattering cross

section per thorium atom density) E = Relative energy in the center-of-mass-system M = Mass of absorbing nucleus (mu) KT = Absorber temperature in energy units (ev) p = Reduced mass of neutrons (mu)

References 1.

Lawrence Dresner, Tables for Computing Effective Resonance Integrals Including Doppler Broadening of Nucleas Resonances, USAEC Report ORNL-CF 55-9-74, O a k Ridge National Laboratory, September 19, 1955.

2. D* J. Hughes and R. B. Schwartz, Neutron Cross Sections, USAEC Report BNL-325, Brookhaven National Laboratory, July 1, 1958.


205 Appendix C ENERGY DEPENDENCE OF ETA OF 233U*

C. W. Nestor

A parameter of great i n t e r e s t i n nuclear calculations of a thorium

reactor i s q, the number of neutrons produced per neutron absorbed.

Ex-

perimental information on the energy dependence of q i n the range of 0 t o 10 ev as measured a t the MTR was used t o calculate group averaged f i s sion cross sections, using absorption cross sections calculated from the recent t o t a l cross section data1,* and the s c a t t e r i n g cross section as calculated by Vogt.â&#x20AC;&#x2122;

The q values used i n preparation of the cross sec-

tions was normalized t o a 2200 m/sec value of 2.29.3 I n the energy range of 0 t o 0.8 ev, q was assumed t o be constant a t

I n the range of 0.8 ev t o 10 ev, as mentioned, group averaged

2.29.

values of

Ef= qua

were calculated by numerical evaluation of the i n t e -

grals

i

where Au denotes the lethargy width of the group*

I n the range of 10 ev t o 30 ev q was estimated t o be 2.17; the d a t a of Gaerttner and Yeater4 indicate an average q i n t h i s range of about 0.95 times the 2200 m/sec value.

*Adapted from ORNL-CF 61-6-87 (Rev). !

*


206 From 30 ev t o 30 kev, q w a s assumed t o be 2.25; t h i s i s the value reported by Spiv&â&#x20AC;&#x2122;

e t a l . , a t 30 kev.

From 20 kev t o 900 kev, measurements of q a r e a ~ a i l a b l e ;f~i s s i o n cross sections a r e reported i n BNL-325 f o r the range 30 kev t o 10 MeV. The t o t a l cross section i n t h i s range w a s taken t o be equal t o t h a t of 235U, as suggested by J. A . Harvey.l The value of V was assumed t o be l i n e a r i n energy with a 2200 m/sec value of 2.50 ( r e f . 3 ) and a slope of 0.127 per A p l o t of the experimental q and t h e group averaged values from 0.01 ev t o l k e v i s shown i n Fig. C.l. l i s t e d i n Table A . l of Appendix A.

Group values are a l s o


?

i

207

1

;

,

,

*

-I

.


208 References

1.

J. A. Harvey, ORNL, personal communication t o C. W. Nestor, ORNL, March 1960.

2- M. S. Moore, MTB Nuclear physics Group, personal communication t o C. W. Nestor, ORNL, March 1960.

3.

J. E. Evans, reported a t t h e Argonne National Laboratory Conference on the physics of Breeding, October 1959.

4.

E. R. Gaerttner and M. L. Yeater, Reports t o t h e AEC Nuclear Cross Section Advisory Group, USAEC Report Wash-194, USAEC, February 1958.

5.

P. E. Spivak e t a l . , Measurement of Eta f o r 233U, 235U, and 239Pu with Epithermal Neutrons, J. Nucl. Energy, 4:70 (January 1957).

6 - G. N. Smirenkin e t al., J. Nucl. Energy, 9:155 (1955).

.


209

Appendix D THE MERC-1 EQUILIBRIUM RUCTOR CODE

T. W. Kerlin Introduction The M E N - 1 code automatically calculates the composition of a fluidfuel reactor so that equilibrium and criticality conditions are simultaneously satisfied. MERC-1 uses the MODRIC' multigroup-diffusion-theory code and the ERC-10 equilibrium reactor code as chain links. These codes were previously used separately at ORNL in iterative calculations to determine equilibrium reactor compositions. MODEIC was developed by the Central Data Processing Group at the O a k Ridge Gaseous Diffusion Plant as a replacement for GNU.2 The ERC-10 code was prepared as a later version of ERC-!jO3 The MERC-1 code automatically transmits necessary data from one chain link to the other until sufficient iterations have been performed to cause criticality and equilibrium requirements to be satisfied simultaneously. The output consists of equilibrium concentrations, a neutron balance, and fuel-cycle costs. The system which is considered in the MERC-1 analysis is shown in Fig. D.1. A complete specification of data required for controlling the f l o w and losses in each stream is included in the MERC-1 input. Note in Fig. D.1that material is removed from the reactor system by losses, waste, and sale as well as by nuclear transformation (decay and neutron absorption), and that fresh material is fed to the system. Therefore, the calculated equilibrium concentrations are in equilibrium with respect to the

fee4 and discharge rates as well as with respect to nuclear transformation rates. "heory MODRIC is a typical neutron-diffusion-theorycode. It allows 50 neutron energy groups with downscattering from a group to any of the


210

-

OR NL- LR DWG 768 24

Losses

Waste Sales

Recycle I-

2

IC. Loss

4Processing 1Plant

-&-

Waste

=8- Sales

*The reactor may have os many as two active regions, ondeach region may contain either or &oth of the fluid streams Fig. D.l.

The Reactor System.


211 following ten groups. It will perform concentration searches on specified elements. The output consists of criticality search converged concentrations, group macroscopic cross sections, normalized nuclear events (absorptions, fission,leakage, etc. ) by region and group, absorptions and fissions by material and region, group f l u distributions, and fission density distributions. ERC-IO requires extensive input. Rewriting of large quantities of input is avoided by using basic input decks which include lnformation ap. plicable to a number of cases. To specify a new case, it is necessary only to specify changes in the basic input deck. For instance, one might prepare a basic input deck for a particular reactor with a given power level. A set of cases with different power levels would need only the basic input deck and the new power level as input. Basically, ERC solves two equations. They are: dN..

f

V.

= Qij

df

+ Rij

+ Fij + Tij

+

Dij

or if the material must be fed to maintain criticality

Equation (2) is just the conservation requirement, saying that enough fissile material must be added (or removed) h iteration 's to Overcame the neutron production deficiency (or excess) in iteration (s-1). These are inner iterations in ERC. The terms are defined as:

vj = volume of Nij

U

t

stream j , cm3,

=,atomsof material i per barn cm of stream j , =

time, sec,


212 = feed rate of material i into stream j, atms/sec, Qij Rij = rate of growth of material i in stream j due to recycle from other streams, atoms/sec , Fij = rate of growth of fission fragment i in stream j, atoms/sec, = rate of growth of material i in stream j due to neutron absorptions in other materials, atoms/sec, Dij = rate of growth of material i in stream j due to radioactive decay of other materials, atoms/sec, = rate coefficient for loss of material i in stream j because of neutron capture, atoms per sec per atom/parn cm, = rate coefficient for loss of material i in stream j because diJ of radioactive decay, atoms per sec per atm/barnâ&#x20AC;&#x2122;cm,

= rate coefficient for loss of material i in stream j because qij of processing removal, atoms per sec per atm/barn cm,

r

ij

= rate coefficient for growth of material i in stream j because

of recycle f r m stream j, atoms per sec per atom/ba;rn cm, â&#x20AC;&#x153;i = neutrons produced per fission in material i, .P A

Cijk

a Cijk

=

reaction rate coefficient, number of fissions in material i per atom/barn cm in stream j in region k per fission neutron born in reactor,

=

reaction rate coefficient, number of absorptions in material i per atm/barn cm in stream j in region k per fission neutron born in reactor.

Superscripts: i = material, j = stream, k = region.

3

The use of stream and region indexes allows reactors with two streatns in the same region to be analyzed. < The equilibrium concentration calculdtions in ERC use reaction rate coefficients (Cijk) obtained frum an earlier MQDRIC calculation. However, the initial concentrations used in the MODRIC calculation will not, in general, agree with the equilibrium concentrations computed by ERC. This new set of concentrations will alter the neutron spectrum and flux level, thereby changing the reactioi rate coefficients. Therefore, it is necess a y to repeat the MOD3IC criticality calculation with the latest value for the estimated concentrations to get new reaction rate coefficients-

\


213

This process is repeated until the MODRIC and ERC concentrations are equal. The flow of information in the code is shown in Fig. D.2. The reaction rate coefficients (Cijk) used in ERC are spectrum-averaged cross sections which are available directly from MODRIC. The MODRIC calculation gives Aik and VFik, the absorptions and neutron productions in material i in region k, normalized to 1.0 total neutron produced. The distribution of nuclear events between two streams in a region is accomplished by introducing the stream volume fractions, fjk, in this manner: atoms of i in stream j in region k a. Aijk = Aik*( atoms of 3. in region k yiFijk = YiFik

, atoms of i in stream j in region k atoms of i in region k

f

The multiplying factor in each term is atoms of i in stream j in region k atoms of i in region k

, j

where the units on these factors are Nij

-- atoms of i in stream

barn cm of stream j

j

'

cm3 of stream j in region k

f = jk

cm3 of region k

The material, stream, and region dependent absorption and production terms are automatically transfe ed from the MODRIC link .t;o the ERC link of the MERC-1 calculatian. ERC obtains the reaction rate coefficients (intensive quantities) from the absorption and production terms (extensive quantities) by dividing by Nij, the stream concentration ~


t

214

OR N L

- LR- DWG 7682 5

First Guess at Concent raf ions

1 - 1 I

I I

I

------

I I -..------------eo--

1 -i I I I I I I

-------- J

MODRIC and ERC agree on all concentrations

~

ffoM Nohs t

Is maximum number of iterations exceeded

Output

t Output Fig. D.2.

The MERC Flow Diagram.


215

Vi Cijk

-

V F i ijk

Nij The absolute reaction rate coefficient

Eijk

is obtained in the ERC cal-

culation using the total neutron production rate as determined by the reactor power -a 'ijk

-- 'iajk

X

3.1

X

10l6 P T X lo-",

where 3.1

X

10l6 = number of fissions per sec per megawatt, P = power level in megawatts,

-V = average number of neutrons produced per fission Nij Cijk f Vi

-- ijk E Ni j Cfijk ijk A similar argument applies for the fission reaction rate.

The restriction to fluid-fuel reactors occurs because the ERC calculation requires that the reactor discharge composition be equal to the mean reactor canposition. This restriction i s satisfied in fluid-fuel reactors but not in solid-fuel reactors where the discharge has experienced a much greater exposure than the mean. In ERC-10, effective, one-group cross sections of individual fission products were calculated by reference to a standard absorber exposed to the MSCR spectrum of neutrons as generated by the multigroup program Modric. The thermal cross section and the resonance integral can be used in a twogroup model to calculate relative one-group cross sections. Thus, set

where the bar denotes effective, one-group values, "e" denotes "eipthermal," and "t" denotes "thermal." Rearranging and taking a ratio of the cross


216

section of the ith material to that of a standard material, denoted by the subscript "s,"

.

The epithermal cross section is defined in terms of the resonance integral by the relation

= -RI

ue =

ue

.

where ue is the lethargy at the lower energy bound of the epithermal group. The ratio

@e is obtained by equating the slowing down current from the

@t epithermal group to neutrons absorbed or leaking in the thermal group.

where C is the "removal" cross section. r,e Let C r 6 CT,JUe r,e where T denotes "total."

Ignoring leakage gives

Combining these results gives

- .

[ut,e

u i = u s [ut,s + K(RI)s] ,


217 where

The spectral index, K, is computed by Modric from input data.

The reference

element was the standard absorber referred to above. The product of the reference element cross section and the effective flux integrated Over the core is computed from Modric output by dividing the fraction of neutrons as absorbed by the reference element by its atomic density, Ns.

This

product is the desired number for use in the ERC calculations; the working equation becomes

The thermal cross sections u

t are computed from the 2200 m/s cross sections by multiplying by a factor that averages them Over a MaxwellBoltzman spectrum around the reactor temperature. This factor was canputed for a l/v enera dependence of the kross section and applied uniformly to all fission product isotopes, except noble gases, In all, 115 fission product isotopes were so treated, linked by transmutation and decay into chains. Provision was made in ERC-11 to remove each at a rate determined by its chemical or physical properties in relation to the processing method. For instance, xenon is removed rapidly by transpiration in an expansion chamber, whereas rare earths are removed only by discard of the fuel salt with a period measured in hundreds of days. Xenon was treated separately in the Modric calculation, not only to determine accurately its effective cross section in the MSCR neutron spectrum but also to permit special treatment of its exceptional behavior. It may be possible to remove xenon rapidly from the fuel solution by circulating a portion of the salt through the dome of the expansion tank mounted over the core (Sec. 4.2.2) provided the xenon does not diffuse rapidly into the moderator graphite.

Provision was made in ERC-11 to calculate

the extreme cases (complete absorption in graphite vs. zero absorption) as well as intermediate situations where remmal competes with absorption.


218

,-

wSamarium was also treated separately because of its importance. The fission product calculation is thought to result in a reasonably good approximation of the poisoning in reactors where the fission products are exposed to neutrons for a lo?zg time. The ingrowth of second and higher generation isotopes by transmutation and decay is treated in detail. The transient period following start-up of a clean reactor with an initial loading of 235U is ignored, and all concentrations are calculated at their maximm, equilibrium values. Hence, the poisoning is overestimated - somewhat, thus providing a margin of safety in respect to assignment of the cross sections and resonance integrals. The fission-product reaction rate coefficient is obtained by reference to a specified standard absorber:

where CFp = fission-product reaction rate coefficient,

CR

= reference material reaction rate coefficient,

oFP = effective fission-product absorption cross Fection, crR = effective reference material absorption cross section.

The effective cross section ratio is obtained from a two-group formulation:

FP

(J

-=

R

(J

((Jl g1/g2

(B1 g1/g2

+

u2

R +

(321

where u1

= fast absorption cross section

U

th

U

th

â&#x20AC;&#x2122;


219

-U 2 f12

=

ak;sor>tion cross section averaged Over the thermal flux,

=

fast flux,

=

average thermal flux.

For a two-group treatment, all neutrons removed from the fast group must

either be absorbed or leak from the reactor while thermal: R'

1

$1

=

ca

G~

+ D B ~d2

2

Ignoring leakage,

Also,

where f = thermal spectrum factor

Boltzmann distribution, 02200 =

a

K is calculated as follows:

c

WS,

29' 273 for a -ell+

(5), and (6) into Eq. ( 3 ) gives

where

=

q l ~ ) ~)

2200 m/s absorption cross section.

Substituting Eqs. (CC),

K

=


220

w = -f- a u S =

input t o linkage section of input,

ca &,

= value automatically calculated by MODRIC.

2

The value of a m u s t be specified i f f i s s i o n product option 1 i s speci: iel on card BN=5. f o r K.

This value should be calculated using an estimated value

a i s calThe required values of

If f i s s i o n product option 2 i s specified on card BN-5,

culated by the code using the l a t e s t value of K.

and R I f o r a special reference material are b u i l t i n t o the code f o r a f i s s i o n product option 2. The nuclear constants f o r a l/v absorber with a 2200 m/s cross section of 1.0 barns are b u i l t i n t o the code. Therefore,

cr2200

t o use f i s s i o n product option 2, the reference element must correspond t o an a r t i f i c i a l element i n MODRIC which has cross sections f o r a l / v absorber with

g;*Oo

= 1.0.

R e f ere nce s 1. J. Replogle, MODRIC

- A One-Dimensional Neutron Diffusion Code,

USAM:

Report K-1520, Oak Ridge Gaseous Diffusion Plant, September 1962. 2.

C. L. Davis, J. M. Bookston, and B. E. Smith, GNU-11, A Multigroup One-Dimensional Diffusion Program f o r the IBM-704, GMR-101, General Motors, November 12, 1957.

3.

L. G. Alexander, ERC-5 Program f o r Computing the Equilibrium States of Two-Region Thorium Breeder Reactors, USAEZ Report ORNL CF-60-10-87, Oak Ridge National Laboratory, October 1960.


221 Appendix E

FISSION mODUCT NUCLEAR DATAa

L.

G.

Alexander

Fission Yieldb Number Isotope 233u

26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48

&

m

2 3 5 ~

0.007 0.0045

0.0028 0.0014

0.012 0.019

0.00544

0.032 0.019 0.040

0.010 0.00293 0.0202 0.010 0 249

0.050 0.065 0.065 0.065 0.067 0.070 0.068 0.057

0.0357 0.0479 0.0577 0.0584 0.0603 0.0645 0.0640 0.0633

0.062

0.0627

0.006

B, Decay h, sec-l

0.214 X

0.298 X lom5

Cross Section go, barns

2.1 3.3 45 205 0.16 7 0.06 0.91 0.13 1.3 0.0055 1.31 1 1.2 0.15 1.1 0.076 0.053

Resonance Integral RI, barns

1.4 60 45 201 5.5 29 0.04 0.67 O.2lc (0.6) 0.06 0.78 1.8 9 0.55 28 0.2 0.07

0.229 x 13.9 1.2 2.2

0.053

0.0609

49

0.052

0.0578

0.51

50 51 52 53 54 55 56 57 58 59 60 61 62

0.044 0.048

0.0630 0.0606

0.030 0.024 0.0097 0.016

0.050 0.041 0.018 0.030

0.005 0.0028 0.0015 0.0006 0.0003

0.009 0.0038 0.0019 0.0007 0.00024

0.3 22.2 1.7 5 1.44 0.7 184 6 11 6 10 10.7 0.28

109 34 16 5.6

6.2 140 7 77 11 8 1030 19 76 I2 40 169 10

aCompiled from Appendix E list of references..

ad

bAtoms per fission. C Values in parentheses estimated by comparison with similar nuclides.


222 Fission Product Nuclear Data (continued) Fission Yield Number Isotope 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84

85 86 87

88 89

90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107

,,'ol

1llCd 112Cd l13Cd l14Cd l151n 116Sn 125Te 1.26Te 128Te 127~

lZg1

128xe l2'Xe l3 Oxe 131xe l3 2 ~ e 133Xe 134Xe 135~e l3 6 ~ e 133Cs 134cs 135cs 136cs 137cs 138cs 134Ba 136Ba 137Ba 138Ba 139La 140ce 14%e 142ce 143~e 141142-

23321

0.0004 0.00025 0.0002 0.0002 0.0002 0.0002

23 5u

0.0003 0.00019 o.ooo1 0.0001 0.0001 0.0001

0.0024 0.010 0.0039 0.02

0.0005 0.0037 0.0013 0.009

0 037 0.051

0.0293 0.0438

0 066d 0 067 0.069 0.062

0 0806d 0.0641 0 0646 0.0659

0.067e

0.0641

0.072

0.0615

0.068

0.061

0.0574 0-0655 0.0644

0 057

0.0595

0.059

0.064

0.052 0.041 0.030 0.023

0 0598 0 0567 0.0395 0.0307

0.064

143Pr 144-

143Nd 144Nd la5Nd 146Nd 147Nd

-

dIncludes indirect yield from e Included in Xe yields.

13Q.

Decay A, sec'l

Cross Section ga, barns 91 2 1 59,500 1.2

c

228

Resonance Integral RI, barns u20 52 13 652 15 3300

0.006 1.56 (0.8) 0.8 12 0.3 2 6.2 154 27 39 5 45 45 302 5 45 I20 806 0.2 1.8 190 I270 0.152 X 0.2 0.6 0.211 x 10'~ 3.344 X 10 0.6512 X 10 0.15 0.1 0.296 X 420 28 1400 0.110 x 10-7 137 62.0 8.7 0.617 X 0.11 0.3 0.666 X ''ol 62 0.362 X 8-7 2 (1) 0.4 (0.2 1 4.9 (2.5) 0.68 0.3 11.0 8.9 0.5 0.66 0.251 X 0.94 1.3 0.601 X 6 (3) 11.5 23.5 18.0 (9.0) (0.45 x 10) 89 0.660 X 308 130 12 5 67 245 10 25 180 2510 0.710 X


223 continued )

Fission Product Nuclear Data ~

-

~~

~

~~~~~~~

Fission Yield Number

Isotope

108 109 110 111 112 113 114 115 116 117 118

148Nd l4 'Nd lS0Nd lS1Nd

0.012

147h 148-

0.017

0.0048

0.0067 0.0238

0.0062

0.0113

0.0026 0.0017

0.0045 0.00285

0.00037 ~ 0.00095 ~

0.00077 0.0015

0.700 X 0.846 X 0.151 X

0.301 X

6 E ~

0.00015 0.0003 0.00005 0.00013 0.5 x 10-~0.78 x 10-4 ~ 0.1x 10-4 0.2 x 10-4 0.148 X

(f)

l 35Ba

(f1

160Tb

TOTAL 'Yield

690 50 2440 460 3565 2500

1500

8490

25 1380 750 4245

1630

44 740 29

0 027

0.020

2.02285

2.06485

46 130 1.5

420 (65) 0.7

5.6 8400 5500

(31 (4200) (2750) (262) 2.6

0.289 X

1 5 1 ~ ~ 1 5 2 ~ ~

13'Te

10,260 194

2510

0.107 X

1351

I39 Ut0

180 27,000 87

58,000 4 0.24 X lo6 3.9

0.5 x 10-5 0.1x 10-4

%r "~r

138

14

5 382

1 5 5 ~ 1 5 6 ~ 1 5 7 ~

l5 'Tb

1.5

X

0.137 X 0.129 x 10'~ 0.521 X

1 5 4 ~

1 5 5~

48

87,770

0.410

1 5 4 ~ ~ 1 5 5m

1 5 8

Resonance Integral RI, barns

3.4

85

15%m l5 2 ~ m 153~m 1 5 3

0.0170

9

Osm

154~m

235u

Cross Section gu, barns

0.963 X

1lc7sm lc8sm l4 '~m

119 120 I21

I22 I23 124 125 I26 127 128 129 130 131 132 133 134 135 136 137

2 3 3 ~

Decay h, sec-l

of 1351 combined with that of 135Xe.

525 0.5


224

References

1. J. 0. Blomeke and M. F. Todd, Uranium-235 Fission-Product Production A s a Function of t h e T h e m 1 Neutron Flux, I r r a d i a t i o n Time, and Decay Time. 11: Summations of Individual Chains, Elements, and t h e Rare-Gas and Rare-Earth Groups, USAEC Report ORNL-2127, Pt. 11, Vols. 1, 2, and 3, Oak Ridge National Laboratory, November 1958. 2.

J. D. Garrison and B. W . Roos, F i s s i o n Product Capture Cross Sections, Nucl. S c i . Eng., 12: 115 (January 1962).

3. E. A. Ne hew, Thermal and Resonance Absorption Cross Section of t h e

233U 9 23%, and 23%u F i s s i o n Products, USAEC Report OW-2869, Oak Ridge National Laboratory, March 1960.

4.

J. J. Pattenden, F i s s i o n Product Poisoning Data, USAEC Report ORNL 2778, Oak Ridge National Laboratory, October 1959.

5. C . R. Greenhow and E. C. Hensen, Thermal and Resonance Fission-Product

Poisoning f o r 23% System, USAEC Report W L - 2 1 7 2 , Knolls Atomic Power Laboratory, October 1961.

6. N. F. Wikner and S . Jaye, Energy-Dependent and Spectrum-Averaged Thermal Cross Sections f o r the Heavy Elements and F i s s i o n Products f o r Various Temperatures and C: "'U Atom Ratios, USAEC Report GA-2113, General Atomic, June 1961.

7.

R. L. Ferguson and G. D. O'Kelley, A Survey and Evaluation of 233U F i s s i o n Yield Data, WAEC R e p o r t ORNL (37-62-3-71, Oak Ridge National Laboratory, March 1962.

8. E . C . Hensen, A C r i t i c a l E W n a t i o n of t h e Uncertainties i n Predicted Gross F i s s i o n Product Poisoning, USAEC Report KAPL-M-ECH-8, Atomic Power Laboratory, March 1962.

Knolls

9. E . C. Hensen and C. R. Greenhow, An Improved Generalized Analysis of

F i s s i o n Product Poisoning and Thermal and Resonance F i s s i o n Fragment Cross Sections, USAEC Report KAPL-M-ECH-7, Knolls Atomic Power Laboratory, September 1960.

10. C . H. Wescott, Effective Cross Section Values f o r Well-Moderated Thermal Reactor Spectra, Canadian Report ~ ~ - 7 (Rev. 8 7 August 1958)s AECL-~~O. 11. J. 0. Blomeke, Nuclear Properties of 235U F i s s i o n Products, USAEC Report OW-1783, Oak Ridge National Laboratory, April 1957.


225

Appendix F TREATMENT OF DELAYED NEUTRONS T. W. Kerlin Summarv

Circulating fuel reactors lose neutrons because some of the delayed neutrons are emitted outside of the core. These losses depend on core residence time, external loop residence time, and decay characteristics of the precursors. A symbolic representation of the system is

+-4-,

Exchanger

where i' j = decay constant o

he ith precursor from fissionable material

j,

= number of ith precursors formed per fission neutron from fis-

sionable material j,

"j

E atms of ith precursor per unit volume of fuel stream in the N i j ~ core resulting from fissions in material j,

= atoms of ith precursor per unit volume of fuel stream in the

N i j ~external loops resulting from fissions in material j,

f

C

=

volume fraction of fuel in the core,


226 f E = volume f r a c t i o n of f u e l i n t h e external loops, Vc = core volume,

VE = external loop volume, V

C @fcVc = r a t e of production of f i s s i o n neutrons j fj fissionable material j .

i n the core from

The precursor concentrations are described by these equations:

mi*c J =

$.

dtC

.v.c

.@- mijc ,

13 J fJ

mi*E

J= -hNijE

,

dtE

where

tC

= time

i n the core,

tE = time i n t h e external loops. The boundary conditions are:

where T

C

= time f o r the f u e l stream t o pass through t h e core,

TE = t h e for the f u e l stream t o pass through t h e external loops.

The solution t o Eqs. (1)and (2) are:

Ni jc

-A. .t

ij

-A. .t

1J C

Y


227

-A

NijE

ijtE

NijE( 0 ) e

Note that the precursor production rate is assumed constant for the fuel stream during its stay in the core. This idealized case would exist only for uniform power density along the fuel stream or for core residence times which are short compared to the half-life of the precursor. The boundary conditions become

-

Nijc ('1

-

-A E

NijE(o) e

ijTE

Eliminating N (0)in Eqs. ( 7 ) and (8) apd substituting the result in ijc Eq. ( 6 ) gives (9)

i

NijE -

L

1-e

J

The rate of decay of precursors in the external loops is

1-e

f V .Ni LE = E E TE

The total rate of precursor the fraction of the delayed NijEfEvE

TE BijVjCfjg5fcVc -

ij E

cay (at equilibrium) is f3. V . C

g5f V

J fj

C C

.

Thus

utrons which appear in external loops is

(

fEVE 1

- e-hi jTc ) [1 - 2-%,jTE)

1'

-A. .(Tc + TE)

(11)


228

Since

f ~ VT~~ , Eq. Tc =

(119 becomes

fcVc

( 1 - e-A

!?LE=

T ij c)

(

- e -hijTE

I'

-A. .(Tc + TE)

pij

For using these results in an equilibrium reactor code such as ERC-51 the term V.C .# Fay be replaced by a neutron production rate given by J fJ

where

cf

3

= reaction rate coefficient for fissions in material j.

Using this in Eq. (12) gives the following result for the number of neutrons lost in the external loops per neutron produced:

(

. 1 - e +ijTc)

f losses = 1 N j C j v j

3

(

-Ai jTE 1-e

-A. .(Tc

i

+

TE)

I

The necessary constants for 232Th, 233U, 235U, 238U, and 239Pu are:2

Group 1 2 3 4 5 6

Ai( s e' 0.0128 0.0315 0. I25 0.325 1.55 4.5

)

232m

233u

235u

238~

239pu

0.00085 0.0035 0.0045 0.0120 0.0045 0.0009

0.00020 0.00075 0.00105 0.00075 0.00025

0.0003 0.0018 0.0022 0.0023 0.0007 0.0002

0.00015 0.0017 0.0028 0.0071 0.0042 0.0015

0.0001 0.0006 O.OOOCt5 0.00085 0=0003

0.02625

0.0030

0.0075

0.01745

0.0023


229 References 1.

L. G. Alexander, ERC-5 Program for Computing t h e Equilibrium S t a t e s of Two-Region, Thorium Breeder Reactors, ORNL-CF-60-10-87 (October 20, 1960)

.

2-

A. M. Weinberg and E. P. Wigner, The Physical Theory of Neutron Chain Reactors, p. 136, The University of Chicago Press, Chicago, 1958.


230

Appendix G TREATMENT OF XENON ABSORPTION IN GRAPHITE

L.

G.

Alexander

Introduction Some 135Xe is formed directly during fission; however, the major part (>9@) is formed by the decay of 1351which has a half-life of 6.7 hours. The iodine remains in solution as the iodide ion (I-)= Thus, at equilibrium, the rate of formation of xenon in the fuel is proportional to the sum of the direct fission yields of xenon and iodine, here taken to be 0.066 atoms per fission. All of the xenon is released in the fuel salt, and since t h e halflife for the decay of 1351is long compared to the time required for the fuel to make one complete trip around the fuel circuit (about 15 seconds), the rate of release of 135Xe is nearly uniform throughout the fuel volume, being augmented somewhat in the core by the direct fission yield. For purposes of this study, the concentration of xenon in the salt was assumed to be uniform. The solubility of noble gases in molten salts is low (l), especially in mixtures of LiF-BeF2. A concentration of only 2 X loi5 atoms of xenon per cc of salt at 1200째F is in equilibrium with a partial pressure of xenon in the gas phase of one atmosphere. In the reference design case, the equilibrium pressure is about 0.06 atmosphere. Xenon thus tends to leave the salt at any phase boundary.

It may

form microbubbles clinging to the surface of the graphite moderator. It will tend to diffuse into the pores of the graphite. It can be removed rapidly by spraying a portion of the circulating stream into a space filled with helium or by subsurface sparging with helium. Xenon is also removed from the system by decay to 135CS and by reaction with neutrons to form 136Xe, which is stable and has a low neutron capture cross section.


231

1

Analy sis Watson and Evans1 have analyzed the equilibrium xenon poisoning resulting from the interaction of all these modes of production and removal, obtaining an equation which may be rendered in the form

Xe P.F. =

-?iy V

where

P.F.

=

Poison &action, or number of neutrons absorbed by xenon per neutron absorbed in fissile isotopes,

-q = neutron productions, number of neutrons produced from all sources per neutron absorbed in fissile isotopes (2.21 in reference design),

Y

I

=

c

V =

sum of fission yields of 135Xe and per fission,

13%,

taken as 0.066 atoms

neutron yield, or number of neutrons produced from fission per fission ( 2.50 in reference design),

gc = mean effective flux in reactor core (neut-cm/cm3 sec), a

=

Captures effective 13%e neutron capture cross section neut-cm Atom Xe, cc

vfc = volume of fuel stream in core (600 ft3), Vf = volume of fuel stream (2500 ft3),

'i\ = decay constant for 135Xe, 2.09 X 10'5/sec, ti

- rate of diffusion of xenon into graphite, atoms/sec, d = rate of removal of xenon by sparging, atoms/sec, S

nf

=

number of atoms of xenon dissolved in fuel salt.


232

The produce calculation.

C

f--

a i s r e a d i l y evaluated by reference t o a multigroup

u4

Thus =FVA

gcaNcVc

where N

V

C

C

=

concentration of xenon atoms i n the cor

=

volume of core, cc,

I

tom

/C

F = f i s s i o n r a t e i n core, fissions/sec, A = f r a c t i o n of a l l neutrons captured i n 135Xe.

The f i s s i o n r a t e i s readily calculated from the power, using the conversion:

3.1

X

10l6 fissions/Mw-sec.

Fc

0 =

3.1

Solving f o r the product;

X

loL6 P V vC

The r a t i o A/Nc

i s , a t low concentrations, independent of pJc and may

be determined by means of a multigroup neutron calculation. erence reactor it has a value of 2.22

X

By insertion of numerical values, it i s found t h a t captures/ c c atom Xe sec cc

I n the r e f -

gc

a equals

2.54 X

.

The r a t i o is/nf i s equal t o the volumetric r a t e of sparging divided by the volume of the f u e l stream.

Let Q be the f u e l stream r a t e of flow through the core (160 f t 3 / s e c ) and r be the f r a c t i o n of t h i s diverted

through a sparrge or spray chamber. :s/nf

Then = rQ/Vf

For the term Ad, Watson and Evans give a r e l a t i o n which may be rendered n

d

1/ 2

= N* A

g g

D(gc a + A)]

i


233

where N* = gas-phase xenon cancentrationithat is in equilibrium with xenon dissolved in the salt, atoms/cc, A

=

e

=

area of interface between salt and moderator gra hite, (For 810 logs 8 in. diam X 20 ft long, A = 31.5 X lo6 cm8 ), g porosity of graphite or fraction of graphite volume accessible to xenon,

D = coefficient of diffusion of xenon in graphite, cm2/sec.

The value of N* is related by Henry s law to the concentration in the salt. g

Nf =

Ng”

(4)

KRT

where Nf = concentration of xenon in the salt, atoms/cc,

K

=

Henry’s law constant for xenon in salt, 3.2 X 10’’ moles Xe cc of salt, atom ’

R = gas constant, 82 cc-atom/mole

OK,

T = absolute temperature, 922OK,

Values Of K for xenon dissolved in various salts at various temperatures are given by Watson and E3ans.l For MSCR salt at 1200°F, K is about 3.2 X 1Q moles Xe/cc salt-atom. Noting that nf Nf Vf, one has, from E;

Eq* (31, Ag[e

D(FC a +

tid/nf = mwf

Substituting these results and numerical values into Eq. (1)yields 1 + 42.6

Xe P.F.

=

0.0584 -k

lOOOr

X

lo4

(e

D)l12

+ 46.0 x lo4 (e


2

234 Values of the poison f r a c t i o n calculated by means of t h i s equation f o r various values of r and the product e D axe displayed below i n Table G.1. Table G.l.

0.1

0.01 0.001

Xenon Poison Fraction i n MSCR

0.054 0.054

0.0445

0.0173

0.0006

0.053

0.0445

0.0052

0.054

0.054

0.053

0.0251

~~

aI n cm2/sec.

Reference 1. G. M. Watson and R. B. Evans 111, XenoiDiffusion i n Graphite: Effects of Xenon Absorption i n Molten S a l t Reactors Containing Graphite, USAEC Report ORNL CF-61-2-59, Oak Ridge National Laboratory, February 1961.


235 Appendix H THE EQUILIBRIUM STATE AS A BASIS FOR ECONOMIC EVALUATION OF THORIUM REACTORS T. W. Kerlin Introduction

The equilibrium condition is currently being used as the basis for fluid-fuel reactor economic evaluation and for new computer code development. Because of this increasing application of calculations based on the equilibrium state, it is advisable to clearly define equilibrium and to assess the validity of evaluations based on the equilibrium condition. These problems were considered in this study for thorium-fueled reactors, fueled initially with 235U, and particularly for the molten-salt converter reactor (MSCR). The equilibrium condition is defined as that condition in which the reactor composition is time independent because of a balance between nuclide production rates and loss rates. It is important to note that these are the total production and loss rates (including feed, recycle, discharge, processing losses, etc.) and are not restricted to nuclear transformation rates. The mathematical formulation of the equilibrium state is obtained by setting the time derivative of the nuclide concen-

where Ni

=

nuclide concentra

t = time, f Qi = feed rate,

ad

Ri = recycle rate,

of ith nuclide,


.

Fi

.

236

= f i s s i o n fragment formation r a t e ,

= growth r a t e due t o neutron capture i n other materials, Ti = growth r a t e due t o radioactive decay i n other materials, Di

ti

di

= l o s s r a t e due t o neutron capture i n ith nuclide, = l o s s r a t e due t o radioactive decay of ith nuclide,

qi = processing l o s s r a t e . Equation (1)should be v a l i d f o r reactor evaluation i f Che nuclide concentration i s near equilibrium (90-.95$) Over a large f r a c t i o n (90-.95$) of the r e a c t o r ' s operating l i f e . Estimates of the s a t x a t i o n behavior of the nuclides of i n t e r e s t i n thorium-fueled reactors were made using the methods discussed below. Methods The time-dependent behavior of the nuclide concentrations i n a thorium\

fueled reactor, fueled i n i t i a l l y w i t h 235U, w a s calculated by solving

m= (1)without

the r e s t r i c t i o n t h a t dNi/dt = 0.

The treatment f o r the individual isotopes along w i t h special assumption f o r each case i s given below: 1. Fission Products

Assume t h a t f i s s i o n products xre produced a t a constant r a t e , S, and are removed by neutron capture, radioactive decay, and processing.

leads t o

mi= si dt

where T =

time t o process a complete reactor volume.

Solving Eq. ( 2 ) gives

This


t

t

I

It I

237

The fractional saturation is Nib)

--

-1-e

-(ti + di + &) 7

( 41

Ni(a)

A lower limit on the fractional saturation at any time may be obtained by

setting ti = di = 0 .

This gives

2. Uranium-233 Neglecting processing losses and captures in 233Pa, the concentration of 233U is given by

where N23 = concentration of 233U, c

NO2 = concentration of 232Th (assumed invariant), 2'

= effective capture cross section of 2 3 2 ~ ,

$5 =

neutron flux (defined below),

!

I

u23 = effective absorption cross section of a

233uD

The neutron flux is given by

Wf@= P x 3.1 X lolo

,

(7)


238

P 3.1

lolo

X

.>'

0

f

where

P = reactor power (watts),

v

= fuel volume (cc),

Nf = concentration of fuel (atoms/cc),

af = effective fission cross section (average over all fissile nuclides)

.

The term, VNf, is the total number of fissile atoms. Taking the average

atomic weight of the fissile nuclides as 235 gives I

VNf = W

X

2.563

X

lo2'

,

(9) i

where

w

= mass of fissile material (grams).

Substituting Eq. (9) into Eq. (8) gives

B =

) ;1

x

1.21 x c

af

Solving Eq. (6) and using Eq. (10)gives

- -P x -x 3.816 X N23(t) = 1 - e N23(m)

T

af J

where

T = time (years). The term, P/W, is the specific power of the fuel in units of watts/g. The ratio, ua* ~ / u ~is , about 1.2 for the molten-salt converter reactor (MSCR).


239

3. Uranium-234 The concentration of 234U is given by

where N24 = concentration of 234U, 023

C

=

capture cross section of 2 3 3 ~ ,

u24 = absorption cross section of 234U.

a

Solving Eqs. (6), (lo), and (12) simultaneously gives

N"(a3)

-e

c

(13)

$3 -

1--

a

a24

a

The ratio, c724/uf, is about 0.4 for the MSCR. a

4.

uranium-235

The concentration of 2 3 5 ~ t any time is that required by the criticality condition. For these estimates, it is assumed that the fissile inventory is constant. This gives N25(t) = N25(a)

+ N23(~)- N23(t),

N25(t) = 1 + (a) N2*5

N2'

(a)


--

240

wIf c~~~a = u25, a as is approximately the case in the MSCR, then

N23 (a)

CR

-

N25(a)

1

- CR ,

(15)'

where CR =

corrversion ratio (assumed constant).

Using Eqs. (ll) and (15) in Eq. (14) gives

N25 (t)

CR

=1+

N25(a)

1

- CR

a23 - -P 2x 3.816 e

w

x

T

u43

I

The assumptions of constant fissile inventory and constant conversion ratio are very crude, but will suffice for the qualitative evaluation desired in this study. 5.

Uranium-236 The concentration of 236U is given by

where ~~6

=

025 = C

concentration of 2 3 6 ~ , capture cross section of 2 3 5 ~ ,

u~~ = absorption cross section of 236U0

a

Solving Eq. (17) and using Eqs. (10) and (16) gives

I


241

N26(tl

- -P -o6:

I'

=l-e

N 2 6 ( a)

x 3.816 x

T

uf

23

P aa

+

CRCR

1

- -P -x 3.816 ~ 1 0 T' ~ Ui6

X

3.816X10'4

T

-e

'

023

a

1--

o2 a

The r a t i o , 026/of, i s about 0.1 f o r the MSCR. a

c

6.

(18)

Uranium-238 The 238U appears only i n the feed along with 235U

-= (235U

- N28

feed r a t e )

dt

$8

a

@

'

,

where N~~ = concentration of 2 3 %

'

a28 a = absorption cross section of 238U. The 235U feed r a t e is equal t o the nuclear transmutation r a t e

235U feed r a t e 5

= N25(t)

ui5

@ =

N25(a)

+

N23(a)

- N23(t)

u:~ @

I

(20)


i

!

242 Solving Eq. (20) gives

N28(t)

r

28

-

- - -a' = 1 - e w Of

x 3.816 x

. .

T

1 1-

N28 (a)

23

- - -a,

(--J

1 1

1

- CR

The r a t i o ,

e

X

3.816 X

T

af

a23 ($-1)

i s about 0.1 f o r t h e MSCR.

a

Results and Conclusions The r e s u l t s are shown i n Figs. H.l through H.6.

They are discussed

individually below: 1. Fission Products

Figure H . l shows the saturation behavior of a material removed only I n a c t u a l operation, the approach t o s a t u r a t i o n would be It i s c l e a r from Fig. 1 t h a t f a s t e r because of nuclear transformations.

by processing.

the f i s s i o n product nuclides which s a t u r a t e slowly with respect t o t h e i r nuclear reaction r a t e s are a t equilibrium (9@ or g r e a t e r ) f o r 9% 30-year reactor l i f e f o r processing times of less than 500 days.

Of a A 1000-

day processing time ( t y p i c a l f o r the MSCR) gives equilibrium (9w or g r e a t e r ) f o r only 8% of t h e time.

Therefore, the equilibrium treatment

i s doubtful f o r fission-product nuclides with l o w cross sections or long half l i v e s .

However, these e f f e c t s a r e small.


243

1

Reactor Operating T i m e ( y e a r s )

Fig. H.1.

Fission Product Saturation.

1


244 ORNLLR-

1.0

.9

!

.8 1

(si

=

1

1

1

1

1

l

l

specific power (watts/gram)

.7

-6 h

8 m

v

(u

'== .5 4J

v

m (u 2

.4

.3

.2

.1

0 0

2

4

6

8

10

I2

14

16

18

20

Reactor Operating Time (years)

F i g . H.2.

233U Saturation.

22

24

26

28 30


245

1.0

.9

WG. 728t

5

/

.8

.7 j s

*6 8

v

*N

2 > .5 +,

v

.t N

e-4 .4

!

I ) ;1

=

specific power (watts/gram)

.3 i

.2

.1

0

6

8

10

12

14 16

18

20

Reactor Operating Time (years )

Fig. H . 3 .

234U Saturation.

22

24

26 28

30


246

[;I

ORNGLELDWG. 72864 = s p e c i f i c power (watts/gram)

CR = -9

10

-

8

8 v

22 6 2 9

4

W

In N

2 0 0

4

8

I2

,

5 n

16

I

20

24

I

4

8 W

: 3 2

‘r=.

+>2

v

y\

N

1 0 0

4

8

I2

16

20

24 ,

4 8 3

v

N

2

L2

0 0

4

8

16

Reactor Operating Time (years)

Fig. H.4.

235U Saturation.

20

24


247

1.0

u)

tu

0

.4

cu 0

.2 0 0

1.0

.8 8

w

4

I

8

12

16

1

20

/

24

2000

28

-

1000 750

.6

.2

0 0

4

8

12

16

20

24

28

16

20

24

28

1.0 .8

0 0

4

8

12

Reactor Operating Time (years)

Fig. H.5.

236U Saturation.


248

P W- = specific power (watts/gram)

.4 .2 . 0

CR =

1.0

-8

-8 8

v

-6

to N

z5

\ to N

-4

b

.2 0 0

4

8

12

8

24

28

20

24

28

CR = .7

1.0

--

20

16

.8 -6

to N

2

to N

.4

z *

.2 0 0

4

8 Fig. H.6.

12

16

238U Saturation.


249

2. Uranium-233 Figure H.2 shows the saturation behavior of 233U. The curves show that equilibrium (9w or greater) exists for 90$ of the reactor life only if the specific power is greater than about 1800 w/gram. A specific power of 750 w/gram (typical f o r the MSCR) insures equilibrium (90$ or greater) for 8 9 of the 30-year reactor life. 3. Uranium-234

Figure H.3 shows that 234U saturates very slowly for all practical values of specific power. A specific power of 750 w/gram (typical f o r the MSCR) gives an average concentration over the 30-year reactor life only 65% of the equilibrium value. 4.

Uranium-235

Figure 4 shows the saturation behavior of 235U. For a conversion ratio of 0.8 and a specific power of 750 w/gram (typical M3CR conditions), the 235U concentration is within 1% of the equilibrium concentration for 65$ of the reactor operating life.

5. Uranium-236 Figure 5 shows that the approach to equilibrium is quite slow for 236U* For a conversion ratio of 0.8 and a specific power of 750 w/gram (typical M C R conditions), the 236LJ concentration never reaches 9C$ of the I 1

equilibrium concentration.

.

6. Uranium-238 .

.

Figure H.6 shows the approach to saturation for 238U. For a conversion ratio of 0.8 and a sp fic power of 750 w/grm (typical MSCR conditions) equilibrium (9% reactor operating lifetime.

greater) is insured for 2% of a 30-year

The results f o r the six materials considered in this study are Shawn in Table H.1 for MSCR conditions. Consideration of the methods used to obtain the results in Table H.1 does not indicate high accuracy. However, the results should be qualitatively


,

250

Table H.l. Percentage of MSCR Lifetime Having Nuclide Concentrations Within 1% of Equilibrium

Nuclide

MSCR Lifetime Having Nuclide Concentration Within 1% of Equilibrium

(4) Fission Products ( 1000-day processing)

80

233u

80

234u

20

23 5 u

65 0 20

236u 2 3 8 ~

correct; and they should create some concern Over the validity of the equilibrium state as a suitable condition for reactor economic evaluation. Table H . 2 shows the expected direction of the error introduced by assuming the equilibrium condition for the nuclides considered. It is apparent from Table H . 2 that it is not possible to predict whether equilibrium calculations are intrinsically optimistic or conservative from these results. Also, since the relative magnitude of the competing effects depends on reactor type, the characteristics of the particular reactor must enter into an assessment of the direction and magnitude of the error associated with the equilibrium assumption. The magnitude of the effect of these factors on reactor economic evaluations is not known. However, an extensive study should be made to examine these problems. Reactor evaluations based on the equilibrium condition are so convenient, economical, and unambiguous that they should be used if possible.


!

251

Table H.2. Effect of the Equilibrium Assumption on Calculated Reactor Performance Material

Effect on Calculated Performance

Fission Products

Conservative (overestimate)

2 3 3 ~

Optimistic (overestimate) Conservative" (overestimate)

234u 2 3 5 ~ 236u 23813

Optimis tic ( underestimate ) Conservative* (overestimate3 Conservative* (overestimate)

*These conclusions are based on an equilibrium calculation with no corrections. These materials are actually calculated using adjustment factors which average the concentrations over the life of the reactor. The direction of the expected error under this assumption is not known.

c


252 Appendix I ESTIMATES OF PHYSICAL PROPERTIES OF LITHIUM-BERYLLIUM MSCR FUEL AND COOLANT SALTS

R. Van Winkle Introduction Estimated physical properties of three fuel salt mixtures at 1 2 O O O F and one coolant salt at 1062OF f o r use in heat transfer and pressure drop calculations of the MSCR study are listed below in Table 1.1, Table 1.1,

MSCR Salt Properties

MSCR No, 1

Mixture Temperature , OF Composition mole % LiF-BeF2-ThF,+

MSCR No. 2

MSCR No. 3

Coolant

1200

1200

1062

71-16 13

68-23-9

66-29-5

66-34-0

941

887

860

851

66,03

56,2

46.2

33.14

215 6

190 1

163.0

120.5

3 454

3.045

2.610

1,931

Viscosity, lb/hr-ft

24.2

21

18.9

20.0

Thermal Conductivity Btu/hr- f t - O F Heat Capacity Bt u / l b - O F

2.67

2.91

3.10

3.5

0,318

0,383

0.449

0 ,526

Liquidus Temp,

OF

Mol. Wt. Density, l b / f t Density, g/cc

3

,

1200

-

.

.

The bases for these estimates, some temperature-dependent relationships and atom number densities are given for cases of interest,


253 Viscosity Experimental v a l u e s of t h e v i s c o s i t y of s e v e r a l d i f f e r e n t composit i o n s of t h e mixture LiF-BeF2-UF4-ThF4

(and some e s t i m a t e d v a l u e s ) are

l i s t e d i n T a b l e I , 2 a t temperatures of 600째, 700" and 800째C,

These are

p l o t t e d i n Fig, 1.1 which a l s o i n c l u d e s p l o t s of what appear t o be r e a sonable estimates of t h e v i s c o s i t y of MSCR S a l t s 2 and 3,

The e s t i m a t e s

on t h e s e two s a l t s depend on t h e fact t h a t t h e i r compositions l i e between t h o s e of Mixture 75 and Mixture 133 (which is t h e same as MSCR No. 1); hence, t h e i r v i s c o s i t y curves may be expected t o l i e between t h e curves

of t h e two known mixtures.

V i s c o s i t i e s and temperature-dependent v i s -

c o s i t y equations of MSRE f u e l and coolant salts a r e l i s t e d i n Tables 1.3 and 1.4 for comparison, The v i s c o s i t y equation for Mixture 133 (MSCR No. 1) is (1): rl

= 0.0526 exp(4838/T

OK)

centipoise

Some P h y s i c a l P r o p e r t i e s of Various T a b l e 1.2. Lithium-Beryllium Molten Fluoride S a l t s

Mixture

Composition, m/o LiF-BeF2LUFq-ThF4

V i s c o s i t y , cp 600 700 800째C

74

69

31

0

0

7,s 7,25

4,9 4.54

75

67

30.5

2.5

0

80

5.5*

111

7 1 16

112

50

Number

12

Heat Capacity(:! a t 700째C Mol

. Wt .

3.45(5) 3.10(6)

0.67*(5)

3.8*(5)

0 57*(5 1

39.5

0 37a ( 5

66.02

0.65*(5 1

36.5

t

13.0

7.1

4.8(5)

22.5

11.6

6.03(6)

50

131

60

36

4

0

3.0

7.96

5,30(6)

132

57

43

0

0

3.4

7.38

4.50(6)

133

71 16

0

13

13.4

7.55

4.76(1)

32.4

. .

--

44.96 35.03

-I

0 e 306 (1)

*Estimated values ( a l l o t h e r s l i s t e d are experimental) Numbers i n p a r e n t h e s i s are r e f e r e n c e s .

.

.

66 03


t

254

.

ORNL-LR-Dwg 62611

950 1000

800

1109

850

TEMPERATURE,OF I 2 0 0 1300 1400 1500 16001700 1800

TEMPERATURE, OC 900 950 I000 I100 TEMPERATURE,

Fig. 1.1.

O K

MSCR Salt Viscosity.

1200

1300


255

-V

Table 1.3.

I.

11. 111.

Composition and P r o p e r t i e s of Fuel S a l t f o r MSRE

Chemical Composition

Mole P e r c e n t

LiF

70

BeF2

23

ZrF4

5

ThF4

1

uF4

1

Molecular w t

43 59 0

Physical Properties Density (above l i q u i d u s ) l b / f t

t in

3

OF

@ 12OOOF Liquidus, OF

154.5 842

Heat c a p a c i t y , Btu/lb-OF Solid

212

Liquid

887

- 806OF

- 1472OF

@ 120OOF

+

0.132

(4,033 x 10'4)t

- (9.99

0.575

V i s c o s i t y , c e n t i p o i s e , T i n OF Range:

Thermal c o n d u c t i v i t y , Btu/hr-ft-OF

t .in

138e 6 0.1534 e6476/T

1122-1472OF

@ 12OOOF

7.64 2.74

+

-

(5.516 x 1 0 m 4 ) t (1.37 x 10-71t2

OF

@ 12OOOF

3.21

Table 1 , 4 * Composition and P r o p e r t i e s of Coolant S a l t f o r MSRE

11.

lO")t

0.455

Heat of f u s i o n ( @842OF), Btu/lb

I.

X

Chemical Composition

Mole P e r c e n t

LiF

66

Be F2

34

Molecular Weight

33 1 4


256 T a b l e 1.4, 111.

Continued

Physical Properties ( a t normal ave operating Temperature, (1062OF) Density, l b / f t

3

120e 5

Viscosity, l b / f t - h r

2000

Specific heat, Btu/lb-OF

- 68OOF) Liquid (896 - 1508OF)

0.210

@ 1062OF

0 526

Solid (122

0.174

.

Thermal conductivity, Btu-ft/ft2-hr-OF Liquidus temperature,

+ +

(4.71

10-~)t

(3.31 x 10'4)t

3.5 851

OF

Heat Capacity I

The temperature-dependent equation for heat capacity of Mixture 133

is (1): C

P

= 0.473

- 0.000238

T cal/g-OC (T = OC)

Values of heat capacity of MSCR S a l t s 2 and 3 a t a given temperature may be estimated by i n t e r p o l a t i n g between published values of o t h e r s a l t s of d i f f e r e n t molecular weights, since heat capacity of molten salts probably v a r i e s inversely with molecular weight,

Published values of heat capa-

c i t y a t 7OOOC of some known salts are shown i n Fig, 1.2, which contains

a p l o t of heat capacity as a function of molecular weight. Heat capacity r e l a t i o n s h i p s for MSRE f u e l and coolant s a l t s are shown i n Tables 1.3 and 1.4.

Thermal Conductivity Like heat capacity, thermal conductivity can be expected t o vary inversely with molecular weight.

An estimate of t h e thermal conductivi-

t i e s of t h e MSCR f u e l s a l t s has been made by extrapolating t h e published


257

0.7

/-E R E Coolant S a l t

(700OC)

e

0.6 0

P

rl \

MSRE Coolant S a l t

5

-P

.

(572째C)

m

3

I

0.5

.?I

s

2 V 0.4

I

0.3

I

-9

ld

I

i 30

40

W I

I

-

I \ I

I \J

50

60

(700OC)

I 70

Mol. W t *

Fig. 1.2.

MSCR Salt Thermal Conductivity and Heat Capacity.

.-

-


258 values of t h e lower molecular weight MSRE s a l t s t o t h e higher weights of t h e MSCR salts.

This extrapolation is shown i n Fig. 1.2.

However, t h e

published values may not have much precision or accuracy (1)

.

Density Temperature dependent r e l a t i o n s h i p s for c a l c u l a t i n g t h e d e n s i t i e s of t h e t h r e e MSCR s a l t s are (2):

P~~~~

NO.

1

%SCR

NO.

2

"MSCR

NO.

3

= 3.934 P

3.480

= 2.993

- 7,4 x 20

-4 T

-

g/cm

(T

= OC)

6.7 x l O - " f

- 5.9

x 10-4T

The b a s i s f o r these estimates is t h e same as given i n reference (31, page 123 with an added term f o r t h e i o n i c volume of thorium equal t o 2.82

- 2.94

x 10-3T.

Liquidus Temperature Figure 1.3 (from reference 4 ) was used t o obtain t h e liquidus t e m perature for t h e MSCR salts.

Mixture 133 (MSCR No. 1 ) has a liquidus

temperature of 505OC (p. 58, ref. 4).


259

ORNL-LR-DWG 37420AR2

Thg 4 4 4 4

A

TEMPERATURE IN OC COMPOSITION IN mole Oi0

.

526

P 465

Fig. 1.3.

E 370

The System LiF-BeFZ-ThF4.

BeF2 548


i

i

c 1

260 References 1. Private communication from H. W. Hoffman. 20

Private communication from S. Cantor.

30 Oak Ridge National Laboratory, MSRP Semiann. Prop, Rept. August 1, 1960 t o February 28, 1961, USAEC Report ORNL 3122.

40 C. F. Weaver e t al., Phase E q u i l i b r i a i n Molten S a l t Breeder Reactor Fuels. 1. The System LiF-BeF2-UF4-ThFbr USAEC Report ORNL-2896, Oak Ridge National Laboratory, December 27, 1960, 50

S o I. Cohen, W. D o Powers, and N o D. Greene, A Physical Property Summary for ANP Fluoride Mixtures, USAEC Report ORNL-2150, Oak Ridge National Laboratory, August 23 e 1956. (Declassified 11/24/59.

.

Bo C. Blanke e t al., Density and Viscosity of Fused Mixtures of Lithium, Beryllium, and Uranium Fluorides, USAEC Report MLM-1086, Monsanto Research Corporation, March 23, 1959. W. L. Breazeale, Revisions t o MSRE Design Data Sheets MSR-61-100, August 15, 19610 .

- Issue No.

4,


261 Appendix J FUEL AND CARRIER SALT COST BASES

W, L o Carter Introduction A survey was conducted among suppliers of ThF4, Tho2, LiF, BeF2,

ZrF4 and NaF t o determine current market prices,

The d a t a are evaluated and a recommended s e t of values t o be used i n molten s a l t converter and breeder r e a c t o r calculations is presented,

The cost data are needed i n

c a l c u l a t i n g f u e l cycle c o s t s , The following values are recommended f o r use i n molten s a l t r e a c t o r calculations: ,

ThF4

$ 6.50 per pound

BeF2

7,OO per pound

L iF

l4,,70 p e r pound

ZrF4

4.00 per pound

One o f t h e purposes of t h e study of thorium breeder and converter r e a c t o r s is t o furnish comparative f u e l cycle c o s t data on t h e various systems as w e l l as nuclear performance data.

The market survey w a s con-

ducted t o obtain current and r e l i a b l e p r i c e information on s e v e r a l chem-

i c a l compounds which w i l l be needed i n r a t h e r l a r g e inventory and f o r i

which t h e consumption rate may be s i g n i f i c a n t .

The i n q u i r i e s were concerned primarily with molten s a l t r e a c t o r materials, namely, thorium

f l u o r i d e , lithium f l o r i d e , beryllium f l u o r i d e , zirconium f l u o r i d e and sodium fluoride; i n addition, p r i c e s were obtained f o r thorium oxide. Several manufacturers of these chemicals were contacted, and a summary of t h e i r p r i c e schedules is given i n Tables J.l through J060

Since it is appropriate t o a s s o c i a t e a date with a market quotation, it may be

noted t h a t these d a t a were obtained during t h e period November 1961 January 1962,

There is one exception:

-

t h e comparative f i g u r e s quoted

from document Y-1312 were published i n December 1959,


Table J.1. Cost Data for Thorium Fluoride cost ~

$/lb ThFba

$/lb ThF4

$/kg Tha

$/kg'Th

GCDb

Y-1312'

GCD

Y-1312

6.50

17 52:

18.98

~~

Vendor or Source of Information Quantity Initial order of 127,000 kg

6.00d

'

m 4

Replacement rate of 37,100 kg T W 4 / J n . Initial or der of 1,271,000

(a)

(a)

6.OOd

17.52d

kg w 4

Replacement rate of 371,000 kg W 4 / Y r %o assay given for the material.

-

bGCD General Chemical Division, Allied Chemical Corporation, P.d. Box 70, Morristown, New Jersey. This vendor says price is only a rough estimate. C R. G. Orrison, Thorium Metal Processes, Y-1312 (December 18, 1959). This price is not based on the indicated quantity. %o distinction made in price because of quantity or rate.

e,


Table 5.2.

Cost and Composition Data f o r Thorium Oxide Cost Data cost $/lb Tho2 $/lb Tho2 $/lb Tho2

Vendor or Source of Information ma Material Designation Code 111 Material Form Powder Cost quoted for Initial order of 127,000 kg

AP

$/kg Th

$/kg Th

$/kg Th

Y-1312b

Code 112 Powder

AP AP Y-1312 Code 111 Code 112 Not stated Powder Powder Not stated

7.00

7.50

5.75-8.50

7.00

17.52

18.78

7.50

17 52

18.78

7.00

7.50

17.52

18.78

7.00

7.50

17.52

18.78

14.39-21.28

Tho2

Replacement rate of 37,100 kg Tho2/Y-r Initial order of 1,270,000 kg Tho2 Replacement rate of 371,000 kg Tho2/Yr

Composition Data Typical Analysis (ppm unless indicated) Vendor or Source of Information Element or Campound Tho2

AP 9976 min

AP 9% rnin

Y-1312

Not given

aAI? -American Potash and Chemical Corporation, 99 Park Avenue, New York 16, New York. bR. G. Orrison, Thorium Metal Processes, Y-1312 (December 18, 1959). This price is not based on the indicated quantity.


...........

. . . . . . . . . . .

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

. .

....

.

,

i

.

..

. .

,

Table J. 2 (continued) Composition Data Typical Analysis (ppm unless indicated) Vendor or Source of Information

AP

Elememt or Compound (continued) Rare earth oxide 50 Sulfate, SO3 100 Phosphate, P205 50 Fe 50 CaO 100 100 Mgo 2000 Na + K + Li Silica, Si02 500 Boron, B Uranium, U 5000 Loss on ignition 10 Sm 1 Esl 10 Gd QY

5

AP 30 50 10 6 10 1 1 5 0.1 10 5000 1-2 0.2 1 1

Y-1312

......

-.


C

i

Table 5.3.

Cost Data'

for Lithium Fluoride b

Cost

Quantity Initial order of 16,000 kg Li

(e)

Replacement rate of 5,100 kg Li/yr (e) a The sources of information are Atomic Energy Commission-Oak Ridge Operations, Oak Ridge Gaseous Diffusion Plant, and Union Carbide Company Y-12 Plant. bThe price is for (c) at. $ 7Li and includes a basic charge for the material produced as the monohydrate (LiOH*H20)plus a conversion cost for LiOH-H20+ LiF plus a feed cost plus an AEC overhead cost. 'C3.as sified information.

-

F


Cost and Composition Data for Beryllium Fluoride

Table J.4.

Cost Data -

Cost $/lb BeF2

Vendor Quantity' Initial order of Replacement rate Initial order of Replacement rate

23,800 kg BeF2 of 6,950 kg BeFz/yr 238,000 kg BeF2 of 69,500 kg BeFn/yr

$/kg Be

BBCO'

BCorpb

BBCo

BCorp

6.66 6.00 6.48 6900

7-25 7.25 6-95 6.95

76.52 68.93 74.45 68.93

83.29 83.29 79.85 79.85

Composition Data Element or Compound BeF2 H2O Fe Ni

Cr.* A1 S

Manufacturing Specifications 99.5 2 0 . 5 wt $ 0.1 0.5 wt 8 400 ppm rnax 100 ppm max 100 ppm max 200 ppm max 500 ppm max

-

Typical Analysis of Manufactured Materiald 99.5 w t $ min 0.06 wt$ 25 PPm 20 ppm PPm 90 PPm 750 ppm \

%rush Beryllium Company, 5209 Euclid Avenue, Cleveland 3, Ohio. bThe Beryllium Corporation, Reading, Pennsylvania. C Shipped as I X 1 X 1-inch lumps. dThe analysis given is f o r BBCo material; no analysis was given f o r BCorp material.

1

I


Table J.5.

Cost and Component Data f o r Zirconium Fluoride Cost Data cost

Vendor Quantity Initial order of Replacement rate Initial order of Replacement rate

55,100 kg ZrF4

of 16,100 kg ZrFq/yr 551,000 kg ZrF4 of 161,000 kg ZrF4/yr

TADa

GCDb

TAD

GCD

4.00 4.00 3.55

4.50 (c ) (c) (c)

16.13 16.13 14.32 14.32

18.15

3.55

Composition Data Typical Assay (wt 46) Vendor Element or Compound ZrF4 Chlorides S â&#x201A;Źif

Fe Ni ~~

-

TAD

GCD

Not given

98.5+

0.007 0.003 0.01 0.03 0.003

~~

"TAD Titanium Alloy Manufacturing Division, National Lead Go., 111 Broadway, New York 6, New York. No assay was given; however, the bid was f o r Hf-free material.

-

bGCD General Chemical Division, Allied Chemical Corp., P.O. Box 70, Morristown, New Jersey. This vendor says the price is approximate. C No cost distinction is made for quantity or aate.


Table J.6.

Cost and Composition Data'

for Sodium Fluoride

Cost Data cost

b

$/lb NaF

$/kg Na

0.135 0.139

0.558

Shipping Containers 100 l b multiwall paper bags 375 l b leverpak f i b e r drums

0.542

Composition Data Typical Assay ( w t $ ) Elements Na Ca A1 Si Fe, Mg cu sc K, Ba B, Mo T i , Mn, Co, N i , V C r , Ag

Major constituent 0.05-0.5 0.034*3

0.01-0.1 0.0005-0.005 each Trace, <0.0001 Not detected, <O. 1 Not detected, <0.01 each Not detected, <0.001 each Not detected, <0.0005 each Not detected, <0.0001 each

' A l l information is from the Blockson Chemical Company, P.O. Box l407, J o l i e t , I l l i n o i s .

bFor an i n i t i a l order of 75,80o-r758,000 kg NaF and a replacement r a t e of 22,100-221,000 kg NaF/yr

.

C)


269

Bases f o r Establishing Prices I t was assumed t h a t p r i c e s would be established f o r a l a r g e molten

s a l t , power-producing system.

Two conditions were visualized:

(a) A

1000 Mwe (2500 Mwt) s t a t i o n and (b) t e n of these 1000 Mwe s t a t i o n s i n simultaneous operation. t o r y requirements

These conditions established t h e i n i t i a l inven-

The consumption rates were calculated by assuming

t h a t t h e f u e l s a l t would be discarded after removal of f i s s i l e material on a 1000-day cycle,

I t was assumed t h a t thorium and carrier s a l t could

not be decontaminated and recovered. Vendors were asked t o quote prices on t h e b a s i s of producing mate-

r i a l s i n t h e q u a n t i t i e s desired by e x i s t i n g methods and according t o current specifications.

I t was not considered appropriate t o a s k a ven-

dor i n an information-seeking survey such as t h i s t o do much research i n t o manufacturing procedures and schedules i f h i s operations would be s i g n i f i c a n t l y affected by these additional q u a n t i t i e s .

Consequently, no

r i g i d s p e c i f i c a t i o n s were affixed t o these chemicals o t h e r than t h e obvious one t h a t a l l materials should have extremely low concentrations of I

high neutron cross section materials,

General Comments on Price Quotations Thorium Fluoride Only one manufacturer was i n t e r e s t e d i n making a quotation f o r ThF,+, and it was admitted t h a t t h i s was a rough approximation. The p r i c e compared favorably with t h e value quoted by Orrison (1) i n a previous market survey, Zirconium Fluoride Vendors were asked t o quote on hafnium-free ZrF4.

The q u a n t i t i e s

requested are apparently q u i t e l a r g e compared with available production.


270 Beryllium Fluoride The suppliers i n d i c a t e t h a t t h e r e would be no problem i n supplying t h e q u a n t i t i e s requested.

Beryllium fluoride is an intermediate compound

i n t h e production of beryllium metal.

Although t h e vendors s t a t e t h a t

t h e p r i c e s quoted are t e n t a t i v e , they are probably r a t h e r accurate. Sodium Fluoride This chemical is a v a i l a b l e i n l a r g e supply. be q u i t e firm.

The quoted p r i c e s should

Sodium fluoride is so inexpensive t h a t i t s use i n a

reactor contributes negligibly t o t h e f u e l cycle c o s t , Thorium Oxide Thorium oxide is a v a i l a b l e i n l a r g e supply; t h e quoted p r i c e s are perhaps rather f i r m .

The quotes compare favorably w i t h t h e values given

by Orrison (1). Lithium Fluoride Lithium f l u o r i d e occupies a s i n g u l a r position i n t h e molten s a l t f u e l system p a r t i c u l a r l y with regard t o a v a i l a b i l i t y and price. Since t h e lithium content must have a high 'Li

assay, t h e only source of mate-

r i a l is from AEC production f a c i l i t i e s .

AEC-ORO,

Y-12 and ORGDP per-

sonnel ( 2 ) were instrumental i n developing a classified p r i c e schedule

for high i s o t o p i c p u r i t y 'Li i n q u a n t i t i e s t o meet t h e requirements of

a large molten s a l t power i n s t a l l a t i o n . Lithium is produced as Li0H.H 0 f o r which a reasonably accurate 2

p r i c e can be computed. However, an uncertainty e x i s t s i n t h e charge for converting t h e hydroxide t o an anhydrous f l u o r i d e product s i n c e t h i s operation has not been attempted except i n small-scale batches; t h i s charge was estimated. A schedule of b a s i c charges was recommended (3,4) for computing t h e p r i c e of lithium f l u o r i d e i n 99.995 at. % 'Li:

i


271 $/kg Li* L i O H H2 0

LiOH*H20

-

L i F conversion

=

--

Cascade feed cost ( 4 )

Base cost Plus 15% AEC overhead (4)

The values have not been released by t h e Atomic Energy Commission as an o f f i c i a l p r i c e f o r high p u r i t y 7 L i ; they are c o n f i d e n t i a l information f o r i n t e r n a l use only. The o f f i c i a l p r i c e was $l?O/kg of L i metal as t h e f l u o r i d e i n November 1959 (5) f o r a grade containing 99.99% of 7L i . adopted f o r material containing 99.995%

This price w a s

7

L i produced i n l a r g e q u a n t i t i e s .

Recommended Values f o r Molten S a l t Fuel Based on t h e values given i n Tables J . 1 , 5 . 3 , J 0 4 , and J . 5 , t h e f o l lowing p r i c e s are recommended for use i n molten s a l t r e a c t o r f u e l cycle calculations.

These p r i c e s were chosen more or less a r b i t r a r i l y from t h e

accumulated d a t a since there was no reason t o have more credence i n one value than another. $/lb of

$/kg of m e t a l

Element

f l u o r i d e compound

(as f l u o r i d e

Thorium

6,50

19 * 00

Beryllium

7.00

80.42

14.70

120 00

4.00

16.13

Lithium Zirconium

*This information is classified.

b


272 References

1, R. G. Orrison, Thorium Metal Processes, Y-1312 (December 18, 1959), 2,

Personal communications with E. E, Keller, AEC-ORO, Y-12 Plant, and Murray Hanig, UCC-ORGDP,

Neal Dow, UCC-

3, Letter from R, G o Jordan (Y-12) t o C, A, Keller (AEC-ORO), No, Y-AO-2314, January 10, 1962, 4,

Personal communicarion with E. L. Keller, AEC-OR0 (January 18, 19621,

5,

Nucleonics, 17(11): 31 (November 1959).

,


273

Appendix K MSCR POWEX LIMITATION RESULTING FROM MODEBATOR THERMAL STRESS R. H. Chapman Summary

I n order t o achieve the m a x i m u m performance from the MSCR, it i s necessary t o operate a t a power l e v e l corresponding t o some l i m i t i n g conc

dition.

The l i m i t i n g condition may be a r b i t r a r i l y imposed upon t h e system

or it may be an inherent f e a t u r e i n the design.

I n c e r t a i n conceivable

s i t u a t i o n s thermal s t r e s s i n the graphite moderator may l i m i t t h e power output of a single r e a c t o r core.

It i s therefore of i n t e r e s t t o estimate

t h e m a x i m u m power output as a function of core diameter f o r the case where thermal s t r e s s i s the l i m i t i n g condition. It i s assumed f o r the purpose of t h i s memorandum t h a t the reactor core i s a cylinder of L/D = 1.0.

Honeycomb shaped f u e l channels are formed by

the proper spacing of hexagonal graphite moderator prisms.

For the purpose

of estimating t h e thermal s t r e s s , it i s a l s o assumed that t h e hexagonal prisms can be approximated by a r i g h t c i r c u l a r cylinder of a diameter equal t o the distance across the f l a t s of the hex.

It i s noted from the geometry

of the hexagonal u n i t c e l l that the f u e l volume f r a c t i o n , fv, the f u e l channel thickness, tc, and t h e distance across f l a t s , dc, a r e i n t e r r e l a t e d . Figure K . 1 shows the r e l a t i o n s h i p f o r the region of i n t e r e s t .

The s i z e

of the u n i t c e l l is obtained by adding the channel thickness t o t h e d i s tance across f l a t s of t h e moderator prisms. Assuming uniform heat generation, the maximum thermal s t r e s s ( a t the surface) i s given'

f o r a s o l i d cylinder as 1 aE

* t =2- 1-V -

ro2

q"/

4k

where

u

- t a n g e n t i a l thermal s t r e s s , p s i , t -


274 ORNL-LR-DWG

77111

1.5 c

1.4 7 v)

1.3

1.2 I. I

1.0 3.9 3.8

3.7

82 d t 5 I

g

16

2

3.5

g

3.4

"*

ll

3.3

)I2 3. I 0.075

0.100

0.125

0.150

0.175

0.20 0

0.225

Fv=FUEL VOLUME FRACTION

Fig. K . l .

Hexagonal Unit Cell Relationship.

01250


275

in.

a

= c o e f f i c i e n t of thermal expansion, in.

"F

'

E = modulus of e l a s t i c i t y , p s i , tr = Poisson's r a t i o , BTU

q / / / = volumetric heat source,

r

0

=

> hr f t 3

radius of cylinder, f t ,

k = thermal conductivity

BTU f t "F

' hr

It i s assumed t h a t the uniform volumetric heat generation term can be r e l a t e d t o t h e core heat production r a t e by

where = f r a c t i o n of t o t a l heat which i s produced i n moderator, P Qc = core heat production r a t e , Mwt, f

VM = volume moderator i n core, f t 3 , C

'max

1

= conversion factor = 3.413 X lo6

-

7

BTU hr-Mw'

r a t i o of' peak-to-average core power density.

-

P

The core moderator volume i s given for L/Dc = 1.0 as

Substituting Eq. (2) and ( 3 ) i n t o (1)and remranging, the following expression i s obtained:


276

Assuming values for the various properties of graphite, i.e., ut, a, E,

!

V,

and k, and representative values for the fraction of total heat generated in the graphite and for the peak-to-average power density ratio, a family of curves of heat production versus diameter may be plotted wherein ro is

the parameter. Choosing an allowable stress value of graphite I s a somewhat arbitrary operation. Experience has shown in the production of AGOT graphite for the EGCR (16 in. X 16 in. X 20 ft long columns) that the mechanical properties vary widely between different blocks of the material, vary across the cross-section of the large blocks, and depend on the orientation relative to the direction of extrusion.2 Fracture strain is probably a better criteria for failure than stress, since it has been shown to be fairly constant at about 0.1 to 0.2%. Tests have also shown that strength is not

c

temperature dependent at least up to 1100째F. Since the ratio of stress to modulus of elasticity appears in Eq. ( 4 ) ) the value of the fracture strain can be substituted and thus side-step the issue of fraction stress. A value of 0.1% is assumed as a failure criterion. For Poisson's ratio a value of 0.4 is recommended, and for the thermal coefficient of expansion, a value of 2.7 X loa6 in./in. OF is used.3 A value of 15 BTU/hr-ft*"Fis used for the thermal conductivity at ~ ~ O O " F O ~ A value of 2.5 is assumed for the ratio of peak-to-average power density to account for the maximum thermal stress condition. This is about the value computed for the MSRE.5

A value of 0.05 is assumed for the fraction

of total heat produced in the moderator, essentially the same as for the

Msm. Multiplying the heat rate by the overall net thermodynamic efficiency, one obtains the net power output. An assumed value of 40$ efficiency is used. With these assumptions Eq. ( 4 ) becomes

P 1 - fv

= 0.0196

Dc3 ,2 r0

Mwe (net)

.

(5)

t


277

Figure K.2 i s a p l o t of Eq. ( 5 ) f o r the range of i n t e r e s t .

Also

shown i n the f i g u r e a r e 500 and 1000 Mwe condition f o r f u e l volume f r a c t i o n s of 0.10, 0.15, and 0.20.

With %he figure, and f o r a given s e t of

conditions one i s able t o estimate quickly the m a x i m u m graphite s i z e permitted by thermal s t r e s s considerations.

For example, i f it i s desired

t o provide 1000 Mwe w i t h single core of 20 f t dim and 10 vol.

f u e l , it

i s seen from the f i g u r e t h a t the l a r g e s t graphite moderator prism i s

limited t o about 9 in. across f l a t s . It should be pointed out t h a t the data used i n constructing the f i g u r e

a r e subject t o considerable uncertainty.

Design data f o r large sections of

graphite such as l i k e l y t o be used i n the MSCR axe, of course, unavailable

a t t h i s time.

However, i t i s believed t h a t the r e s u l t s obtained from t h e

f i g u r e w i l l be conservative.


278 QRNL-LR-DWG

76828

L5" 1800

I600

4.0"

I40C

l20C 0.5" Q

L

. tom 5rO"

8oa

64" 6oa

7.0 " 4oc

880"

ro=IO" 20c

I

0

I

I

1

I

Cube of Core Diam, ft. I

IO

I

I

I

17.5

15

12

19

Core Diarn, ft.'

Fig. K.2.

Thermal Stress Power Limitation.

I 20


279 References

1. Ralph Hankel, S t r e s s Temperature Distributions, Nucleonics, Vol. 18, No. 11, pp. 168-169 (November 1960 )

.

2.

B. L. Greenstreet, Oak Ridge National Laboratory, personal communication.

3.

S . E. Moore, Oak Ridge National Laboratory, personal communication.

4.

L. G. Alexander e t a l . , Thorium Breeder Reactor Evaluation, P a r t 1 Fuel Yield and Fuel Cycle Costs i n Five Thermal Breeders -Appendices, ORNL-CF-61-3-9, Appendices, p . 33 (May 24, 1961).

5.

C. W, Nestor, WRE Preliminary Physics Report, ORNL-CF-61-4-62 ( A p r i l 9, 1961).


I

280 Appendix L VOWMES OF FUEL SALT AND INTERMEDIATE COOLANT SALT FOR 1000 Mwe MOLTEN SALT CONVERTER REACTOR D. B. Janney

I n t r cduct i o n The volumes of t h e s a l t streams (including heels i n dump tanks, e t c . ) were calculated f o r t h e reference design described i n Sec. 4 . Fuel S a l t Volume Table L.l.

Volume of WCR Fuel S a l t

A.

Reactor

1360 f t 3

B.

Piping

320 f t 3

c. pumps

130 f t 3

D.

Primary Heat Exchangers

575 f t 3

E.

Dump Tanks, Control Tanks

115 f t 3

TOTAL

2500 f t 3

These volumes were calculated a s follows: A.

Reactor Core ( L = 20 f t , D = 20 f t )

0.785 (20)3

X

1% = 630 f t 3

Annulus ( L = 20 f t , t = 1 i n . )

I2 (20)2 Top Plenum ( h = 13 in., D = 20 f t , i n c l . core hold-down grid* of = 4 in., t = 1 in.) ht3 *See Fig. 4 . 8 .

=

105 f t 3


281

p.785 ( 2 0 ) 2

g] - [0*785 5.2 ( 2 0 ) 2 x 24 [$I 6 X

41 = 340

- 48

=

290 f t 3

31 = 288

- 36

=

250 f t 3

Bottom Plenum ( h = 11 i n . , D = 20 f t , i n c l . core support grid* of h t = 1 in.)

(20>2

E]- [

0.785 6 x i .220 ) 2 x 24

I$]

Dome ( L = 6 , D = 6 )

0.785 ( 6 ) 3 x

1

2

( l i q . vol.) =

REACTOR TO!TAL

B.

85

ft3

1360 f t 3

Piping

Pump Suction ( 6 f t , 14 i n . , Sch. 2 0 )

" x - 140.5 144 Pump Discharge ( 2 0

ft,

12 i n . . Sch.

t3

-

20 x 111.9 x 8 = 144 Reactor Inlet ( 2 5 ft, 10 n. Sch.

0)

25 x

82.5 x 144

8

=

Misc. Piping (Estimate)

115 f t 3 30 f t 3 320 f t 3

PIPING TOTAL

C.

125 f t 3

Pump Bowl (8) (Equiv. ann. 1 f t , equiv. depth 1 1 / 2 f t ) T X

l(3 1 / 2 )

X 1

1/2

X

8 =

130 f t 3


282

D.

Primary Heat Exchanger ( 8 ) ( S h e l l I D 43.75 in., s h e l l length 13 1 / 2 f t )

0.785 (43.75)2

E

+ 0.667 x

- 0.785 T

(0.5) 144

4050] 13.5 x 8

($j)3 (1 -

g)] + 8 = 530

45 =

575 ft3

coolant S a l t volume Table L.2.

Volume of Coolant S a l t 2425 f t 3

A.

Superheaters

B.

Reheaters

485 f t 3

C.

Primary Heat Exchangers

510 ft3

D.

Piping

E.

Pumps

F.

Flush S a l t

2385 ft3

TOTAL

8615 ft3

2710 f t 3

100 f t 3

These volumes were calculated as follows:

A.

Superheaters ( 16 ) ( S h e l l I D 31.5 in., U-shell length 58 f t , 0.5 i n . tube bundle annulus) (0.0029 f t 3

S

+ ["

~

yz

x 785 x 58 x 16) ~

(31) X 58

~

X

161 = 2112

~

+

~

*

313 = 2425 ft3

J


283 B.

Rehaaters (8) ( S h e l l , I D 3 1 i n . , U-shell length 23.6 f t , 0.5 i n . t o be bundle annulus) 3

ft

C.

s a l t vol. f t tube

X

766 X 23.6 X 81

Primary Heat Exchangers (8) (Tube ID 0.43 i n . , tube length 25 f t , 2025 tubes)

[0*7851240*43)2

x 25 x 2025 x 81 + 100 (H-X heads) = 410

D.

+

100 =

510 f t 3

Piping (avg. l e n g t h s ) Primary H-X o u t l e t (155 f t , 14 i n . Sch. 2 0 ) 155

X

8 = I250 ft3

Primary H-X i n l e t (125 f t , 14 i n . Sch. 2 0 )

I25

-

140.5 X 8 = 144

970 f t 3

Superheater i n l e t / o u t l e t ( 5 f t , 8 in. Sch. 20)

-

5 51 8 144

X

32 =

60 f t 3


t

284

n

w-

Reheater i n l e t / o u t l e t ( 5 f t , 5 in. Sch. 4 0 )

-

5 20 144

X

16 =

10 ft3

Pump Suction (70 f t , 12 in. Sch. 20)

118

70 -144 X PIPING TOTAL

E.

8 = 460 f t 3 2710 f t 3

h p s (System pressurizing and s a l t sampling volume only) 100 f t 3

F.

Flush S a l t (Amount required equal t o reactor system volume )

2385 f t 3


2 85

Appendix M EVALUATION OF A GRAPHITE REFLECTOR FOR THE MOLTEN SALT CONVERTER REACTOR

T o W, Kerlin Introduction Nuclear c a l c u l a t i o n s on l a r g e molten s a l t c o n v e r t e r r e a c t o r s (MSCR) i n d i c a t e t h a t t h e neutron leakage i s l a r g e enough ( 2 t o 4% o f t h e neutrons produced) t o warrant c o n s i d e r a t i o n o f a g r a p h i t e reflector,

A r e a c t o r was

chosen from a set p r e s e n t l y under s t u d y t o e v a l u a t e t h e d e s i r a b i l i t y o f a reflector.

This r e a c t o r , which c u r r e n t r e s u l t s i n d i c a t e is n e a r t h e

optimum w i t h r e s p e c t t o f u e l c y c l e costs, has t h e c h a r a c t e r i s t i c s given i n T a b l e M.1.

Table M . 1 .

Typical C h a r a c t e r i s t i c s of MSCR

Diameter of core, f t Height of core, f t Carbon-to-thorium r a t i o Fuel salt composition Fuel s a l t volume f r a c t i o Fuel p r o c e s s i n g rate, f t /day

17.7 17.7 293 68LiF-23BeF2-SThF,+ 0.10 2

9

The core c o n s i s t s of a g r a p h i t e matrix i n s i d e a 2-inr-thick vessel.

INOR-8

Because of t h e d i f f e r e n t c o e f f i c i e n t s of thermal expansion, t h e

vessel w i l l move away from t h e g r a p h i t e when t h e reactor is a t power, c r e a t i n g an annulus.

Since no adequate metal-to-graphite

seal is a v a i l -

able, t h i s annulus w i l l c o n t a i n fuel s a l t . The d e s i g n e r may choose t o p l a c e a reflector between t h e core and t h e annulus by merely i n c r e a s i n g t h e s i z e of t h e g r a p h i t e r e g i o n and o m i t t i n g ,

f u e l channels i n t h e outer p o r t i o n .

The d e s i g n e r a l s o might choose t o p i n g r a p h i t e blocks t o t h e i n s i d e of t h e v e s s e l so t h a t t h e r e f l e c t o r moves with t h e v e s s e l , c r e a t i n g an annulus between t h e core and t h e reflector.


286 A t h i r d p o s s i b i l i t y is a combination of t h e above ,NO methods, crea ing an

annulus between two r e f l e c t o r regions. Calculations were made t o determine t h e r e l a t i v e c h a r a c t e r i s t i c s of t h e MSCR with (a) no r e f l e c t o r , (b) a 15-in0 r e f l e c t o r outside t h e annulus, ( c ) a 7.5-in.

graphite region between t h e core and annulus and a 7,S-in.

graphite region outside t h e annulus.

The core composition was determined

by an i t e r a t i v e procedure t o achieve equilibrium with respect t o a 2 ft3/day f u e l processing rate.

'

'Results

The r e s u l t s are summarized i n Table M.2. t h e sum of t h e inventory and replacement costs.

Here t h e materials cost i s The leakage includes a l l

neutrons which escape from t h e system, are captured i n t h e vessel, o r are captured i n t h e reflector. Table M.2. Materials Cost and Nuclear C h a r a c t e r i s t i c s of MSCR as a Function of Reflector Condition

.... .

Reflect o r

Case

Materials Cost (mills/kwhr)a

Conversion Ratio

Leakage

%

1

None

0.783

0 828

2.62

2

Outside annulus

0.740

. .

0 849

logo

3

Between core and annulus

0,805

0.812

3.16

4

Between core and annulus

0 784

0,823

2.81

plus outside annulus.. .. .

.

These r e s u l t s show t h a t t h e r e f l e c t o r outside of t h e annulus improves performance s l i g h t l y , but t h a t t h e o t h e r r e f l e c t e d r e a c t o r s have poorer performanceo This behavior can be c l a r i f i e d by examining t h e power density d i s t r i b u t i o n s shown i n Fig, M.1. The r e a c t o r with t h e r e f l e c t o r outside t h e annulus shows a l a r g e peak i n power i n t h e annulus because of t h e l a r g e thermal f l u x from t h e r e f l e c t o r .


c 0-LR-DWG.

Fig. l a Unreflected Case 1

-

0

40

80

Radius ( i n . )

120

40

80

120 160 200

160

200

240

280 320 360

240

280 320 360

Radius ( i n . )

Fig. lb

0

69530

Fin. IC

Fig. I d

240

Radius ( i n . )

Fig. M . l .

280 320

360

0

40

80

120

160

200

Radius ( i n . )

Power Density Distribution in MSCR.


288

Since a peak i n t h e f i s s i o n rate occurs, t h e source of neutrons aimed out of t h e r e a c t o r is increased; however, t h e r e f l e c t o r r e t u r n s many of t h e The n e t effect, as shown i n Table M.2,

neutrons leaving t h e annulus.

is a

s l i g h t reduction i n leakage, The r e a c t o r with t h e r e f l e c t o r between t h e core and annulus experiences a considerable f l a t t e n i n g of t h e power d i s t r i b u t i o n .

The f i s s i o n

r a t e i n t h e annulus remains l a r g e and furnishes a l a r g e source of neutrons adjacent t o t h e reactor periphery.

This source is l a r g e r than f o r t h e un-

r e f l e c t e d case, and a higher leakage results. The r e a c t o r with t h e graphite regions on each s i d e of t h e annulus combines t h e bad f e a t u r e s of cases 2 and 3 0 The f i s s i o n rate is l a r g e i n t h e annulus,' giving a l a r g e source for neutrons t o t h e r e a c t o r periphery. Thus it is seen t h a t t h e main reason t h a t a r e f l e c t o r has such a small effect is t h e presence of t h e f u e l annulus. Any addition of reflect o r increases t h e f i s s i o n rate i n t h i s region.

These f i s s i o n neutrons are

close t o t h e r e a c t o r periphery, where they may be absorbed i n t h e v e s s e l

or leak out of t h e reactor.

A r e f l e c t o r would be much more b e n e f i c i a l i f

a design which eliminated t h e f u e l annulus could be devised.

Conclusions Use of a r e f l e c t o r i n t h e MSCR improved t h e r e a c t o r performance only

s l i g h t l y L0.02 increase i n conversion r a t i o and 0.04 mills/kwhr (electrical) decrease i n f u e l cycle costs].

a

O f t h e reflected-reactor configurations

considered, t h e r e a c t o r with t h e r e f l e c t o r outside of t h e f u e l annulus was t h e only one which improved performance.

However, a method of pinning

graphite t o t h e v e s s e l without leaving l a r g e cracks is unknown.

Also, t h e

extra cost of f a b r i c a t i n g an INOR-8 vessel 2.5 ft l a r g e r is unknown. Therefore, i n view of t h e s l i g h t b e n e f i t t o be gained and t h e added design u n c e r t a i n t i e s and complexity, it appears t h a t no r e f l e c t o r should be used i n t h e MSCR study.

If t h e s e u n c e r t a i n t i e s should be removed and s l i g h t

improvements i n performance become important, t h e gains a v a i l a b l e with a r e f l e c t o r should be exploited.

-

W


289 Appendix N DETAILED ESTIMATE OF 1000 Mwe MSCR CAPITAL INVESTMENT*

C, H, Hatstat** Summary The cycle chosen f o r t h i s analysis is shown i n elementary form i n Fig. 4.3,

A 2500 M w t r e a c t o r is cooled with a fuel-bearing molten s a l t ,

from which heat is t r a n s f e r r e d t o an i n e r t s a l t i n e i g h t v e r t i c a l s h e l l and-tube heat exchangers; t h e heat i n t h e i n e r t s a l t is removed i n a system of 16 shell-and-tube superheaters and e i g h t reheaters, of a design similar t o t h a t of t h e superheaters. Approximately 63% of t h e superheated

steam flows t o a system of four Loeffler b o i l e r s , where it produces satur a t e d steam by mixing with t h e turbine feed water,. The remaining superheated steam is delivered a t 2400 p s i , 1000째F, t o t h e t h r o t t l e of t h e steam turbine.

Exhaust steam f r o m t h e high-pressure turbine elements is reheated

t o 1000째F i n t h e r e h e a t e r s and flows t o t h e intermediate pressure t u r b i n e ,

from which it flows t o a 6-flow low-pressure unit.

Condensate i s returned

t o t h e Loeffler b o i l e r s through e i g h t s t a g e s of feed-water heating.

gross power output of t h e cycle is 1083 Mwe approximately 8 x lo6 l b / h r and a condenser The p l a n t design is based on an Atomic site i n Western Massachusetts, The site i s

The

a t a t h r o t t l e steam flow of pressure of 1.5-in.

Hg.

Energy Commission reference assumed t o have an adequate

source of c i r c u l a t i n g water for t h e turbine,

Because of t h e low vapor

pressure of t h e r e a c t o r coolant, high-pressure containment is not considered necessary; t h e r e a c t o r and i t s auxiliaries are contained i n a sealed,

s t e e l l i n e d concrete structure which forms

ai

p a r t of a subdivided biolo-

g i c a l s h i e l d with a t o t a l thickness of 10 feet.

The turbine-generator and t h e o t h e r components of t h e steam-condensate system are housed i n a conventional s t e e l frame building. The t u r b i n e

*&-bracted from SL-1554, SL-19?4. *%argent and Lundy, Engineers, Chicago, I l l i n o i s .


290

w-

building and t h e r e a c t o r building are arranged so t h a t one t r a v e l i n g bridge crane services both buildings. Other s t r u c t u r e s on t h e s i t e which are included i n t h e c o s t estimate are t h e crib house, c i r c u l a t i n g water intake and discharge flumes and tunnels, waste disposal building and stack, and foundations for o i l and condensate tanks,

Road and r a i l access are a l s o provided f o r t h e plant.

The c o s t estimate includes a l l systems and components necessary for

a complete plant. In addition t o t h e energy conversion components, t h e following equipment and/or systems are estimated i n d e t a i l . 1.

Radioactive waste treatment and disposal systems and building.

2.

Cover gas supply and d i s t r i b u t i o n system,

3.

Reagent gas supply and disposal system.

4.

7.

UF4 addition f a c i l i t y , Fuel s a l t handling, sampling and storage systems. Reactor vessel and primary pumps. Thermal s h i e l d and cooling system.

8.

Emergency shutdown cooling system.

9.

Reactor c o n t r o l system.

5.

6.

10.

Fuel s a l t chemical treatment system.

11.

Intermediate salt chemical treatment system.

12.

Intermediate s a l t handling, sampling and storage systems.

13.

Coolant pump l u b r i c a t i n g o i l systems.

14.

Hot sampling f a c i l i t i e s ,

15

Remote maintenance f a c i l i t y .

16 Subdivided shielded areas â&#x201A;Źor r e a c t o r a u x i l i a r i e s i n t h e r e a c t o r building.

d

Investment Requirements The c a p i t a l investment required for t h e concept which i s described i n t h i s r e p o r t has been estimated on t h e b a s i s of preliminary design and

material q u a n t i t i e s prepared by O a k Ridge National Laboratory and Sargent f Lundy.

The estimating d a t a for t h e heat cycle, a u x i l i a r y systems, and

primary and intermediate system components were prepared by Oak Ridge

P

h,

.


291

National Laboratory.

The cost estimate was prepared by Sargent f Lundy,

u s i n g accounting procedures s p e c i f i e d by t h e U. S. Atomic Energy Commission i n t h e Guide t o Nuclear Power Cost Evaluation.

The d i r e c t c o n s t r u c t i o n cost and i n d i r e c t c o s t are summarized below. The d e t a i l e d estimate is presented on subsequent pages. Estimated Direct Construction Cost I n d i r e c t Costs T o t a l C a p i t a l Investment for S t r u c t u r e s and Equipment Coolant S a l t Inventory p l u s 1,5% f o r Interest During Construction

Total Investment, Excluding Fuel S a l t

\

ti

. 3

$ 89,341,200

48,582,600 $137,923,800 10,951,800

$148,875,600


Table N . l . 1000 Mwe Molten Salt Converter Reactor Plant -Estimate of Capital Investment ONE (1) 2500 MWt MOLTEN SALT REACTOR ONE (1) 1000 MWe REHEAT TURBINE GENFZATOR UNIT C.C.6F 40" L.S.B. 1000째F 1000째F) (2400 Psi.

-

-

(Prices as o f 11-1-62 and Based on a 40 Hour Work Week) QUANTITY

-

ACCOUNT 21 STRUCTURES & IMPROVEMENTS 211 Ground Improvements Access Roads for Permanent .1 Use .ll Grading 1 .12 Surfacing 1 .13 Culverts .14 Bridges & Trestles ) .15 Guards & Signs .1.6 Lighting 1 General Yard Improvements .2 .21 Grading & Landscaping Roads Sidewalks & Parking .22 Areas .23 Retaining Walls, Fences & Railings .231 Fence, Post, Gates

MATERIAL OR EQUIPMENT

LABOR

TOTALS

In Place

15 Miles

>

c1

$6,000

$19,200

$25,200

47,000 SF

16,500

7,600

24,100

2,450 LF

8,500

3,200

11,700

Lot


Table N . l (continued)

QUANTITY

-

ACCOUNT 21 STRUCTURES & TMPROVEMENTS (Cont’d.) 211 Ground Improvements (Cont’d.) .2 General Yard Improvements (Cont’d.) .24 Outside Water Distribution Systems Including Fire Hydrants &Water Tanks for General Use .241 Domestic Water System ,2411 500 G.P.M. Deep Wells,) Including Pump & 1 Accessories 1 .2412 Storage Tank, 300 Gal.) & Controls 1 .2413 Water Softener, 1 Piping & Controls 1 .2414 Piping 1 .242 Fire Protection System .2421 Water Storage Tank .2422 2000 GPM Fire Pump 6 Motor Drive .2423 Other Fire Protection Equipment .2424 Piping, Including Hydrants .2425 Hose & Hose Houses

Lot

MATERIAL OR EQUIPMENT

LABOR

TOTALS

13,000

17,600

30,600

27,500

27,500

55,000

) )

1 )

1 1 1 )

Lot


Table N. 1 (continued)

QUANTITY

TOTALS

EQUIPMENT

-

STRUCTURES & IMPROVEMENTS (Cont'd.) 211 Ground Immmvements (Cont'd.) .2 General Yard Impravements (Cont d ) Sewers & Drainage Systems: .25 .251 Yard Drainage & Culverts .252 Sanitary Sewer System .2521 Septic Tank .2522 Dosing Syphon .2523 Distribution Box .2524 T i l e Field (Drainage) )

ACCOUNT 2 1

LABOR

MATERIAL OR

.

,253 Storm Sewer System: ) ,2531 Excavation & Backfill .2532 V i t r i f i e d Clay T i l e (6'' & 8") .2533 Reinforced Concrete Pipe ) (27" & 30") .2534 Manholes .2535 Outfall Structure .26 Roadway & General Lighting .261 Security Fence Lighting) 1 .262 Roadway Lighting .263 Parkway Cable .264 Trenching f o r Parkway ) Cable

I

h

Lot

4 000

7 000

11,000

Lot

$12,000

18,400

30,400

h,

0 &

Lot

13,000

11 200

24,200

Lot

8 000

,

11,200

19 ,200

4

e. L


c Table N. 1 (continued)

QUANTITY

-

ACCOUNT 21 STRUCTURES & IMPROVEMENTS (Cont'd.) 211 Ground Improvements (Cont'd.) .3 Railroads .31 Off Site .3l1 Grading .312 Bridges, Culverts & Trestles) 5 Miles .313 Ballast & Track 1 .314 Signals & Interlocks .32 On Site .321 Ballast & Track 265 LF TOTAL ACCOUNT 211

MATERIAL OR EQUIPMENT

LABOR

TOTAU

135,000

132,000

267,000

1,500 $245,000

1,600 $256,500

3,100 $501,500

11,500 45,200 11,400 4,400 75,000

450,400

212 Buildings

.1 .ll .12 .13 .14 .15

Excavation & Backfill Earth Excavation Rock Excavation Backfill Disposal Dewatering

11,500 CY 5,650 CY 6,350 CY 10,800 CY Lot

2,000

-

11,500 45,200 9,400 4,400 75,000

.3 .31 .32 .33 .34 .35 .36

Substructure Concrete Forms Reinforcing Concrete ) 6,750 CY Waterproofing Patch & Finish ) Conc. Miscellaneous Anchor Bolts,) Sleeves Etc. Embedded in Concrete

232,000

218,400

-


Table N. 1 (continued)

QUANTITY

-

ACCOUNT 21 STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildinns (Cont'd.) 212A Turbine Generator Building Including Off ices Control Roam, Cable Ram. Switch Gear Room(Cont'd.) Superstructure .4 .41 Superstructure Concrete .411 Forms 1 34,000 SF ) .412 Reinforcing of Floor .413 Concrete 1 .42 Structural Steel & Miscellaneous Metal 1,650 T .421 Structural Steel .422 Stairs, Ladders, Railings, Walkways, Gratings, Etc. Lot .43 Exterior Wall s .431 Masonry 56,400 SF .432 Sasulated Metal Siding Roofing & Flashing .44 .441 Pre-Cast Roof Slabs ) .442 Built-up Roofing & ) Flashing 1 35,600 SF .443 Poured Concrete Roof Deck 1 .444 Insulation 1

-

MATERIAL OR EQUIPMENT

LABOR

TOTALS

$57,500

$49,000

$106,500

535,000

128,000

663,000

55,000

79,000

-

21,000

134,000

46,400

-

180i400

32,000

36 ,000

68,000

Iu 0 \D


C

Table N. 1

QUANTITY

-

ACCOUNT. 21 STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212A Turbine Generator Buildinq Including Office, Control Room, Cable Room. Switch Gear Roam .4 Superstructure (Cont'd.) .45 Interior Masonry & Partit ions .451 Structural Tile 29,800 SP .46 Doors & Windows -461 Doors Lot .462 Windows 12,600 SF .47 Wall and Ceiling Finish .471 Glazed Tile 1 .472 Metal Ceiling 1 6,200 SF -473 Plastering Including ) Lathing and Furring ) A 7 4 Acoustical Ttle 1 .48 Floor Finish ) .481 Cement Lot .482 Tile 1 .49 Painting Glazing and Insulation .491 Painting Lot ,492 Glass and Glazing

-

.5

Stack (Heating Boiler and Auxiliary Boiler )

1

(continued)

MATERIAL OR EQUIPMENT

LABOR

TOTAL

$15,100

$20,300

$35,400

11 ,500 48 ,000

4,400 20,000

15,900 68 ,000

5,000

4,400

9,400

30,000

38, loo

68,100

10,500

32,400

42,900 Incl. 462

4,000

1 ,600

5,600

-

-


Table N. 1 (continued)

QUANTITY

-

ACCOUNT 21 STRUCTURES & I M P R O V ~ ~ ( C o n t ' d . ) 212 Buildings (Cont Id.) 212A Turbine Generator Building Including Office Control Room, Cable Roam, Switch G e q Room (Cont'd.) Building Services .6 Plumbing & Drainage Systems ) .61 ,611 Plumbing ,612 Drainage 1 Lot .613 Duplex Sump Pump 1 .614 Domestic Cold Water Tank 1 .615 Damestic Hot Water Tank 1 .62 Heating Boiler & Accessoriea .621 Heating Boiler ) .622 Unit Heaters .623 Discharge Ducts 1 .624 Condensate Pump & Receiver ) .625 Flash Tank Lot 1 .626 Pip ing ) ,627 Fuel Oil Transfer Pump .628 Heating Oil Tanks Day ) & Storage 1 .6221 Berm for Fuel Oil Storage j Tank ) .6222 Foundation for Heating Oil ) Day Tank 1 .63 Ventilating System .64 Air-conditioning System 1 .641 Air-conditioning Control Roam Lot .642 Office Air-Condit ioning 1 .643 Laboratory Air Conditioning

LABOR

TOTALS

$60,000

$32,000

$92,000

77,000

50,400

127,400

MATERIAL OR EQUIPMENT

-

(i

55,000

28, ooo

83 ,000

c.i


Table N. 1

QUANTITY

-

ACCOUNT 2 1 STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 21% Turbine Generator Building Including Office, Control loom, Cable Room, Switch Gear Room (Cont'd.) .6 Building Services (Cont'd.) .66 Lighting & Service Wiring .661 Control Panels & Cabinets 1 .662 Condui 1 .663 Wiring 1 .664 F i x t u r e s Switches & Receptacles ) .67 F i r e P r o t e c t i o n System (Water Lines, Hose, Sprinkler, E t c .) TOTAL ACCOUNT 212A

212D Waste .1 .ll .111 .112 .12 ,121 .13 .131 .132 .15 .151 .3 .31 .32

Disposal Building Excavation and Backfill Excavation Earth Rock Backfill Earth Disposal Earth ) Rock ) Dewatering Pumping Substructure Concrete Including, Forms, Anchored S t e e l Bottom Slab ) Walls t o Grade )

(continued)

MATERIAL OR EQUIPMENT

TOTALS

LABOR

Lot

$46,500

$39,200

$85,700

Lot

12,000 $1,422,100

2,400 $920,500

14,400 $2,342,600

a5

100

- 100

45 c.y.

100

100

.50 c.y.

50

50

1,000

1,000

3,200

6,700

-

C.Y.

.

Lot

100 c.y.

3,500

-

i u

3

I


Table N . l

(continued)

QUANTITY

-

ACCOUNT 21 STRUCTURES & IMPROVEMENTS (Cont'd.) 212D Waste Disposal Building (Cont'd.) 4 Superstruc ture .42 Structural Steel and Miscellaneous Steel 105 Tons .421 Structural Steel and Girts .422 Miscellaneous Steel Galleries Stairs, Landiilg, Handrailing, Lot Ladders, Etc. .43 Exterior Walls 6,800 s.f. .431 Insulated Metal Siding .44 Floors, Barrierc, Including Reinforcing, Forms, Etc. 135 c.y. .441 Walls Above Grade 80 c.y. .442 Floors 2,200 8.f. ,4421 Pre-Cast Roof Slab .45 Interior Masonry and Partitions .46 Doors and Windows Lot .461 Doors 1,200 s.f. .462 Windows 3,500 s.f. .48 Floor Finish (Cement) .49 Exterior and Interior Finishes .491 Painting Floor and Walls .492 Painting Structural and) Miscellaneous Steel Lot .493 Heavy Duty Coating 1 .494 Exterior Coating Below Grade

C)

MATERIAL OR EQUIPMENT

LABOR

TOTALS

$34 ,000

$8,000

$42,000

4j 500

2,000

6 ,500

13,600

4,900

18,500

w

0 0

6 ,500 3,200 2,000

4,000 1,800 2,000

10,500 5,000 4,000

2 ,000 4,500 500

1,000 2,000 1 ,500

3 ,000 6,500 2,000

25 ,000

20,000

45,000


Table N. 1 (continued)

QUANTITY

~ T E R I A Z .OR

LABOR

TOTALS

EQUIPMENT

-

ACCOUNT 21 STRUCTURES & IMPROVEMENTS (Cont'd.) 212D Waste Disposal Building (Cont'd .I .6 Building Services Plumbing and Drainage

.67

Conduit Fire-Protection System TOTAL ACCOUNT 212D

212F Miscellaneous Structures .1 Gate House .2 Electrical .3 Waste Storage Pond T ACCOUNT 212F 212G Reactor Plant Building .1 Excavation & Backfill .ll Earth Excavation .12 Rock Excavation .13 Backfill .14 Disposal .15 Dewatering

Lot

$4,800

$2,200

$7,000

Lot Lot

1,200 4,000 $109,300

1 800 1 Io00 $56,650

3,000 5;000 $165,950

Lot Lot Each

$5,500 3,000 2 ,800 $11,300

$5,200 2,800 5,200 $13,200

$10,700 5 ,800 8 000 $24 ,500

$5,700 21,800 1,200 2,400 55 ,000

$5,700 21,800 1,200 2,400 55,000

5,655 c.y. 1,090 c.y. 755 c.y. 5,990 Cay. Lot

CJ

8


Table N. 1 (continued)

QUANTITY

-

STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212G Reactor P l a n t Building (Cont'd.) Substructure Concrete .3 Forms .31 Reinforcing 1 .32 Concrete 1 .33 Waterproof ing .34 Patch & Finish .35 Miscellaneous Anchor Bolts, ) .36 Sleeves Etc. Embedded i n ) Concrete Superstructure .4 Superstructure Concrete .41 .411 Forms 1 .412 Re inf orc l n g 1 ) .413 Concrete I n t e r i o r Structural Steel & .42 Miscellaneous S t e e l .421 S t r u c t u r a l S t e e l & Reactor Supports .422 S t a i r s , Ladders, Railings, Walkways, Grating, E t c . Exterior Wa 11s .43 .431 Masonry ,432 Insulated Metal Siding .433 Concrete Walls

ACCOUNT 21

c',

5,730 c.y.

MATERIAL OR EQUIPMENT

$200,000

LABOR

TOTALS

$185,000

$385,000 w

8 6,981 c.y.

365,000

225,000

590,000

964 T

305,000

80,000

385,000

30,000

14,500

44,500

Lot

54,100 s.f. 5,150 c.y.

-

-

-

110,000 250,000

38,000 157,000

148,000 407,000


c

c Table N. 1 (continued)

QUANTITY

-

ACCOUNT 21 STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212G Reactor Plant Building (Cont'd.) .4 Superstructure (Cont'd.) .44 Roofing & Flashing .441 Pre-Cast Roof Slabs 1 .442 Built-up Roofing & Flashing) .443 Insulation .45 Interior Ma .451 Structural Tile .453 Hot Cells .46 Doors & Windows .461 Doors .462 Windows .48 Floor Finish .481 Cement .49 Painting Glazing Insulation ,491 Painting .492 Glass and Glazing .493 Insulation of Reactor Chamber .5 Stack (When Supported on Building) .6 Building Services .61 Plumbing & Drainage System .611 Plumbing ) .612 Drainage ) .613 Sump Pump )

23,300 s . f .

Lot

MATERIAL OR EQUIPMENT

LABOR

$21,000

$21,600

$42,600

-

-

-

400,000

80,000

TOTALS

480,000

w w

0

Lot 7,250 s.f.

2,500 27,000

1,200 12,000

3.700 39;OOO

35,000 s.f.

12,000

14,400

26,400

7,000 18,500 Included .462

25,500

Lot

Incl. in Acct. 221.32 Incl. in Acct. 212A

Lot

15,000

8,000

23,000


Table N . l

(continued)

QUANTITY

MATERIAL OR EQUIPMENT

LABOR

Lot

$130,000

$70,000

$200,000

Lot

17,000

19,600

36 ,600

Lot

8 500 $1 ,900,000

1 500 $1,032,400

10 000 $2,932,400

$3,442,700

$2,022,750

$5,465,450

15,000

16,000

31,000

$15,000

$16,000

$31 ,000

$15,000

$16 ,000

$31,000

$3,702,700

$2,295,250

$5,997,950

TOTALS

-

ACCOUNT 21 STRUCTURES 61 IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212G Reactor Plant Building (Cont'd.) .6 Buildings (Cont'd.) .62 Cooling System ) .63 Ventilating System ) .66 Lighting & Service .661 Control Panels & ) Cabinet 1 .662 Conduit 1 .663 Wiring 1 .664 Fixtures, Switches ) & Receptacles 1 .67 Fire Protection System (Water Lines, Hose, Sprinkler, Etc .) TOTAL ACCOUNT 212G TOTAL ACCOUNT 212 218 Stacks 218A Concrete Chimney .1 Excavation and Backfill ) .2 Substructure Concrete .4 Concrete Chimney 1 .6 Obstruction Lighting ) TOTAL ACCOUNT 218A TOTAL ACCOUNT 218 TOTAL ACCOUNT 21

1

Âś

Y

I.


Table N . l

(continued)

QUANTITY

ACCOUNT 22 221

LABOR

Lot

$15,000

$8,000

$23,000

Lot

7,540,000

560,000

8,100,000

37,500 7,000

Included 6,400

37,500 13,400

6,200 42,400 1,200 300

1QO 800 100 100

6,300 43,200 1,300 400

75,000 16,000 200 2,400 200 2,600 500 25,000 I n c l . Account 228

91,000 2,600 2,800 25,500

- REACTOR PLANT EQUIPMENT

Reactor Equipment .1 Reactor Vessel and Supports .ll Reactor Vessel Supports .12 Vessel and I n t e r n a l s ) .13 Pump Suction Columns ) .14 Graphite Rods 1 .15 Heaters .16 Insulation .2 Reactor Controls -21 Reactor Control S a l t Addition Tank .22 Reactivity Control Drain Tanks .23 Drain Tank Condenser .231 Condensate Pump .24 BF3 I n j e c t i o n System .241 BF3 Cylinders .25 Piping, Valves, Etc. .3 Reactor Shielding .31 Thermal Shield System .311 Thermal Shield & Supports ,312 Surge Tank, 2000 Gal. .313 Circulating Pumps .314 Heat Exchanger .315 L'iping, Valves, and I n s u l a t i o n .32 Biological Shielding Insulation, Shield Plugs, Etc. .34 Shield Cooling System .341 Closed Loop Liquid System ,3411 Shield Cooling Heat Exchanger (4000 F t . Surface Admiralty)

-

-

TOTALS

MATERIAL OR EQUIPMENT

cJ

g

Not Included I n c l . Account 228 1 1 2 1 Lot 1

Lot

2

255,000

124,000

379,000

35,000

2,500

37,500

I

I

.

,

,,,, ,

,,,.,,,,,

,

,

,,,, ~

P

,

-.

,

*_*I


.

. ... . .

.~

~

.

. . . _. . . . . . .

.

,

"

. .

.

-.

.

..

.

Table N. 1 (continued)

QUANTITY

-

ACCOUNT 22 REACTOR PLANT EQUIPMENT (Cont'd.) 221 Reactor Equipment (Cont 'd .) Reactor Shields (Cont'd.) .J .34 Shield Cooling System (Cont'd.) .341 Closed Loop Liquid System (Cont ' d .) .3412 Shield Cooling Circulating Pumps & Motors (2500 GPM 3 75 HP Motor) Lot .3413 Piping & Valves .3414 Cooling Coils Embedded in Concrete (16000 Ft. Lot 1" Steel) .3415 H20 Storage Tank 1 3000 Gallons Reactor Plant Cranes &Hoists .7 TOTAL ACCOUNT 221 222 Heat Transfer Systems Reactor Coolant Systems .1 Reactor Salt Circulating Pumps .ll 9075 GPM Including 1600 HP Motors .12 Reactor Salt Piping Insulation .13 Intermediate Coolant System .2 .21 Pumps Including Supports .211 Coolant Salt Pumps 13,900 GPM Including 2,000 HP Motors .212 Auxiliary Pumps and Drives .213 Insulation

C)

-

LABOR

$7,500 800 Included in Account 228 17,500

32,500

1,200 300 Included in Account 251 $8,070,800 $752,500

TOTALS

8,300

50,000

25,000 20,000 5,600

2,793,000 235,000 10,600

8

3,640,000

30,000

3,670,000

-

tn

$8,823,300

2,768,000 215,000 5,000

-

w

0

1,500

8 Lot Lot

-

MATERIAL OR EQUIPMENT

-

Included in Account 228

c.s


Table N. 1 (continued)

QUANTITY

MATERIAL OR EQUIPMENT

LABOR

$1,925,000 50,000

172,000 48,000

$2,097,000 98,000

2

70,000

12,800

82,800

8

2,320,000

28,000

2,348,000

4 16 8

4,000,000 6,350,000 1,400,000

160,000 48,000 10,000

4,160,000 6,398,000 1,410,000

1

60,000

5,000

65,000

-

ACCOUNT 22 REACTOR PLANT EQUIPMENT (Cont'd.) 222 Heat Transfer Systems (Cont'd.) .2 Intermediate Coolant System (Cont'd.) .22 Intermediate Coolant Piping and Valves .221 Pipe, Valves, Supports, E t c . .222 I n s u l a t i o n .223 Coolant S a l t Drain Tanks Including Heaters and I n s u l a t i o n ,23 Primary Heat Exchangers & Supports .3 Steam Generators Superheaters C Reheaters .31 Loeffler Boilers .32 Superheaters .322 Steam Reheaters .35 Auxiliary Start-up Boiler (300 P s i . 50,000 l b l h r . o i l Fired) .36 I n s u l a t i o n f o r Above Equipment .4 Reactor Coolant Receiving Supply and Treatment .411 F e r t i l e S a l t Addition System .4111 F e r t i l e S a l t Addition Tank .42 Reagent Gas System -421 H2 Supply .422 HF Supply .423 Piping .431 Reactor S a l t P u r i f i c a t i o n System .4311 Chemical Treatment Tank .4312 F e r t i l e S a l t Storage Tank

Lot Lot

TOTALS

a

Included i n Account 228

1

1,200

50

1,250

Not Included Not Inc luded Included i n Account 228

1 1

46,600 27,500

400 300

47,000 27,800

0

i l


Table N. 1 (continued)

QUANTITY

MATERIAL OR EQUIPMENT

Lot

LABOR

TOTALS

$67 500

$1,600

$69,100

1

19,200

200

19,400

1

30 000

1,000

31 ,000

1

3 ,000

500

3,500

1

30,000

1 ,000

31 000

-

ACCOUNT 22 REACTOR PLANT EQUIPMENT (Cont'd.) 222 Heat Transfer Systems (Cont'd.) .4 Reactor Coolant Receiving Supply and Treatment (Cont * d .) ,431 Reactor Salt Purification Sy stem (Cont ' d ) .4313 Radioactive Salt Sampler .432 Intermediate Salt Purification System .4321 Chemical Treatment Tank .44 Reactor Salt Charge System .441 Reactor Salt Preparation System .4411 Salt Melt Tank and Appurtenances .4412 UP4 Addition System .442 Intermediate Salt Charge System .4421 Intermediate Salt Preparation Tank and Appurtenances

.

b

b


Table N. 1 (continued)

QUANTITY

-

LABOR

TOTALS

.

ACCOUNT 22 REACTOR PLANT EQUIPMENT (Cont'd .) 222 Heat Transfer Systems (Cont'd.) .4 Reactor Coolant Receiving Supply and Treatment (Cont 'd .) .45 Cover Gas Supply and Purification System .4531 Dryer .4532 Heater .4533 02 Removal Unit .4534 Coolers A535 Pure He Reservoir .4536 0 2 Analyzer .454 Piping .5 Intermediate Coolant Storage Tanks, Etc. TOTAL ACCOUNT 222 223 Fuel Handling and Storage Equipment .31 Fuel Salt Drain and Storage System .311 Drain Tanks & Cooling Jacket .312 Drain Tank Condenser .3121 Condensate Pump .313 Decay Storage Tank Including Cooling Jacket .314 Fuel Withdrawal Transfer Tank '

MATERIAL OR EQUIPMENT

200 50 400 100 3,000 200 200 100 400 100 7,000 500 Included Acct. 228

250 500 3,200 300 500 7,500

2

70,000 $23,039,200

2,000 $570,500

72,000 $23,609,700

54 1 2

$1,269,000 8,000 1,400

$16,800 300 100

$1,285,800 8,300 1,500

5

55,000 37,000

1

1,000 100

56,000 3,800

c*,

0 \c)


Table N. 1 (continued)

QUANTITY

TOTALS

LABOR

-

REACTOR PLANT EQUIPMENT (Cont'd.) 223 Fuel Handling and Storage Eauivment .321 Fuel Withdrawal Metering Tanks .33 Flush S a l t Storage Tanks .34 Piping TOTAL ACCOUNT 223

ACCOUNT 22

MATERIAL OR EQUIPMENT

225 Radioactive Waste Treatment & Disposal Liquid Waste Systerns .1 .ll High Level Storage Tank .111 H.L. Storage Tank Pump .112 H.L. Waste Evaporator .113 H.L. Waste Condenser .114 Evaporator Recycle Pump Demi 8 t e r .13 H.L. Concentrated Waste .14 Storage Tank KOH Scrubber .15 .151 KOH Make-up Tank & Pump H2 Burner .16 .161 H Burner Condenser PZp ing .17 Gas Waste .2 Stack Blower .21 .211 Absolute F i l t e r . .212 Roughing F i l t e r

1 5

$300 $6,500 $6,200 2,800 155,300 152.500 Included i n Account 228 $1,495,800 $21,400 $1,517,200

w

~~

P

12,400 400 2,000 600 400 600

2 1 1

600 100 100 50 100 50

1

13.000 500 2,100 650 500 650

2,500 150 12,100 400 1,400 100 225 75 1,000 100 Included i n Account 228

2,650 12,500 1,500 300 1,100

20,000 4,000 1,200

22,400 4,400 1,400

2,400 400 200

0

E:,


c

c Table N. 1 (continued) QUANTITY

ACCOUNT 22 REACTOR PLANT EQUIPMENT (Cont'd.) 225 Radioactive Waste Treatment & Disposal (Cont'd .) .22 H .F. Absorbers Including Charcoal .221 Absorber Coolers .222 Vacuum Pump .223 H.F. Absorber Vacuum Tank Air Cooled Absorbers Including Charcoal 1.5" Finned Tubes .232 3" Finned Tubes ,233 6" Finned Tubes .234 Dilution Air Duct & Dampers .24 Water Cooled Absorbers .241 112" Tubes .242 1" Tubes .243 1-112'' Tubes .244 2" Tubes .25 Decay Tank .26 Absorber Cooling Water Condenser .27 Helium Recycle Compressor .28 BF3 Stripper .281 Vacuum Pump .29 Piping TOTAL ACCOUNT 225

MATERIAL OR EQUIPMENT

LABOR

TOTALS

, -

226

Instrumentation and Controls .1 Reactor .2 Heat Transfer Systems

$16,000 100 500 300

$2,400 50 50 50

$18,400 150 5 50 350

3,600 8,000 15,000 8,000

400 600 800 3,230

4,000 8,600 15,800 11,200

23,200 60,.400 51,400 49,600 12,500

800 1,900 1,900 1,800 800

24,000 62,300 53,300 51,400 13,300

100 1,000 900 2,000 200 2,200 2,400 30,400 28,000 50 550 500 Included in Account 228 $338,825 $22,325 $361,150

$300,000 70,000

$170,000 50,000

$470,000 120,000

c-' CI, p


Table N. 1 (continued)

QUANTITY

MATERIAL OR EQUIPMENT

LABOR

TOTALS

Cr

-

ACCOUNT 22 REACTOR PLANT EQUIPMENT (Cont'd.) 226 Instrumentation and Controls (Cont'd.) .3 Service to Fuel Handling and Storage .4 Service to Radioactive Waste 6 Disposal .5 Radiation Monitoring .6 Steam Generators .7 Control & Instrument Piping & Wiring .8 Electrical Connections .9 Other Miscellaneous TOTAL ACCOUNT 226 .

227 Feed Water Supply and Treatment Raw Water Supply .1 Make-up Water Treatment .2 Evaporator .21 Ion Exchange Equipment, .22 Filters, Etc .23 Acid h Caustic Transf. Pumps & Drives .24 Demineralized Water Storage Tanks .25 Caustic Tank Acid Tank .26 .27 Foundation Piping & Valves .28 Insulation .29 Steam Generator Feed.3 Water Purification

.

$120,000

$80,000

50 000 60,000 80,000 120,000 Included in Account 235

$200,000 110,000 200 ,000

Included in Account 235 Included in Account 235 Included in Account 235 $670,000 $430,000 $1,100,000 Included in Account 211

1 Lot

1 Lot

$45,000

$10,000

$55,000

600

200

800

2 2 1 1 1 Lot 1 Lot

30,000 2,200 2 200 3,500 J

Included 30,000 400 2,600 400 2,600 6 ,000 2,500 Included in Account 228 Included in Account 228


*

t

.I

C

Table N. 1 (continued) QUANTITY ACCOUNT 22

227

LABOR

TOTALS

- REACTOR PLANT EQUIPMENT

(Cont'd.) Feed Water Supply and Treatment (Cont'd .) Feed-Water Heaters .4 .41 Deaerating H e a t e r s "E" 4,020,000#/Hr. 150 P s i g . .42 Closed H e a t e r s .421 L.P. Heater "A" .422 L .P Heater "B1* ,423 L .P. H e a t e r "C" .424 L .P Heater "D" .425 H.P. Heater "F" .426 H.P. Heater "G" .427 H .P. Heater "H" Feed-Water Pumps and Drives .5 Feed-Water Pumps & Drives .51 .511 6600 GPM Pumps 2465 P s i g Hd. .512 11,300H.P. B.F. Pump Turbine Drive 5600 RPM Motor Driven S t a r t - u p F.W. .52 Pump .521 6000 GPM Pump 850 P s i g . Hd. .522 3500 H.P. S t a r t - u p FW Pump Motor Heater "A" Drain Pumps and .53 Drives .531 620 GPM Pump 285 P s i g . Hd. ) .532 125 H.P. Heater "A" Drain Pump Motor 1

-

. .

I

MATERIAL OR EQUIPMENT

-

-

2

$240,000

$15,000

$255,000

3 3 3 3 3 3 3

75,000 63,000 63,000 81,000 315,000 429,000 441,000

5,000 5,000 3,000 3,000 3,000 3,000 3,000

80,000 68,000 66,000 84,000 318,000 432,000 444,000

3

405,000

12,000

417,000

3

750,000

30,000

780,000

1 1

70,000 55,000

3,000 2,000

73,000 57,000

3

22,500

1,500

24,000

$1


Table N. 1 (continued)

QUANTITY ACCOUNT 22 227

LABOR

TOTALS

- REACTOR PLANT EQUIPMENT

(Cont'd.) Feed Water Supply and Treatment (Cont ' d .) Feed-Water Pumps and .5 Drives (Cont'd.) Heater "C" Drain Pumps .54 and Drives .541 700 GPM Pumps 210 Psig. Hd.) .542 100 H.P. Heater "C" Drain ) Pump Motor 1 Boiler Steam C i r c u l a t o r s .55 and Drives .551 5,300,000 #/Hr. Steam Circulator 2500 1675'F .552 5000 H.P. Turbine Drive f o r Steam C i r c u l a t o r 500 # Steam; 10,000 RPM .553 5000 H.P. Motor f o r Steam C i r c u l a t o r Including Gear and Mag. Coup1ing TOTAL ACCOUNT 227

3

Steam, Condensate, Feed Water, and a l l Other Piping, Valves E t c . For Turbine P l a n t , Crib House and Other Reactor P l a n t A u x i l i a r i e s .1 Pipe, Valves, F i t t i n g s , E t c . .ll Turbine Plant 1 .12 Other I n t e r i o r Piping ) .13 Yard Pipe Etc. 1

$30,000

$2,500

$32,500

w

-

228

MATERLAL OR EQUIPMENT

97 5,000

40,000

1,015,000

560,000

30,000

590 ,.ooo

100,000 $4,758,000

7,000 $181,500

107,000 $4,939,500

$4,025,000

$2,575,000

$6,600,000

P .fc

-

C)

1 Lot

c:,


Table N. 1 (continued)

ACCOUNT 22 228

QUANTITY

MATERIAL OR EQUIPMENT

1 Lot 1 Lot

$455,000 155;OOO $4,635,000

-

REACTOR PUNT EQUIPMENT (Cont'd.) Steam, Condensate, Feed Water, and a l l Other Piping, Valves E t c . For Turbine P l a n t , Crib House and Other Reactor P l a n t A u x i l i a r i e s (Cont'd.) .2 Insulation .21 Piping I n s u l a t i o n .22 Equipment I n s u l a t i o n TOTAL ACCOUNT 228

LABOR

TOTALS

-

229

Other Reactor Plant Equipment .2 Remote Maintenance Facilities .4 Coolant Pump Lube O i l System Storage Tank ) .41 .42 Pumps 1 .43 O i l Coolers ) .44 Filters 1 .45 Piping .62 Intermediate S a l t Sampling System .621 Sampler and Appurtenances TOTAL ACCOUNT 229 TOTAL ACCOUNT 22

$520,000 195,000 $3,290,000

$975,000 350,000 $7,925,000

E

in Lot 1 2 2 2

3,000,000

Included

3,000,000

29,400

600

30,000

Lot Lot

Included i n Account 225 16,500 $3,045,900

2,000 $2,600

18,500 $3,048,500

$5,270,825

$51,324,350


Table N.l

(continued)

QUANTITY

MTEIUAL OR EJUIPMIWT

1 U;ic 1 Lot

$175,000 10,Ooo

-

ACCOUNT 23 TURBINE GDTEFtATOR UNITS 231 Turblne Generators * Turbine Foundations Concrete Including Reinforcing Steel, Eto Miscallaneoue Turbine Generators Turbine Generator Units As Follows8 1000 MWe Rehest W M n e Generator Unit C.C.6F. 40” L.S.B. Complete with Acceesories Steam C O n d i t i O n 8 2i@ P d 1000.F -1000°F Generators: 1,280,OOO KVA Total .85 P.F. and .64 SCR Accessoples Other Than Sttrndard Generator Exciter (Motor Driven) Reserve Exciter l”AL ACCOUNT 231

-

.

LABOR

TOTAU

-

-

-

232 Circulating Water System .1 -ping and Regulating Systana .11 m p s , Drives & Contlola .112 134,000 GPM Vertical Mixed Flow Circulating Water p\rmps Head 30 ft.

,t

Y

1

19,815, OoO Included i n Account 2 3 1 . a Included i n Account 231.21 Included i n Account 231.21

1

6

4

360,000

18,000

m,m


. I

*

h

c

Table N. 1 (continued)

-

TURKlNE GEIWUTOR UNITS (Canted.) 232 Circulating Mater System (Conttd.) -1 b p b g a d Regulating Systms (Contfd.) .ll Pumps, Drives & Contmls (Cont'd.) ,113 1250 H.P, Motor Drive f o r Circulating Water Pumps e12 Traveling Screens, Etc. ,121 Traveling Screens Complete with Kotors e 1 2 2 1200 GPM Scr 230 Ft. Discharge Head .123 100 H.P. Motor f o r Screen Wash hrmp ,124 Trash Rake Complete with Appurt enances .I25 Pipe & Valves e2 Circulating Water Lines .21 Supply Lines To Condenser .2ll Circulating Water Piping, Valves, Fittings, Etc. e 2 l l l S t e d Circulating Water Piping, Valves Expansion Joints, Fittings, Etc. e22 Discharge Lines From Condenaer ,221 Circulating Water Piping, Valves, Fittings, B c .

ACCOrDUI) 23

6

$ 8 0p 000

7

122,500

2

5,000

2

4, 500

1 1 Lot

-

-

1Lot

,

27 500


Table N. 1 (continued)

LABOR

QUANTITY

-

ACCOUNT 23 TUiZBINE GaJERA'KR UNITS (Cont'd.) 232 Circulating Water System (Cont'd.) -2 Circulating Water Lines (Cont'd.) .22 Discharge h e 8 From Condenser (Cont'd.) .2211 Steel Circulating Water Piping Valves, Expansion Joint 8 F i t t i n g s , Etc Intake and Discharge .3 S t ructuree 31 Intake Structures .311 River Dredging & Rock

-

.

.312 .3121 .3122 323 .313

Removal Intake Flume Intake Flume Proper Floating Born

Concrete Retaining Walls Intake Crib House 3131 Substructure e3132 Superstructure *3133 S t e d -3134 E l e c t r i c a l Work e32 Discharge .321 Seal Well & Discharge

Tunnel .322 Discharge Flume

Included i n Account P2.P

1Lot

$12,000

1 Lot 1 LQt 1 2

45,000 10,400 29,200

1 Lot

-

35 T 1 Lot Lot Lot

29,500

4,500

28,400 22 400

57,900 26,900


a

c

b

L

b'

i,

Te-le N. 1 (continued)

QUANTITn

LABOR

TOTALS

EQUIPMENT

-

TURBINE GENEWi'IUR UNITS (So3t'd.) 232 Circulating Water Systaf, (Coct'd.) .4 Fouling, Corrosion C o n t r o l

ACCOUNT 23

MATERIAL OR

and Water Treatmat

.W Chlorinating System .w1 Chlorination Equipcent 4 2 Chlorine Handling Facilities TOTAL ACCOUNT 232

1 Lot 1 Lot

233 Condmsers and Auxiliaries Condensers Foundation8 Condenser Shell and Appurtmances .121 225,000 Sq. Ft. Single Pass Condensers Complete with Appurtenances Including Shell, Water Boxes, Tube Sheets, Tube Supports, Hot Well, Mended Neck with Expansion Joint , Etc. .13 50 F t . Long Admiralty Condenser Tubes 2 7 Instruments & Accessories .2 Condensate Pumps .P Pwnm & Drives ,211 1875 aCondensate h p s Complete with Appurtenances, Mscharge Head 325 F t .

.1 .U .12

-

3

3 3

Sets

3

Sets

6

Included

15,000

6,000

93,000


Table N. 1 (continued)

LABm

QUANTIm ACCOUHI 23

- TURFSINE GENERATOR UNITS (Contpd.)

233 Condensers and A w d 2 i a r i e s (Cont'd.) Condensate Pump8 (Cont'd.) Pumps & Drlvea (Cont'd.) -212 400 H.P. Motors for Condensate Pumps -22 Suction Piping .3 A i r Renoval Equipment and Pipa -31 Steam Jet Air Ejector, with I n t e r & After Condensera .32 Air Suctlon Piping -33 PrinLing Ejectors T@TAL ACCOUIV" 233

.2

.a.

234

C e n t m 2 Lubricating System .1 Treating & -ping Esuiranent .2 Storage-Tanks & App&e-mcea

.3

Fire Protection TOTAL ACCOUNT 234

235 Turbine ?lant Boards Instrument8 & Controls .1 Control - p e n t .11 blechanicd. Contml Boards ) -12 Isolated Contmller, 1 Transmitters B c . 1 .2 Isolated Recording Gauges ) Meters & Instruments 1

6

F L I 200 $51,000 Included i.n Accoun, 233.121

Int

6

100,OOO

Lot Lot

1 Lot 1 Lot 1LOt

1 Lot

9,oo 1098 Included i n Account 228 Included

2,000

3.m Included Account

$275,000

19, Ooo 17.000

k? 5


Table N. 1

(continued)

QUAI lTITP

-

TURBINE GENERATOR UNITS (Contcd.) 235 !lâ&#x20AC;&#x2122;urbine Plant Boards Instruments & Controls .3 Control dc Instrument Piping & Tubing .4 Electrical Connectione m A L ACCOUIE 235

ACCOUNT 23

MATERIAL aR

LABOR

TOTAM

pi[xT=

-

1 Lot 1 Lot

$2o..ooo

236 Turbine Plant Piping .1 Main Steam Between Stor, Valves and Turbine I n l t .2

Included i n Account 231.2

Drip, Drain and V e n t Piping and Valves TOTAL 236

237 AwdliaryEqu i m e n t for GeneratoFB .1 Excitation Panels. Switches 13iiheostk,ts .2 Generator Cooling Water Systens .21 Lubricating O i l Cooling systan .22 Qazerator Hydrogen Cooling system .23 Generator Liquid cooling System

Included in Account 228 Included i n Account 228

IncludcZ i n Account 231.2

Lot

60,000

:.?,reo

72, OOO


Table N. 1 (continued)

QUANTITY

-

TOTALS

Other Turbine Plant Equipment .1 Gland Seal Water System .2 Vacuum Priming System TOTAL ACCOUNT 238

-

-

-

-

Lot

$50,000

$15 ,000

$65,000

$110,000

$27,000

$137,000

tInc luded i n Account 228 ----C

1

Included i n Account 228 e Included i n Account 228

TOTAL ACCOUNT 23 ACCOUNT 24

LABOR

EQUIPMENT

ACCOUNT 23 TURBINE GENERATORS UNITS (Contld.) 237 Auxiliary Equipment f o r Generators (Cont'd.) .3 Central Hydrogen Cooling Sy6fClB .4 F i r e Extinguishing Equipment) Including Piping and C02 ) System Exclusively f o r 1 Generators 1 .5 F i r e Extinguishing 1 Equipment f o r O i l Ram, E t c . ) TOTAL ACCOUNT 237

238

MATERIAL OR

f.

le6 403

$26 ,843,700

$4,000 1,600 800 19,200

$42,000 15,600 6 ,800 29 ,200

-

ACCESSORY ELECTRIC EQUIPMeNT 241 Switchgear .1 Generator Main and Neutral C i r c u it s .ll Generator P o t e n t i a l Transformer Compartment .12 Surge Protection Equipment .13 Generator Neutral Equipment .14 Miscellaneous Items

2 2

1 Lot

$38,000 14,000 6,000 . 10,000


.

1 1 1 1

.

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

I

.

............................ w

(1

I

._-.....

.I". . . . . . . . . . . . . .

4

Table N. 1

-

ACCOUNT 24 ACCESSORY ELECTRIC EQUIPMENT 241 Switchgear. (Cont'd.) .2 Station Service .21 13.8 KV Switchgear .22 4160 V. Switchgear .23 480 V. Switchgear TOTAL ACCOUNT 241

243 Protective Equipment .l General Station Grounding E qu ipme n t .2 Fire Protection System TOTAL ACCOUNT 243

. . 1 1 1 1 _

___...l__l ...............I .... ......

(continued)

QUA" ITY (Cont ' d .I

242 Switchboards .1 Main Cont .2 Auxiliary Power Battery & Signal Board .21 Battery & Battery Charging Pane1s .22 D.C. Control & Auxiliary Pane1s .23 A.C. Control & Instrument Panels .24 Motor Control Centers .25 Miscellaneous Panels & Boards TOTAL ACCOUNT 242

........................... ........ -1411-_1. . \

(.

MATERIAL OR EQUIPMENT

LABOR

TOTALS

365,000 110,000 $543 ,000

-

-

51.200

416,200

$82 ,000

$31,200

$113,200

1

15,000

5,600

20,600

2

18,000 7,000 80,000

4,800 1,600 13,600

22,800 8,600 93,600

Lot

16,000 $218,000

11,200 $68,000

27 I 200 $286,000

Lot Lot

$60,000 14,000 $74,000

$48,800 81800 $57,600

$108,800

Lot Lot Lot

Lot

1 Lot

=

.....


Y

Table N. 1 (continued)

QUANTITY ACCOUNT 24

LABOR

TOTALS

$21,600 72,000

$35,600 152,000

7 000

8,000

15,000

5,000 $106,000

5.600 $107,200

10.600 $213,200

Lot

$25,000

$61,200

$86,200

Lot

22, OoO

54,400

76,400

Lot

6 ,000

6,800

12.800

Lot

9,000 5.000 $67 ,000

15,200 5.600 $143,200

24.200

Lot

$576,000 110,000

$49,600 16,000

$625,600 126,000

EQUIPMENT

-

ACCESSORY FUCTRIC EQUIPMENT (Cont'd.) 244 E l e c t r i c a l Structures .1 Concrete Cable TUnnels, Compartmenta and Cable Trenches i n Earth Lot .2 Cable Traya 6 Supports 192,000 l b . P i p e and S t e e l Frames .3 and Supports Lot .4 Foundations & Pads f o r E l e c t r i c a l Equipment Lot TOTAL ACCOUNT 244

245 Conduit .1 Conduit .ll Power Conduit .12 Control and Instrument Conduit .2 Concrete Envelopes .21 10" Transita Pipe Duct Run .22 Iron Conduit Enclosed i n Concrete .3 Manholes & Covers TOTAL ACCOUNT 245 246

MATERIAL OR

Power and Control Wirinq .1 Main Power Cables and Bus Duct .ll Isolated Phase Bus Duct (Generator) .12 Main Power Cables

5

1

$14,000 80,000

,


c

C {continued)

Table N . 1

. QUANTITY ACCOUNT 25

- MISCELLANEOUS POWER PLANT

EQUIPMENT (Cont ' d ,) 251 Cranes and Hoisting Equipment (Cont'd.) .2 Miscellaneous Cranes and Hoists TOTAL ACCOUNT 251

252

253

Compressed A i r and Vacuum Cleaning System .1 Compressors and Accessories ,11 200 C.F.M. S t a t i o n A i r Compressore Including Motor Drives .12 250 C.F.M. Control A i r Compressors including Motors .13 A i r Drying Equipment f o r Control A i r System .14 Receivers ,141 S t a t i o n A i r .142 Control A i r .2 Pipe Valves and F i t t i n g s .3 Vacuum Cleaning System TOTAL ACCOUNT 252 Other Power P l a n t Equipment .l Local Communication, Signal and C a l l System ,2 F i r e Extinguishing Equipment .21 2000 GPM F i r e Pump Including Drive and Accessories .22 Other F i r e P r o t e c t i o n Equipment

Lot

2 2

Lot Lot

MATERIAL OR EQUIPMENT

TOTALS

$23,000 $173,000

$2,000 $22,000

$25,000 $195,000

$13,500

$1,200

$14,700

15,500

1,200

16,700

9,000

500

9,500

1,300 1,300 -Included

300 1,600 300 1,600 i n Account 228 4

16.ooo

4.(30Q

20,000

$50,000

$44,800

$94,800

$56,600

Lot

LABOR

$7,500

$64,100

Included i n Account 211.24 Lot

$19,000

$1,000

$20,000


Table N. 1

(continued)

QUANTITP

-

ACCOUNT 25 MISCELLANEOUS POWER PLANT (Cont'd.) 253 Other Power Plant Esuiranent (Cont'd.) Furniture and Fixtures .I Lockers, Shelves, and Catd.net8 .4 Cleaning Equipent 05 Machine Tools & Other Station .6 Maintenance Equipnent Laboratory, Test & .7 Weath er Instrmnents 71 Radiation Monitoring Equipment 72 Miscellaneous Laborat.org, Telst & Weather Instruments Diesel Generator Unit loo0 KW 09 Including Oil Tank TOTAL ACCOUNT 253

TOTAL ACCOUNT 25 TOTAL DIRECT CONSTRUCTION COST

INDIRECT COSTS

LABm

TOTAIS

Lot Lot Int Lot

240,000

$10,000

Lot

23,000

2,000

Lot

20,Ooo

1 897,300

$11,166,574

Contractor's O'HD and PmfMt 20% SUB-TOTAL

23 333,300 $91,674,500

mer&and Administrative 6.3% SUB-TOT&

5,775 9 5 0 0 $97,440 ,m

Miscellmneous Construction 3%

974,500 $98,424,50(?

SUB-TOTAL

C)

HATEFUL OR EQUIPMENT


Table N. 1 (continued)

QUANTXTY g G I N E E R I N G DESIGN AND INSECTION A E Design and Inspection SUB-TUTAL

-

MATEXAL OB EQUIPMENT

LABOR

TOTAM

U.lk

Nuclear Engineering 3 . s SUB-rnAL

start SUB-TOTAL Land and Land Right8

360,000 4bU-4,6 ~ ~ ~ 7 0 0

SUB-TUTAL Contingency l@ SUB-TOTAL Interest During Construction SUETCYl'AL

11,461,200

9.G

Fuel Charge Int ennediate Coolant Salt Investment, and 1.5% fop Interest During Construction

TOTAL CAPITAL INlUiSDENT

$126 ,072,900

Not Inc lwi ed l0,790,000 161,800

8U.8 875,600


!

328

zi.

Appendix 0 DESIGN REQUIREMENTS FOR THE MSCR MODERATOR

Lo

Go Alexander

Introduction The r e a c t o r vessel contains t h e moderator and holds it i n q s t a b l e position during a l l phases of operation, provides f o r accepting f u e l discharged from t h e heat exchangers, passage of f u e l through t h e moderator, and discharge t o t h e f u e l pumps, The reactor m u s t be designed t o expose t h e f u e l t o neutrons a t a spec i f i e d r a t i o of graphite volume t o f u e l volume, The graphite must be supported and r e s t r a i n e d under a l l circumstances, including t h e drained condition.

Allowance must be made for d i f f e r e n t i a l expansion between

graphite and vessel,

Provision must be made f o r d i s t r i b u t i n g t h e flow of

f u e l over t h e core entrance, and for c o l l e c t i n g t h e flow a t t h e e x i t ,

A

free surface i n an expansion chamber must be provided somewhere i n t h e f u e l c i r c u i t , and c i r c u l a t i o n through t h e expansion chamber must be maintained. Provision must be made for preheating t h e r e a c t o r v e s s e l p r i o r t o charging molten s a l t , and f o r cooling t h e reactor v e s s e l during operation and after shutdown. This cooling must be accomplished without t h e generation of excessive thermal stresses.

The v e s s e l must be designed i n conformance with

t h e pressure vessel codeo Means of sparging t h e f u e l i n t h e expansion chamber with an i n e r t gas must be provided t o remove xenon and o t h e r vola-' t i l e materials. Excessive thermal stress i n t h e g r a p h i t e must be avoided, and it must be composed of pieces s u f f i c i e n t l y small t h a t d i f f e r e n t i a l shrinkage due t o exposure t o neutrons w i l l be t o l e r a b l e i n each piece, Stagnation of t h e f u e l between adjacent blocks o f graphite or between graphite and metal s t r u c t u r e must be avoided i f such stagnation leads t o ' excessive temperatures or stresses i n e i t h e r t h e f u e l , graphite, or metal structure,

Temperature a t t h e fuel-graphite i n t e r f a c e should be below

t h a t a t which chemical reaction, i f any, takes place a t an appreciable

rate, and below t h e temperature a t which any important c o n s t i t u e n t of t h e \

,

t


329

s a l t (other than v o l a t i l e f i s s i o n products) has appreciable vapor pressure (e.g.,

i f UF4 were t o vaporize appreciably and d i f f u s e i n t o pores i n t h e graphite, t h i s would be disadvantageous and perhaps hazardous

Temperature gradients i n t h e graphite near stagnation areas should not exceed those corresponding t o t o l e r a b l e thermal stresseso I n t h e MSCR it is desirable, i n order t o reduce t h e graphite surface exposed t o permeation by s a l t and f i s s i o n product gases, t o use moderator elements of l a r g e r diameter than i n t h e MSRE.

On t h e o t h e r hand, s e t t i n g

an allowable thermal s t r a i n of 0.001 imposes an upper l i m i t ,

A diameter

of about s i x t o e i g h t inches appears t o be a s u i t a b l e compromise between t h e c o n f l i c t i n g requirements, It would be d e s i r a b l e f o r t h e logs t o extend t h e length of t h e core, but r a d i a t i o n damage may induce a tendency f o r t h e logs t o bow outward and t h i s would increase t h e volume f r a c t i o n of f u e l i n t h e core.

This

increase is controlled and l a r g e l y eliminated by using logs 24 in. long and stacking t h e s e i n a v e r t i c a l position, of pins and sockets,,

The ends are mated by means

D i f f e r e n t i a l shrinkage of t h e graphite is now accom-

modated by a s l i g h t r o t a t i o n about t h e pins. Fuel s a l t should permeate the graphite not more than O o l % by volume. With t h i s penetration, and 10 volume per cent of f u e l i n t h e core, about one per cent of t h e fuel w i l l be i n t h e pores i n t h e graphite.

This is

probably t o l e r a b l e , e s p e c i a l l y i f t h e accessible pores are those near t h e surface, as seems t o be t h e case, Fuel stagnation i n cracks between blocks and i n pores i n t h e blocks

is c l o s e l y r e l a t e d t o t h e problem of a f t e r h e a t , since the f u e l so involved i s probably not r e a d i l y drainable.

Means of flushing t h e fuel-

s a l t thus retained must be provided i f possible, and i f t h i s is not possible, means must be provided of removing t h e heat generated i n t h e core after drainage, Temperature r i g e and thermal stress i n t h e r e a c t o r v e s s e l must be limited t o t o l e r a b l e l e v e l s .

D i f f e r e n t i a l thermal expansion w i l l leave a

gap between t h e graphite s t r u c t u r e and the r e a c t o r v e s s e l not less than 1 in, i n the r a d i a l direction.

It w i l l be necessary t o provide some flow through t h i s annulus not only t o remove t h e heat generated t h e r e , but a l s o t o cool t h e r e a c t o r vessel,


: 330

n

It w i l l be necessary t o o r i f i c e the flow channels, or otherwise vary t h e i r width systematically i n order t o d i s t r i b u t e t h e flow through

w*

the core compatibly w i t h the power density d i s t r i b u t i o n t o achieve uniform teGerature r i s e i n a l l f u e l chaonels.

Moderator I n the &ERE (l), the moderator i s constructed of square graphite bars measuring 2 inches on the sides and 63 inches long.

Channels 0.4

ioches deep and 1.2 inches wide a r e machined i n t o a l t e r n a t e faces of t h e bars, which a r e pinned loosely t o beams lying across the bottom of the vessel.

The channels occupy 22.5% of the volume of the matrix.

The shape, size, and spacing of the MSRE moderator elements a r e not suitable f o r the MSCR.

The pieces need t o be larger and the volume of

f u e l needs t o be of the order of 10 percent. The void f r a c t i o n i n a matrix formed by closely packed cylinders of uniform diameter i s 0.093.

Such a matrix appears t o be s t r u c t u r a l l y

stable, i s e a s i l y assembled i n a close array, and provides a minimum of contact between individual pieces.

This l a s t i s important i n t h a t the

amount of stagnant f u e l i s probably roughly proportional t o the t o t a l area of contact.

Small v a r i a t i o n s i n radius, c i r c u l a r i t y , and s t r a i g h t -

ness of the cylinders can be tolerated, and a voidage a s l o w as lo$ s t i l l be achieved.

Higher voidages can be obtained by machining away portions

of the s w f a c e of the cylinders; i n f a c t , t h e voidage can be systemat i c a l l y varied both r a d i a l l y and a x i a l l y i n t h i s way.

The r e l a t i v e

positions of the logs a r e fixed by contact of the unmachined portions a t the top and bottom.

It should thus be possible t o reduce power peaking

i n the reactor and, by a combination of o r i f i c i n g and power f l a t t e n i n g , t o obtain a good match between radial d i s t r i b u t i o n s of flow and power density and thus achieve approximate equal temperature r i s e i n a l l f u e l channels. I n the MSCR study, fuel-volume f r a c t i o n s i n the range from 0.1 t o

0.2 were investigated.

The optimum fraction appears t o be s l i g h t l y

.

1


331 g r e a t e r than 0.1.

It appears undesirable t o design a matrix having a

f u e l f r a c t i o n much smaller than t h i s .

A t low f u e l f r a c t i o n s , dimensional

t o l e r a n c e s i n t h e machining o f t h e logs and assembling o f t h e matrix introduce u n c e r t a i n t i e s t h a t become an appreciable f r a c t i o n of t h e f u e l volume. This u n c e r t a i n t y can be reduced somewhat by using l o g s o f l a r g e c r o s s s e c t i o n , but there are l i m i t s , v i z , :

(a) a l i m i t on t h e s i z e of

l o g t h a t can c u r r e n t l y or i n t h e foreseeable f u t u r e be manufactured; ( b )

a l i m i t imposed by thermal stress i n t h e log, which becomes excessive with i n c r e a s i n g size.

Graphite logs measuring 16 inches i n diameter can

be made now (1962); while t h e s e are not of a grade s a t i s f a c t o r y f o r use i n t h e MSCR, it is not a g r e a t e x t r a p o l a t i o n of c u r r e n t technology t o p o s t u l a t e t h e a v a i l a b i l i t y o f g r a p h i t e logs o f s a t i s f a c t o r y grade i n s i z e s up t o 8 - i n c h e s i n diameter, and t h i s appears t o be as l a r g e as thermal s t r a i n considerations w i l l allow. If t h e volume f r a c t i o n of f u e l i n t h e core is small, then any s l i g h t

v a r i a t i o n i n t h i s volume f r a c t i o n would have an appreciable effect on t h e r e a c t i v i t y and might r e s u l t i n power excursionso These v a r i a t i o n s might

arise i n any of a number of ways. i

i

For instance, r a d i a t i o n damage might

r e s u l t i n t h e accumulation o f stresses i n t h e g r a p h i t e which, upon sudden removal of external restraints (such as by t h e f a i l u r e of t h e hoops,

etc.) or by t h e y i e l d i n g of t h e m a t e r i a l i t s e l f , might r e s u l t i n g r o s s movement of t h e matrix, and a sudden increase i n r e a c t i v i t y , Also, a t very low volume f r a c t i o n s , t h e optimum concentrations of

f i s s i l e and f e r t i l e i s o t o p e s i n t h e f u e l stream becomes high, and t h i s imposes requirements on t h e design of t h e external system i n regard t o hold-up,

v e l o c i t i e s , etc., t h a t are d i f f i c u l t t o meet.

Permeation of Graphite by Salt Graphite is not impervious t o salt.

The presence of an appreciable

f r a c t i o n of t h e s a l t i n stagnant pockets introduces a number of problems (such as t h a t a s s o c i a t e d with t h e f a t e of f i s s i o n products generated i n stagnant f u e l ) .

The " t h e o r e t i c a l " d e n s i t y of g r a p h i t e is 2.25 grams/cc.

MSRE g r a p h i t e has a bulk d e n s i t y o f about 1.83, and t h u s t h e pore


, 4

332 f r a c t i o n is about 0.16.

The pores accessible t o a wetting f l u i d , such

as kerosene o r a gas, however, amount t o only 0.07% of t h e volume,

Graphite is not wetted by molten f l u o r i d e salts, and t h e penetration is

an order less than t h a t of a wetting f l u i d . Treatment by any one of s e v e r a l methods (5) reduces t h e penetration f u r t h e r , e i t h e r by f i l l i n g t h e pores, by closing them, or by making t h e entrances smaller. which

Tests i n

CGB-X graphite, newly proposed f o r t h e MSRE, was exposed t o s a l t

a t 130OOF and 150 p s i f o r 100 hours r e s u l t e d i n volume f r a c t i o n permeat i o n s of 0,001 and 0,0002 (2, p, 931, The occlusions of s a l t l a y mostly a t t h e surface, and were presumably i n reasonably good d i f f u s i o n a l communication with t h e bulk f l u i d , The pieces t e s t e d were necessarily small, and t h e proportion of surface exposed was high, Estimates based on expected frequency of surface pores i n l a r g e r pieces l e d t o a predict i o n of a penetration not greater than 0.0016 i n t h e MSRE graphite a t 65 p s i pressure i n t h e s a l t ( 2 , p. 91).

Taking advantage of t h e fact

t h a t t h e number of surface pores can be reduced by proper o r i e n t a t i o n of t h e surface grains and t h a t with l a r g e r bars t h e r a t i o of surface-tovolume is less, it appears plausible t h a t a s a l t - a c c e s s i b l e pore f r a c t i o n of 0.001 can be assumed f o r MSCR graphite.

Now, i f t h e graphite occupies

90% of t h e matrix, salt-accessible pores i n t h e graphite amount t o

roughly 0,1%, which is only 1% of t h e t o t a l f u e l f r a c t i o n ,

However, t h e

fuel i n these pores may be retained when t h e r e a c t o r is drained, and t h i s may present a s e r i o u s operational d i f f i c u l t y i n regard t o cooling t h e r e a c t o r after shutdown, e s p e c i a l l y i f t h e core i s drained s h o r t l y

after operation a t power,

Graphite Shrinkage The graphite w i l l , of course, be subjected t o r a d i a t i o n damage, mostly from fast neutrons.

A t t h e temperatures a n t i c i p a t e d i n t h e MSCR,

t h e graphite w i l l shrink,

Since t h a t s i d e of a l o g c l o s e s t t o t h e center

of t h e core w i l l absorb more r a d i a t i o n than t h e o u t e r s i d e , t h e logs

w i l l tend t o bow outward, and increase t h e volume f r a c t i o n of t h e fuel,

A

LJ

These effects w i l l t a k e place slowly, of course, and can e a s i l y be b

c


333

compensated, i n r e s p e c t t o c r i t i c a l i t y , by i n c r e a s i n g t h e c o n c e n t r a t i o n of thorium i n t h e f u e l or by decreasing t h a t of t h e uranium. The breeding

r a t i o w i l l n e c e s s a r i l y decrease, however, due t o s h i f t i n g of t h e C/Th/U r a t i o from t h e optimum. As mentioned above, t h e effect i s minimized by t h e u s e of s h o r t logs.

Graphite Replacement The p r e c i s e effect of g r a p h i t e shrinkage on t h e performance (as meas u r e d by t h e f u e l c y c l e cost) has n o t been determined; however, i f it proves t o be s e r i o u s , one or more of s e v e r a l countermeasures may be taken. The s i m p l e s t would be t o r e p l a c e t h e g r a p h i t e p e r i o d i c a l l y . The excess s i n k i n g fund cost ( @6.75%) over t h a t corresponding t o s i n k i n g fund amort i z a t i o n over a t h i r t y - y e a r l i f e (loll%, and which i s charged o f f t o cap-

i t a l costs) is l i s t e d below i n Table 0.1 f o r s e v e r a l replacements p e r i o d s . Table 0,l. Power Cost Increment f o r Replacement of Graphite* Moderator i n MSCR ~~

Replacement Period (years 1

Incremental Power Cost (mills/kwhr) ~~

1

0.58

5

0.096

10

0.037

15

0.018

20

0,008

*@$6/lb. It is seen t h a t , i f t h e replacement occurs no o f t e n e r t h a n once i n

t e n y e a r s , t h e incremental cost is t o l e r a b l e , being less t h a n 10% of a typical f u e l c o s t of 0.75 mills/kwhr. If t h e s h r i n k a g e rate is such as t o r e q u i r e replacement more o f t e n

than t h i s , then p a r t of t h e f u e l volume f r a c t i o n i n c r e a s e might be avoided by r e s t r a i n i n g t h e core and preventing t h e bowing of t h e logs. Hoops of


334 molydbenum, which have a c o e f f i c i e n t of thermal expansion very nearly equal t o t h a t of graphite ( 3 ) could be placed around t h e core (4, p. 32). O f course, should t h e hoops f a i l suddenly, an excursion i n t h e power

l e v e l might r e s u l t .

On t h e o t h e r hand, . i f t h e hoops held, accumulated

stresses i n t h e graphite might r e s u l t i n t h e formation of cracks i n which t h e f u e l might stagnate with deleterious effect, The f u e l volume f r a c t i o n can be made self-preserving i n s p i t e of r a d i a t i o n induced shrinkage by use o f interlocking blocks of graphite. The moderator elements are cubes having four holes, or sockets, i n t h e

four quadrants of t h e upper face and four corresponding pegs extending from t h e lower face. After a l a y e r of blocks has been l a i d , t h e blocks of t h e next l a y e r are positioned so t h a t t h e a x i s of each block l i e s over t h e i n t e r s e c t i o n s of t h e f u e l channel planes between adjacent blocks i n t h e lower layer, with i t s pins f i t t i n g i n t o sockets i n four blocks i n

t h a t layer.

Thus each cube is pinned t o four overlapping cubes i n t h e

l a y e r above and four i n t h e l a y e r below. along an edge, and a void f r a c t i o n of

With cubes measuring 8 inches

lo%,

t h e thickness of t h e passage

between adjacent cubes is approximately 0.4 inches when t h e blocks rest d i r e c t l y on t h e blocks below, and 0.3 inches when uniform clearance is provided on a l l s i x s i d e s of t h e blocks.

Now, as the blocks shrink, t h e

f u e l volume fraction i s iavariant, since t h i s i s determined by the spacing of t h e pegs and sockets and t h e dimensions of t h e blocks. This arrangement, while solving one problem, introduces others, chief of which i s a much increased resistance t o flow of t h e f u e l through t h e core, which may increase by a f a c t o r of perhaps 20,

This would in-

crease t h e pumping c o s t and t h e design requirements for pumps, heat exchangers and r e a c t o r vessel.

If d i f f i c u l t y with short-circuiting of t h e f u e l through t h e annular

gap between moderator and r e a c t o r v e s s e l i s encountered, t h i s could be controlled by eliminating t h e gaps between t h e o u t e r blocks of graphite

so t h a t t h e blocks f i t t i g h t l y one against t h e other.

The annulus t h u s

becomes a channel unconnected t o t h e voids i n t h e moderator, and t h e flow through it could be o r i f i c e d a t t h e top where t h e p i l e floats up against t h e upper support g r i d ,

.


335 The foregoing discussion of t h e problem of graphite shrinkage indi-

cates some solutions t h a t might be applied i f the problem should prove to be s e r i o u s , but it should not be i n f e r r e d therefrom t h a t t h e problem

is known t o be s e r i o u s , f o r i n fact, it may not be,

There are indica-

t i o n s (6) t h a t r a d i a t i o n damage may anneal and s a t u r a t e a t t h e temperat u r e of operation.

I n t h a t event, t h e core would be designed so t h a t ,

after t h e steady s t a t e is reached, t h e f u e l volume f r a c t i o n is a t t h e desired value.

D i f f e r e n t i a l Expansion The c o e f f i c i e n t of thermal expansion of graphite is smaller than t h a t of INOR (compare Tables 3.2 and 3.1).

Thus, i n a 20-ft core with t h e

graphite j u s t f i l l i n g t h e vessel a t room temperature, t h e r e w i l l be a gap about 2/3-in0 wide between t h e graphite and t h e vessel w a l l a t llOO째F. If f u r t h e r allowance is made f o r dimensional tolerances between metal and

graphite during assembly, t h e gap cannot be much less than one inch t h i c k

a t operating temperatures, This gap w i l l contain f u e l s a l t , and t h i s fact must be taken i n t o account i n evaluating t h e performance of t h e r e a c t o r and i n t h e design of t h e core and t h e r e a c t o r vessel.

I n a 20 f t c y l i n d r i c a l core t h e volume

of f u e l i n t h e annulus amounts t o

%

100 f t 3 , which is an appreciable

fraction of t h e volume of f u e l i n t h e matrix

(%

600 f t 3 ) .

This f u e l l i e s

i n a region of low neutron population so t h a t it adds l i t t l e t o t h e reactivity,

Also, it i s nearly opaque t o thermal neutrons which are

absorbed, multiplied by eta, and re-emitted as fast neutrons, thus increasing t h e leakage.

This concentrated source of fast neutrons and

concomitant gamma r a d i a t i o n adjacent t o t h e r e a c t o r v e s s e l w a l l i n t r o duces design problems,

It may be necessary t o provide a thermal s h i e l d

between t h e f u e l annulus and t h e w a l l .

The most s u i t a b l e material f o r

t h i s s h i e l d is INOR, but t h i s must be cooled,

The only a v a i l a b l e

cooling medium i n t h i s s i t u a t i o n i s t h e f u e l s a l t , and i t s use f o r t h i s purpose f u r t h e r increases t h e nonactive inventory of valuable materials.


336 It may be possible t o use t h e f u e l annulus as a downcomer for f u e l coming from t h e heat exchangers.

This may or may not r e s u l t i n some .

saving i n f u e l inventory, depending on t h e location of t h e exchangers and t h e i r requirements f o r draining, etc,

The annulus is so used i n t h e

MSRE (1).

References

1. A, Lo Boch e t al., The Molten-Salt Reactor Experiment, Power Reactor Experiments, Vol. I , pp. 247-292, I n t e r n a t i o n a l Atomic Energy Agency, Vienna, 1962.

.

2.

Oak Ridge National Laboratory, MSRP Semiann. P r o p . Rept, Feb. 28, 1962, USAEC Report ORNL-3282

3.

S. E. Beall e t al., Molten-Salt Reactor Experiment Preliminary Hazards Report, USAEC Report ORNL CF-61-2-46 (with Addenda 1 and 21, Oak Ridge National Labo=atory, February 1961.

4.

Oak Ridge National Laboratory, MSRP Quarterly Progr. Rept. J u l y 31, 1960, USAEC Report OWL-3014.

5.

J. W. H. Simmons, The Effects of I r r a d i a t i o n on t h e Mechanical Propert i e s of Graphite, Proceedings of t h e Third Conference on Carbon, Sections E and F, Pereamon Press. 1959.


337 BIBLIOGRAPHY

1. L. G. Alexander e t a l . , Nuclear C h a r a c t e r i s t i c s of Spherical, Homogeneous, Two-Region, Molten-Fluoride-Salt Reactors, USAEC Report ORNL-2751, Oak Ridge National Laboratory, September 1959. 2.

L. G. Alexander e t a l . , Experimental Molten-Salt-Fueled 30 Mw Power Reactor, USAEC Report ORNL-2795, Oak Ridge National Laboratory, March 1960

3.

L. G. Alexander e t a l . , Thorium Breeder Reactor Evaluation - P a r t I Fuel Yield and Fuel Cycle Costs i n Five Thermal BTeeders, USAEC Report ORNL-CF-61-3-9, Oak Ridge National Laboratory, March 1961, and USAEC Report ORNL-CF-61-3-9 (Appendices ), Oak Ridge National Laboratory, March 1961-

4.

Robert Avery e t a l . , Report of t h e Fluid Fuel Reactors Task Force t o t h e Division of Reactor Development, USAEC, USAEC Report TID-8507, United S t a t e s Atomic Energy Commission, February 1959.

-

5. S. E. Beall e t a l . , Molten-Salt Reactor Experiment Preliminary Hazards Report, USAEC Report OWL-CF-61-2-46, (Addendum 1 and 2 ) , Oak Ridge National Labor a t ory , Febmary 1961. 6.

E. S. B e t t i s e t a l . , The A i r c r a f t Reactor Experiment - Design and Construction, Nuclear Science Engineering, 2:804 (November 1957).

7. E. S. B e t t i s e t a l . , The A i r c r a f t Reactor Experiment

- Operation,

Nuclear Science Engineering, 2: 841 (November 1957).

8.

B. C. Blanke e t a l . , Density and Viscosity of Fused Mixtures of Lithium, Beryllium, and Uranium Fluorides, USAEC Report MLM-1086, Mound Laboratory, March 1959.

9.

J. 0, Blomeke, Nuclear P r o p e r t i e s of 235U F i s s i o n Products, USAEC Report ORNL-1783, Oak Ridge National Laboratory, A p r i l 1957.

10. J. 0. Blomeke and M. F. Todd, Uranium-235 F i s s i o n Product Production, USAEC Report ORNL-2127, Oak Ridge National Laboratory, November 1958.

11

C. M Blood, S o l u b i l i t y and S t a b i l i t y o f S t r u c t u r a l Metal Difluorides i n Molten Fluoride Mixtures, USAEC Report ORNL-CF-61-5-4, National Laboratory, May 1961.

Oak Ridge

12.

A. L. Boch e t al., The Molten-Salt Reactor Experiment, Power Reactor Experiments, Vol. I, I n t e r n a t i o n a l Atomic Energy Agency, Vienna, 1962, pp. 247-292.

13*

J. 0. Bradfute e t a l . , An Evaluation of Mercury Cooled Breeder Reactors, USAEC Report ATL-A-102, Advanced Technology Laboratory, October 1959.


338 14

R. C. Briant and A. M. Weinberg, Molten Fluorides as Power Reactor Fuels, Nuc. Sci. Eng., 2:797 (November 1957).

15

J. Bulmer e t a l . , Fused S a l t Fast Breeder, USAEC Report ORNL-CF-56-

v

8-204 (Del. ), Oak Ridge National Laboratory, August 1956. 16. D. 0. Campbell and G. I. Cathers, Processing of Molten S a l t Power Reactor Fuels, Ind. Fhg Chero., 52:4l (January 1960).

.

17

M. C. Cannon e t a l . , Adsorption of Xenon and Argon on Graphite, USAEC Report ORNL-2955, Oak Ridge National Laboratory, November 1960.

18. W. H. C a r r e t al., Uraniwn-Zirconium Alloy Fuel Processing i n the ORNL V o l a t i l i t y P i l o t Plant, USAEC Report ORNL-2661, Oak Ridge National Laboratory, July 1962. 19

W. L. Carter, Thorium U t i l i z a t i o n Program: A Survey of Processing Methods f o r Thorium Reactor Fuel, USAEC Report OTNL-TM-241, Oak Ridge National Laboratory, July 1962.

20.

W. L. Carter, Thorium U t i l i z a t i o n Program: Fuel Cycle Cost estimates for the Molten S a l t Converter Reactor, USAEC Report ORNL-CF-62-12-55, Oak Ridge National Laboratory, December 1962.

21

W. L. Carter, R. P. Milford, and W. G. Stockdale, Design Studies and Cost Estimates of Two Fluoride V o l a t i l i t y Plants, USAEC Report ORNL-TM ( i n preparation), Oak Ridge National Laboratory, October 1962.

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26

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27

S. I. Cohen and T. N. Jones, A Summary of Density Measurements on Molten Fluoride Mixtures and a Correlation for Predicting Densities of Fluoride Mixtures, USAEC Report ORNL-1702 (Declassified), Oak Ridge National Laboratory, J u l y 1954.

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339 28.

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29.

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30

T. A. Eastwood and R. D. Werner, The Thermal Neutron Capture Cross Section and Resonance Capture I n t e g r a l of Protactinium-233, Can Jr. Phys. , 38: 751 (June 1960).

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32

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34.

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35

J. D. Gabor e t al., Spray Fluorination of Fused S a l t as a Uranium Recovery Process, USAEC Report ANL-6131, Argonne National Laboratory, March 1960.

36

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37.

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38.

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39.

C. R. Greenhow and E. C. Hansen, Thermal and Resonance Fission-Product Poisoning for 235U System, USAEC Report KAPL-2172, Knolls Atomic Power Laboratory, October 1961.

40

W. R. Grimes e t a l . , Reactor Chemistry Division Annual Progress Report f o r Period Ending January 31, 1960, USAEC Report ORNL-2931, pp. 87-88, Oak Ridge National Laboratory, April 1960.

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340 42. R. Gwin and D. W. Magnuson, Determination of Eta by Comparison of 233U and 235U in a Flux Trap Critical Assembly, Nuc. Sci. Eng., U(3): 359, March 1962.

43

9

R. Gwin and D. W. Magnuson, The Measurement of Eta and Other Nuclear Properties of 233U and 239Pu in Critical Aqueous Solutions, Nuc. Sci. Eng., l2(3):364, March 1962.

44

J. A. Harvey, O a k Ridge National Laboratory, personal communication to C. W. Nestor, Oak Ridge National Laboratory, March 1960.

45

E. C. Hensen, A Critical Examination of the Uncertainties in Predicted Gross Fissioa Product Poisoning, USAEC Report KAPL-M-ECHi8, Knolls Atomic Power Laboratory, March 1962.

46

E. C. Hensen and C. R. Greenhaw, An Improved Generalized Analysis of Fission Product Poisoning and Thermal and Resonance Fission Fragment Cross Sections, US.4EC Report KAPL-M-ECH-7, Knolls Atomic Power Laboratory, September 1960.

47

I).

48

The Industrial Graphite Engineering Handbook, National Carbon Company, Division of Union Carbide Corporation, 30 East 42nd Street, New York 17, New York, 1962.

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49. T. Jarvis et al., Reactor Design and Feasibility Problem, USAEC Report ORNL-CF-53-10-26(Casssified), pak Ridge National Laboratory, August 1953. 50.

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51.

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52

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53

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54

B. W. Kinyon and F. E. Romie, Two Thermodynamic Systems for Molten Fluoride Reactors, Trans. Am. Inst. Chem. Engrs. Preprint No. 177, Third Nuclear Science and Engineering Conference, March 1958.

55.

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r


341 56

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58.

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59.

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60.

H. G. MacPherson e t a l . , A Preliminary Study of a Graphite-Moderated Molten S a l t Power Reactor, USAEC Report CF-59-1-26, Oak Ridge National Laboratory, January 1959.

61.

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62

H. G. MacPherson, Molten-Salt Breeder Reactors, USAEC Report ORNL-CF59-12-64 (Revised), Oak Ridge National Laboratory, December 1959.

63

H. G. MacFnerson, Molten-Salt Reactors: Report f o r 1960 Ten-Year Plan Evaluation, USAEC Report ORNL-CF-60-6-97, Oak Ridge National Laborat o r y , June 1960.

64.

B. Manowitz, Fuel Reprocessing Costs, Nucleonics, 20(2):60, February 1962

65

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66

R. P. Milford e t al., Recovering Uranium Submarine Reactor Fuels, Ind. Ehg. Chem., 53:357, May 1961.

67

J. W. Miller, Thorium Resonance Cross-Sections f o r Thermal Breeder Reactor Study, USAEC Report ORNL-CF-61-1-26, Oak Ridge National Laboratory, January 1961.

68

E. C. Moncrief, Corrosion of t h e V o l a t i l i t y P i l o t Plant INOR-8 Hyctrof l u o r i n a t o r and L-Nickel Fluorinator after 2 1 Nonradioactive Dissolut i o n Funs, USAEC Report ORNL-TM-186, Oak Ridge National Laboratory, March 1962.

69.

M. S. Moore, MTR Nuclear Physics Group, pezsonal communication t o C. W. Nestor, O a k Ridge National Laboratory, March 1960.


342 70

t

71

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T'nermal and Resonance Absorption Cross Section of the Fission Products, USAEC Report ORNL-2869, Oak Ridge National Laboratory, Mach 1960.

233U,

23%,

j3'Pu

72

C. W. Mestor, Multigroup Neutron Cross Sections, USAEC Report ORNL-CF61-6-87 (Rev. ), Oak Ridge National Laboratory, June 1961.

73.

Oak Ridge National Laboratory, Chemical Technology Division Unit Operat i o n s Section Monthly Progress Report, USAEC Report ORNL-TM-34, p. 44 (7.0 V o l a t i l i t y ) Oak Ridge National Laboratory, June 1961.

74

Oak Ridge National Laboratory, Reactor Chemistry Division Annual Progress Report f o r Period Ending Jazuary 21, 1951, U " U C Report ORXL3127

75. Oak Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report f o r Period Ending October 31, 1957, USAEC Report ORNL-2431-

Oak Ridge Xational Laboratory, Molten-Salt Reactor Program Quarterly Progress Report f o r Period Ending June 31, 1958, USAEC Report ORNL2551.

76

77

9

Oak Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report f o r Period Ending October 31, 1958, USAEC Report ORNL-2626

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78.

Oak Ridge National Laboratory, Molten-Salt Reactor Project Quarterly Progress Report f o r Period Ending January 31, 1959, USAEC Report ORNL-2684.

79

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80.

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81.

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82.

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m


343

83.

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84.

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85.

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87.

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89.

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90.

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91

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92

C. C. Randall e t al., Study of Mercury Binary Cycles f o r Nuclear Power P l a n t s , Report WCAP-1832, Westinghouse E l e c t r i c Corp., J u l y 1961.

93 *

J. E. Ricci, Guide t o t h e Phase Diagrams of t h e Fluoride Systems, USAEC Report ORNL-2396, Oak Ridge National Laboratory, November 1958.

94.

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95.

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344 97.

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J. W. â&#x201A;Ź3. Simmons, The Effects of I r r a d i a t i o n on the Mechanical Prop e r t i e s of Graphite, i n Psoceedings of the Third Conference on Carbon, Sections E and F, Pergamon Press, 1959.

99.

I. Spiewak and L. F. Parsly, Ebaluation of External Hold-up of Circul a t i n g Fuel Thermal Breeders as Related t o Cost and F e a s i b i l i t y , USAEC Report ORNL-CF-60-5-93, Oak Ridge National Laboratory, May 1960.

lQ0. P. E. Spivak e t a l . , Measurement of E t a f o r 233U, 235U, and 239Pu with Epithermal Neutrons, J. NUC. Energy, 4:70, January 1957.

101. R. E. Thoma, C r y s t a l l i z a t i o n Reactions i n t h e Mixture LiF-BeF2-ThF4 (67.5-17.5-15 mole 4 ) BELT-15, USAEC Report ORNL-CF-59-4-49, Oak Ridge National Laborstory, A p r i l 1959.

102. R. E. Thana ( E d i t o r ) Phase Diagrams of Nuclear Reactor Materials, USAEC Report ORNL-2548, Oak Ridge National Laboratory, November 1959.

103. R. E. Thoma and W. R. Grimes, Phase Equilibrium Diagrams f o r Fused S a l t Systems, USAEC Report ORNL-2295 (Decl. ), Oak Ridge National Laboratory, June 1959.

USAEE, Summary Report: AEC Reference Fuel Processing Plant, USAEC Report WASH-743, October 1957.

105. W. T. Ward e t a l . , Rare Earth and Y t t r i u m Fluorides - S o l u b i l i t y Relations i n Various Molten NaF-ZrF4 and NaF-ZrFb-UF4 Solvents, USAEC Report ORNL-2421, Oak Ridge National Laboratory, January 1958.

106

W. T. Ward e t a l . , S o l u b i l i t y Relations Among Rare-Earth Fluorides i n Selected Molten Fluoride Solvents, USAEC Report ORNL-2749, Oak Ridge National Laboratory, October 1959.

107.

G. M. Watson and R. B. Evans 111, Xenon Diffusion i n Graphite: Effects of Xenon Absorption i n Molten S a l t Reactors Containing Graphite, USAEC Report ORNL-CF-61-2-59, Oak Ridge National Laboratory, February 1961.

108. C. F. Weaver e t a l . , Phase E q u i l i b r i a i n Molten S a l t Breeder Reactor Fuels, 1 - The System LiF-BeFz-ThF4, USAEC Report ORNL-2896, Oak Ridge National Laboratory, December 1960. 109.

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iu

345

iP

!

I

!

110. C. H. Wescott, Effective Cross SectionTalues for Well-Moderated Thermal Reactor Spectra, Canadian Report CRRP-787, Chalk River, Ontario AECL-670, August 1, 1958 (Rev. 1958).. 111. N. F. Wikner and S. Jaye, Energy-Dependent and Spectrum-Averaged Thermal Cross Sections for the Heavy Elements and Fission Products for Various Temperatures and C:235U Atom Ratios, USAEC Report GA2113, General Atomic Division, General Dynamics Corporation, June 1961

112. H. U. Woelk, Molten Salts in Nuclear Technology, Chemie-IngenieurTechnik, 32:765, 1960. AEC-TR-4774 by A. R. Saunders and H. H. Stone, Oak Ridge National Laboratory, August 1961.

19

i


- *

I


347

Internal Distribution 1. 2-20. 21. 22. 23 24. 25. 26. 27. 28. 29. 30. 31. 32 33. 34. 35. 36 37. 38. 39. 40. 41. 42. 43. 44. 45. 46. 47 48. 49. 50. 51 52 53 54 55. 56. 57

58 59. 60 61* 62. 63 * 64 65. 66. 67.

T. L. H. S. L. E. F. E. A. E. C. R. F. D. S. W. G.

R. R. C. W. C. F.

S. J. R. D. J. W. A. J. W. E.

H. W. A. C. R. P. E. H. D. W. P.

R. M. T.

J. C.

D. Anderson G. Alexander F. Bauman E. Beall L. Bennett S. B e t t i s F. Blankenship P. Blizard L. Boch G. Bohlmann J. Borkowski B. Briggs R , Bruce 0. Campbell Cantor L. Carter I. Cgthers H. Chapman S. Carlsmith W. Collins H. Cook W, Craven L. Culler G. Delene H. DeVan G. Donnelly A. Douglas R . Engel K. Ergen P. Fraas H. Frye, Jr. R. G a l l H. G i f t E. Goeller R. Grimes G. G r i n d e l l E. Guthrie P. Hammond N. Haubenreich

C. Hise W. Hoffman B. Janney (K-25) H. Jo,rdan R. Kasten J. Ked1 T. Kelley W. Kerlin A. Lane E. Larson

68. C. G. Lawson 69. R. B. Lindauer 70 * J. L. Lucius 7 1 M. I. Lundin R . N. Lyon 72 73. H. G. Macherson 74. H. C. McCurdy 75. W. B. McDonald H. F. McDuffie 76 77. R. P. Milford A. J. Miller 78 79. R. L. Moore 80. J. C . Moyers 81. R. W. Olson 82. H. R. Payne 83. A. M. Perry 84. B. E. Prince 85. R. C. Robertson 86. M. W. Rosenthal 87. H. W. Savage 88. A. W. Savolainen 89. D. S c o t t 90. J. H. Schaffer 91. M. J. Skinner I. Spiewak 92 93. W. G. Stockdale 94. A. Taboada J. R. Tallackson 95 96. R. E. Thoma 97. D. B. Trauger 98. J. W. Ullman D. R . Vondy 99 100 R. Van Winkle 101. G. M. Watson 102. A . M. Weinberg 103. J. H. Westsik 104. G. D. Whitman 9

105. K. J. 106. 107.

108-109. 110-111

112-113

114-115. 116-145 146.

YOSt

Gale Young Biology Library Reactor Division Library MSRP D i r e c t o r ' s Office, Room 219, Building 9204-1 ORNL Y-12 Technical Library, Document Reference Section Central Research Library Laboratory Records Department Laboratory Records, ORNL, R.C.


348

_-. -

Lli External Distribution 147. 148. 149-150 151. 152. 153. 154. 155 156 157-158

-

159. 160. 161 162. 163. 164. 165 166 167 168. 169 170 171-185.

G. Beasley, TVA, EGCR Site, Oak Ridge, Tennessee W. R. Cooper, W A , EGCR Site, Oak Ridge, Tennessee D. F. Cope, AEC, OR0 W. E. Dean, Jr., TVA, Chattanooga, Tennessee Division of Research and Development, AEC, OR0 J. Ebersole, TVA, EGCR Site, Oak Ridge, Tennessee W. F. M o n s , TVA, Knoxsrille, Tennessee R. W. Garrison, AEC, Washington, D. C . R. Harris, AEC, OR0 C. A. Hatstat and W. Chittenden, Sargent and Lundy, Chicago, Illinois S. Jaye, General Atomic, San Diego, California R . W. McNamee, Manager, Research Administration, UCC, New York, N. Y. G. P. Palo, TVA, Knoxville, Tennessee R. E. Pahler, AEC, Washington, D. C. P. F. Pasqua, University of Tennessee, Knoxville, Tennessee R. L. Phillipone, Reactor Division, AEC, OR0 H. M. Roth, Division of Research and Development, AEC, OR0 F. P. Self, AEC, Washington, D . C. W. L. Smalley, Research Division, AEC, OR0 J. M. Vallance, AEC, Washington, D. C. M. J. Whitman, AEC, Washington, D. C. J. Wett, AEC, Washington, D. C. Division of Technical Information Extension (DTIE)

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