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This report was prepared by a t a s k force composed of f i f t e e n highly qualified engineers and s c i e n t i s t s from the Atomic E=nergy Commission, its national laboratory and contractor organizations, and representatives of the architect-engineering and u t i l i t y industry, The t a s k force w a s named the Fluid Fuel Reactors Task Force and was organized and convened by t h e Evaluation and Planning Branch of tihe Office of C i v i l i a n Reactors t o perform a c r i t i c a l evaluation of the t h r e e f l u i d f u e l reactor concepts (aqueous homogeneous MR, molten salt MSR, and l i q u i d metaJ. f u e l LMFR) under development by t,he Commission,

-

-

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The Task Force met continuously during January and February of

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1959 and evaluated information presented by the national, l a b o r a t o r i e s and i n d u s t r i a l contractors developing the concepts. This document i s t h e report of the Fhuid Fuel Reactors Tcsk Force t o the Division of Reactor Development and represents the group evaluation and judgment o f the three f l u i d f u e l reactor concepts.

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3 4 4 5 6 0024205 7 I


ii

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E


TABU O F CON!lE3TS

, I.

11.

III.

. . . . . . . .

INTRODUCTION

OWECTIVES OF F L U I D FUEL REACTORS TASK IKIRCE STUDY SUMMARY

. . . . . . . . .

0

1

.

3

0

4

. . . . . ... ... ... SYSTEIG . . . . . . 9 OF STATE V. COMPARATIVE . . . . . . 11 Fuels and Materials . . . .. .. .. 14 Primary System and Components . Reactor . . .. .. . .. .. . 1164 Primary System . 3 . Primary fiuxiliary System . 4. Instrumentation . . .. . .. . 1917 Operation . . .. .. .. .. .. . .. 21 Maintenance . Control . . .. .. . . .. .. 23 Hazards Chemical Reprocessing . . . 2431 FEASIBILITY BREEDING . . . 39 Breeding Potential . . . . .. . . 3942 Additional Cost to Attain Breeding . Necessary Developments . .. .. .. .. . 4243 Status Basic Physics v1:r. . . . . . . . e 4 4 Preparation the Comparative Power Cost Tabulations & B. Conclusions . . . . .. .. .. . LS47 Presentation Cost Information A. B. C. D.

IV.

Present State of Development and Technical Feasibility 4 Technical Feasibility of Breeding 5 Power Cost . 6 Research and Development Programs 0 7

REFEJREJACE REACTOR

ANALYSES TiiKNICAL F E S I B I L I T P

OF DEVEIDPIJE'iT AND

PRESENT

0

A.

0

B.

0

1, 2.

e

*

0

20

C. D. E.

0

F, G.

VI.

*

0

0

OF

TECHNICAL A. B. C. D.

C.

0

0

0

Data

of

POWER COSTS

A.

11

0

of

of

iii

i E


i

VII. POWER COSTS (Continued)

Table V I I - 1 Table VII-2 Table V I I - 3

VIII. IX.

X.

XI, XII.

-- Overall Summary o f Power C o s t s Power Plant Investment . - Summary of Fuel Costs .

RESEXRCH AND DEVELOPMENT PROGRAMS

MOLTEN SALT FBACTOR

0

0

. .

.. . . . .

. . . . . . HOMOGE3M)US REACTOR . . ACKN0WLE;DGEIJIDITS . . . . . . . .

AQUnOnS

APPENDIX B

50 52

53

55

LIJUID FETAL FUEmD REACTOR

APPENDIX A

49

- Ad

. . . . . .i n . . .

Hoc Committee's Statements on Fluid Fuel Reactors

- List of Items Included Investment Cost

iv

Power Plant

85 1-33

176

177

179

t

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I.

L

INTRODUCTION

The Commission organized the Ad Hoc Advisory Committee on Reactor Policies and Programs i n the F a l l of 1958 to: "(1) review the Commission's c i v i l i a n power program; "(2) advise the Commission i n connection with the formul a t i o n of a sound basic policy i n the l i g h t of our current economy and the present stage of development of nuclear power technology;

" ( 3 ) assist the Commission i n establishing new goals and redefining the problems connected with the foreseeable expansion of commercial u t i l i z a t i o n of nuclear power and the concomitant development of a vigorous nuclear equipment manufacturing industry; and

It(4) recommend immediate and long-range programs t o achieve the Commission's new goals." fllhe Committee presented t h e i r report t o the Commission, January 2, 1959. One of t h e i r recommendations f o r the technical progr,m i n the near future i s reproduced below: Yhe aqueous homogeneous, molten bismuth, and molten salt reactors all offer the p o s s i b i l i t y of reducing the cost of the f u e l cycle, and the last two offer the possibility of high temperature operation. These three concepts f o r power reactors should be c r i t i c a l l y compared and work concentrated on the concept t h a t appears the most promising. I t Further comments of the Committee dealing with Fluid h e 1 Reactors are reproduced i n Appendix A. The hraluation and Planning Branch of the C i v i l i a n Reactors Office, Division of Reactor Development, organized and convened the Fluid Fuel Reactors Task Force i n January, 1959, t o make the comparison of the aqueous homogeneous, molten salt, and liquid metal fueled reactor concepts i n accordance with the Ad Hoc Committ e e ' s recommendations.

L t


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The Task Force consisted of the following members: Robert Avery Physicist, Argonne National Laboratory, The University of Chicago Robert Blum (Vice Chairman of the Task Force) AEC, Division of Reactor Development R. Beecher B r i g g s Director, Homogeneous Reactor Project, Oak Ridge National Laboratory, Union Carbide Nuclear Co. Jack Chernick Physicist, Brookhaven National Laboratory, Associated Universities Inc, Wilson R, Cooper, Nuclear Development Engineer, Tennessee Valley Authority Joseph E. Draley Metallurgist, Urgonne National Laboratory, The University of Chicago James E. Evans Chemical Engineer, Atomic Energy Division, E. I. du Pont de Nemours & Co., Inc. Edgar E, Hayes Metallurgist, Atomic Energy Division, E. I. du Pont de Nemours & Co., Inc, Titus G. LeClair, Edison E l e c t r i c I n s t i t u t e Representative; Manager, Research and Development, Commonwealth Edison Co. Jack S. B i t e l (Alternate) Mechanical Engineer, Commonwealth Edison Co. H. G. MacPherson Director, Molten S a l t Reactor Program, Oak Ridge National Laboratory, Union Carbide Nuclear Co, Jack B. McKamey, Construction Manager, Ebasco Services, Inc. Francis T. Miles Director, Liquid Metal Fuel Reactor Project, Brookhaven National Laboratory, Associated Universities, Inc. C a r l L. Newman Power Engineer, United Ehgineers & Constructors Inc. Robert W. Ritzman (Chairman of the Task Force) AXC, Division of Reactor Development Vincent A. Walker Chemical Engineer, National Reactor Testing Station, Phillips Petroleum Co. The Atomic Energy Commission wishes t o take t h i s opportunity t o thank the members f o r t h e i r devoted and d i l i g e n t p a r t i c i p a t i o n on t h e Task Force.

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TI,

OBJXTIVES O F F L U I D FUEL REc1C’iy)RS TASK FORCE STUDY

The Zvaluation and Planning Branch of the Office of C i v i l i a n Beactors, Division of Reactor Development requested the Task Force t o &ke a r e l a t i v e comparison of t h e f l u i d f u e l r e a c t o r concepts (aqueous homogeneous--AHR, molten salt--MSR, and l i q u i d metal fuelLMFR) t o determine the:

a,

present state of development and technical f e a s i b i l i t y ,

b.

technical f e a s i b i l i t y of breeding,

C.

p o t e n t i a l power cost i n mills per kilowatt-hour i n a system optimized f o r power production, and

d,

research and development program (direction, timing, and c o s t ) necessary t o produce a reactor capable of achieving the cost p o t e n t i a l mentioned above, f o r each concept ,

! h e Evaluation and Planning Branch requested enough information of a comparative nature t o enable the Atomic Energy Commission t o choose t h e proper course of a c t i o n and magnitude of support t h a t would be required t o develop each type,


- 4 I

A.

Present S t a t e of Development and Technical F e a s i b i l i t y

The molten salt r e a c t o r has t h e highest probability of achie i n g technical f e a s i b i l i t y . This is l a r g e l y due t o t h e use of a s o l u t i o n f u e l (as contrasted t o a s l u r r y f u e l i n t h e LMFR and the AHR), and t h e a v a i l a b i l i t y of a s u i t a b l e container material (IHOR-~). Summaries of the r e l a t i v e comparisons of t h e three concepts follow :

1. The technical f e a s i b i l i t y of f u e l s and materials is a c r i t i c a l factor. A t t h e present s t a t e of technology, t h e MSR has t h e best p o s s i b i l i t y of obtaining a s a t i s f a c t o r y f u e l , i f indeed i t does not already have a s a t i s f a c t o r y fuel. S l u r r i e s , as used by the LTYZFR and AHR require a g r e a t e r amount o f development e f f o r t t o e s t a b l i s h f e a s i b i l i t y . "he FER a l s o o f f e r s t h e best p o s s i b i l i t y f o r achieving a s a t i s f a c t o r y container material since t h e LMFR and the AH2 have d i f f i c u l t materials problems at the present stage of technology. However, t h e compatibility of molten s a l t f u e l with graphite which is contemplated f o r use for t h e i n t e r n a l construct i o n of a reactor still remains t o be demonstrated and the problem is judged t o be more severe than i n t h e I;MFF.. The achievement of s a t i s f a c t o r y primary and a u x i l i a r y 2. systems and components depends l a r g e l y on matters of engineering ingenuity and i s believed t o be technically f e a s i b l e ; however, these systems w i l l be complicated and hence expensive. I n comparing t h e molten salt with t h e l i q u i d metal f u e l reactor, no significance has been placed on d i f f i c u l t i e s a r i s i n g from t h e molten salt solution's higher melting point (975째F VS. 525OF) and higher top operating temperature (l225'F VS. 1050째F). D i f f i c u l t i e s i n design caused by t h e higher temperatures are o f f s e t by the fact t h a t t h e MSR primary system w i l l be smaller than. t h e W R system because of t h e higher volumetric heat capacity of t h e salt,

3. From t h e standpoint of operation, i t is anticipated t h a t a l l t h r e e reactor concepts can be designed t o meet load changes, and i t is assumed t h a t they will be able t o operate f o r extended periods of time. However, t h i s has not y e t been demonstrated. When considered from t h e standpoint of r e l i a b i l i t y f o r extended periods of time, the AHR i s a t a disadvantage because of its more extensive and complex a u x i l i a r y systems.

ck

t

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-5Maintenance i s the most important f a c t o r influencing the p r a c t i c a b i l i t y of any of the three concepts. A t present the f e a s i b i l i t y of maintaining a l a r g e f l u i d f u e l r e a c t o r power s t a t i o n i s doubtful because of the need f o r c i r c u l a t i n g a high-level radioa c t i v e , f l u i d f u e l stream. While experience with the HRX-2 has shown t h a t i t can be maintained by the use of wet maintenance, i t i s not known i f these techniques can be applied t o l a r g e plants. The use of remote dry maintenance f o r any of t h e p l a n t s i s unproven. The f e a s i b i l i t y of any maintenance scheme can only be established by a comprehensive design study backed up by extensive full-scale mockup t e s t i n g under conditions simulating a c t u a l requirements.

4.

5. 31.e AHR is e a s i l y controlled, as has been indicated by r e a c t o r experience. Analytical s t u d i e s show t h a t the PER and LEIFR can be controlled but this requires experimental v e r i f i c a t i o n . Both may require shim rods. 6. The AHX i s p o t e n t i a l l y the most hazardous because of its high pressure system, r a d i o l y t i c gas explosion hazard, and the p o t e n t i a l i n s t a b i l i t y of the fuel. However, HRE-2 has operated f o r a long period of time under very disadvantageous circumstances without s e r i o u s release of radioactivity. The PIS2 and LMFR a r e similar t o each other i n t h e i r s a f e t y c h a r a c t e r i s t i c s ,

7. With the exception of the AIR, chemical reprocessing f o r the reference designs has received l i t t l e attention. B.

Technical F e a s i b i l i t y of Breeding

The evaluation of breeding i n the f l u i d f u e l systems on the thorium-U233 cycle leads t o the following conclusions : 1.

One-region r e a c t o r s can breed only i n large s i z e s e

2. It i s highly probable t h a t a l l three of the systems can be developed i n t o "hold own" breeders. Likely breeding r a t i o s f o r t h e AHR, LMFR, and MSR a r e 1.09, 1.05, and 1.05, respectively,

3 . On t h e basis of a reasonable extrapolation of t h e current development program, only t h e AHR has the p o s s i b i l i t y of achieving a reasonably s h o r t doubling time (of the order of 15 years). 4, The LMFR and MSR can a l s o achieve reasonable doubling times i f i n t e r n a l l y cooled. However, t h e required development programs make the internally-cooled, non-aqueous doublers considerably fa:rther away i n terms of both time and money than i s the AHR doubler.


-6-

5. The uncertainty i n eta f o r U-233 does not seriously jeopardize the probability of all three systems being "hold own" breeders, but it i s c r u c i a l i n the f e a s i b i l i t y of the doublers. C.

Power Costs

The cost tabulations summarized below and presented more f u l l y i n Section V I 1 were prepared t o provide a b a s i s f o r evaluating the r e l a t i v e power cost p o t e n t i a l i t i e s and indicated l e v e l of investments of t h e molten salt, l i q u i d metal f u e l , and aqueous homogeneous reactor concepts f o r l a r g e c e n t r a l power stations. Summary of Power Costs (For Reference Reactors: 333 W E , gross) Overall Summary of Power Costs)

(Based upon Table V I I - 1 :

Mills/kwh

Power Plant Investment

d

Chemical & h s t e Disposal Plant Investment

Y

Fuel Inventory Use and h m u p Chemical Plant Operation & Maint. Power Plant Operation & Maint.

- Gross Power Costs - Net

LMFR -

MSR

-

*mu

5.72

6.24

6.70

.

71

.89

1.36

1.37

090

.?6

76

.71 1.46 -

1.57

0

96

1.66

Total Power Costs

10.0

10.7

11.1

Total

10.7

11.1

11.5

-

Notes:

For d e t a i l s and explanation of power plant investment data, see Section V I 1 and Table V I I - 2 .

b

For d e t a i l s and explanation of chemical processing and w a s t e disposal plant investment figures, see Section V. For divergent views of Project Directors on c o s t s of chemical processing, see pages 76, 95, and 169. For d e t a i l s and explanation of f u e l costs, see Section V I I , and Table VII-3.

h


- 7 -

--

Notes:

(cont'd.)

The chemical processing plant operation and maintenance c o s t s are computed as 15% of the chemical processing plant investment shown above, based on Section V, p a r t G. For divergent views of l?roject Directors on c o s t s of chemical processing, see pages 76, 95, and 169. Operating c o s t s are based on estimates of personnel requirements and other expenses. Power plant maintenance has been estimated a t 3% of t o t a l power plant investment, including d i s t r i b u t i v e items. Costs shown f o r AHR are f o r two-region solution core power breeder. Lower c o s t s estimates f o r slurry fueled reactor are presented i n Section X I , page 168. The l e v e l of t h e c o s t d a t a presented on t h e preceding page c l e a r l y i s s u b s t a n t i a l l y higher than costs indicated i n some earlier r e p o r t s on f l u i d f u e l reactor concepts.

I n general, the data do not i n d i c a t e any very s u b s t a n t i a l overall difference i n t h e r e l a t i v e cost p o t e n t i a l of t h e three concepts'. For v a l i d comparison of c o s t s f o r these r e a c t o r concepts with costa of other reactor concepts o r with c o s t s of conventional plants, it i s of t h e utmost importance t o take i n t o account the differences i n t h e many f a c t o r s and conditions entering i n t o t h e various estimates. Conclusion In determining f l u i d f u e l reactor programs, the cost data presented herein should be regarded as secondary t o technical f e a s i b i l i t y and p o s s i b i l i t i e s f o r advances by research and development.

The indicated c o s t s are not s o high as t o preclude a reasonable research and development e f f o r t t o develop and improve f l u i d f u e l technology.

D.

Research and Development Programs

TheaTask Force did not prepare a d e t a i l e d conparison of the required research and development programs f o r the three reactor concepts. The programs presented were prepared by the individual Project Directors and all program include an allowance f o r design, construction, and operation of a reactor experiment and a prototype t o demonstrate f e a s i b i l i t y as a commercial power plant. The programs presented each require i n excess of one hundred million d o l l a r s and ten years of concentrated e f f o r t .


-8After reviewing t h e problems, t h e Task Force believes t h a t any program of development on f l u i d f u e l r e a c t o r s should have as its objective the demonstration of technical f e a s i b i l i t y . The p r a c t i c a l i t y of these concepts depends on the a b i l i t y t o construct these p l a n t s i n a maintainable configuration at a competit i v e cost. Therefore, the programs should include extensive design and c o s t s t u d i e s of a representative large-size plant, with p a r t i c u l a r emphasis on remote maintenance of primary and a u x i l i a r y system components. Simultaneously, t h e program should be directed a t solving the research and development problems. U n t i l these major u n c e r t a i n t i e s are resolved, however, no large-size prototypes should be constructed.

Any of the concepts t h a t a r e pursued should follow t h e l o g i c a l sequence of experiment and prototype before a commitment t o cons t r u c t a large-scale r e a c t o r is made.


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IV. XEFERESICE REACTOR SYSTEMS "he b o j e c t Directors were requested t o furnish as complete information as possible on a reference r e a c t o r system design optimized f o r power production, Brief descriptions of these reference designs We presented below, More d e t a i l e d information can be I X , X, and XI. found in t h e project description s e c t i o n s

-

A.

Molten Salt Reactor

The reference design molten salt r e a c t o r is an moa-8 vessel containing a graphite assembly 12.25 f e e t i n diameter by 1 2 2 5 f e e t high, through which molten salt f u e l flows in v e r t i c a l channels, The f u e l salt i s a s o l u t i o n composed of 0.3 mole percent UF4, 13 mole percent ThF4, 16 mole percent b F 2 , and 70.7 mole percent Li7F,, The fuel salt i s heated from 1075OF t o 1225OF i n the core and I s c i r c u l a t e d from the r e a c t o r vessel t o four primary heat exchangers by four f u e l pumps. A barren coolant s a l t is used ~ 1 8 t h e intermediate heat exchange f l u i d , which superheats and r e h e a t s steam i n a Loeffler b o i l e r system, The r e a c t o r develops 760 PW of heat, By using 2000 psig, 1000/100OoF steam i n a reheat cycle, t h e n e t e l e c t r i c a l output is 318 MWE.

B.

!&quid Metal Fuel Reactor

The reference design l i q u i d metal f u e l r e a c t o r i s fueled with a s l u r r y c o n t s n i n g approldmately 3 w/o ThO2-UO2 i n bismuth, The r e a c t o r core vessel is 14 f e e t i n diameter and 14 f e e t high, cont a i n i n g a graphite core and r e f l e c t o r assembly, The f u e l s l u r r y is heated from 7W째F t o 10W째F through channels i n t h e r e a c t o r core and c i r c u l a t e d through three primary heat exchangers by t h r e e primary, variable speed pumps. The r e a c t o r develops 825 MW of heat. Sodium is used as the intermediate heat transfer coolant, The steam from the secondary heat exchangers supplies a non-reheat, 2090 psi, IOOO*F steam turbine which produces 312 MW of n e t electricity. C.

-

Aqueous Homoaeneous Reactor

Three concepts of aqueous homogeneous r e a c t o r s were presented t o the Task Force: a two-region, s o l u t i o n core, s l u r r y blanket r e a c t o r ; a two-region, s l u r r y core, s l u r r y blanket reactor; and a one-region s l u r r y reactor. The f u e l c a r r i e r and moderator is


- 10 heavy water i n a l l cases, The Project Director chose the two-region, solution core, s l u r r y blanket as t h e reference design reactor. This r e a c t o r c o n s i s t s of a 4 f t . diameter by 1 2 f t , long zirconium alloy core tank i n an 8 f t , diameter by 16 f t , long pressure vessel, A f i v e g/l s o l u t i o n of uranyl sulphate i n heavy water i s circulated through t h e core of t h e r e a c t o r and a t h o r i a s l u r r y containing loo0 g Th/l i s circulated through t h e blanket, Heat is removed from t h e core solution i n two c i r c u l a t i n g systems, each of which contains a steam generator and a c i r c u l a t i n g pump. The blanket s l u r r y i s recirculated through one similar heat removal c i r c u i t . Steam i s generated a t 400 p s i a and 435째F and s e n t d i r e c t l y t o t h e turbogenerator. The heat generation rate f o r one r e a c t o r i s 380 MWT: 320 MUT i n t h e core and 60 MWT i n t h e blanket. Three r e a c t o r s provide steam f o r one turbogenerator i n t h e reference s t a t i o n t o produce 317 MWE n e t , The slurry-fueled r e a c t o r s are similar but l a r g e r , s o f o r each type one 1140 FWT r e a c t o r satisfies the requirements of t h e reference s t a t i o n . Further details of the other systems are included i n the project description, Section X I ,

I

a


- 11 -

V,

COMPARATIVE ANALYSES OF PRESENT STATE: O F DEVF,LODID$T AND TECDJICAL FEASIBILITY

The following s e c t i o n presents an a p p r a i s a l of t h e problems which must be solved before a successful r e a c t o r can be b u i l t and operated, It should be noted t h a t considerably g r e a t e r research and development e f f o r t has been applied t o the aqueous homogeneous power r e a c t o r concept than t o the other two. A s a consequence, i t is l i k e l y t h a t a g r e a t e r proportion of i t s problems have been uncovered than f o r the others. Since these problems have been noted and weighed i n the following analyses, i t i s possible t h a t t h i s concept has been judged more severely than the others. KO a l t e r n a t i v e treatment has been deemed possible,

A.

-Fuels

I

and Materials

For the reference reactor designs presented f o r evzluation t o the Task Force, two-region s o l u t i o n s l u r r y aqueous homogeneous, onethe region s o l u t i o n molten s a l t , a n d one-region (slurry l i q u i d metal molten salt system i s judged t o have fewest f u e l and material problems with respect t o both the current s t a t e of technology and t h e development e f f o r t required before technical f e a s i b i l i t y can be established. The following t a b l e lists estimated r e l a t i v e technical d i f f i c u l t i e s f o r the t h r e e systems. Equal weight i s given t o f u e l and materials, with the l a t t e r broken down t o materials used f o r r e a c t o r i n t e r n a l s , and a l l other materials, The l a r g e r t h e number rating, the smaller is t h e estimated p r o b a b i l i t y o f technical success.

-

L

Xelative Technical D i f f i c u l t i e s

-

--

AHR Solution Core S l u r (Reference)

Mat er lials

I n t e r n a l s (Graphite & Z i r caloy F x t e r n d s (SS, INOR-8, Croloy

Fuel

Total Materials Summations

um - -

MSR

?a.

5

2

2

3 -

-

8

4

-2-

5

1 3 2 -

9

5

15

2

1

3 4 4 8

-

i

s


- 12 -

e

Discussion: 1

1. Materials a.

4

Internals

(Graphite f o r MSR & UiFR

-

Zircaloy f o r LW)

k

The graphite i n t e r n a l s of t h e LNFR have fewer problems than those of the f G R system. The reason f o r t h i s is that the f u e l i n a solution will have more tendency t o penetrate the graphite than will the f u e l i n a slurry. It must be demonstrated whether such penetrat i o n can be avoided and i f not whether adverse e f f e c t s on i r r a d i a t i o n will r e s u l t . These data can be obtained only i n long-time, in-pile loop tests. Accelerated t e s t s may give useful information but the e f f e c t of time under i r r a d i a t i o n may be equally or even more important than integrated flux.

t

The a b i l i t y of a graphite s t r u c t u r e 12-14 f e e t i n diameter by

12-14 feet high t o maintain i t s s t r u c t u r a l i n t e g r i t y for 30 years under conditions o f high flux, temperature gradients, hot spots r e s u l t i n g fron! f u e l collection i n "dead spots," e t c . can only be guessed. An important aspect o f this problem is the degree of engineering ingenuity t h a t is put i n t o t h e design. Despite these problems, t h e graphite i n LMFR is assigned a low r e l a t i v e d i f f i c u l t y , 1. Perhaps the graphite i n the MSR is twice as d i f f i c u l t a problem, due t o more probable penetration by f u e l (mentioned above), and t h e requirement t o avoid s e r i o u s reaction between oxygen and water (absorbed i n the graphite) and t h e fuel,

-

The Zircaloy core tank f o r the two-region s o l u t i o n core homogeneous reactor i s considered much more o f a problem than the graphite i n e i t h e r t h e MSR or MFR. Under conditions of t h e most recent design proposal, -0.025mU02SOq solution, 50 KU/l power density i n solution near wall, and with a core tank wall temperature of 26OOC (obtained by external cooling with t h e blanket s l u r r y plus careful hydrodynamic design i n s i d e ) t h e average corrosion r a t e expected under normal operating conditions and at a velocity of I5 ft/sec !This r a t e w i l l increase t o about i s approximately 15 mils/year. 17 rnils/year a t 5 ft/sec flow r a t e . '&en, f o r design purposes, a s a f e t y f a c t o r of two is applied t o these rates, the l i f e of t h e core tank i s about lyh years on t h e b a s i s t h a t i t is t o l e r a b l e t o corrode completely through a %inch w a l l .

-

Much more rapid penetration of this core vessel w i l l occur i f control of t h e conditions i s not good enough t o avoid inadvertant l o c a l heating. I f high temperature develops and causes deposition of uranium r i c h s o l i d s o r formation of the uranium r i c h second l i q u i d phase, f u r t h e r rapid increase i n l o c a l temperature w i l l occur. Rence, t h e r e l a t i v e l y poor r a t i n g of 5.

1

i i


b,

Externals

INOR-8(79% H i , 17% Mo, 7% C r , 5%Fe)$ t h e basic container and piping material f o r t h e molten salt system, appears t o o f f e r basica l l y a more trouble-free system than e i t h e r the LMFR o r AHR. Its corrosion and m a s s t r a n s f e r properties a r e excellent; no measurable a t t a c k ( l e s s than $? mil) occurring a f t e r one year exposure t o salt a t the m a x i m u m operating temperature. Weldability and f a b r i c a b i l i t y a r e comparable t o Inconel, The best r a t i n g , 1, w a s assigned t o the

m€l*

Although a less expensive material i s proposed f o r the LPIFR, .Z4 Crvsloy, there is a more d e f i n i t e problem with m a s s t r a n s f e r of metal f r o m hot t o cold regions of t h e r e a c t o r system. Additives have minimized but not eliminated the problem. It has not yet been shown whether this Croloy will be a s a t i s f a c t o r y material of cons t r u c t i o n f o r a d e l t a T of 167°C. If Croloy requires a lower d e l t a T for s a t i s f a c t o r y operation, other c o s t s w i l l obviously go up. I f a high s t r e n g t h s t e e l , clad with carbon s t e e l i s required, i t cannot be s a i d t h e system can be constructed i f i t is possible, the cost will be high. The lower r a t i n g of‘ 3 is therefore assigned.

-

In the AHR, s t r e s s corrosion cracking i s a worry in the presence of oxygen. Ia the presence of hydrogen, the s t a i n l e s s s t e e l is not as corrosion-resistant as i n oxygen-bearing s l u r r i e s , There i s a l s o an increased p o s s i b i l i t y of hydrogen embrittlement f o r s p e c i a l purpose materials. For the uranyl sulphate s o l u t i o n , the s t a i n l e s s s t e e l is dependent on r e l a t i v e l y t h i c k corrosion films, Although corrosion r a t e s a r e very low ( a f t e r i n i t i a l corrosion), t h e films must be maintained t o avoid rapid attack. I n addition, 347 s t a i n l e s s s t e e l i s notch-sensitive and poses 2 design problem. The r a t i n g given i s 3. 2,

fiels -

The f u e l f o r the molten s a l t system is considered superior t o t h e A X R and LMFR f u e l s , b a s i c a l l y because i t is a solution. Although i t shows promise of being s a t i s f a c t o r y f o r long cycle t i n e s before requiring reprocessing there are some u n c e r t a i n t i e s t h a t must be resolved, such as e f f e c t of oxygen on p r e c i p i t a t i o n of uranium and the e f f e c t of p r e c i p i t a t i o n of the noble f i s s i o n product metals, Ru, Mo, and Nb. A good r a t i n g is assigned.

Both t h e AHFt and LMFR have s l u r r y problems which can hardly be defined a t t h i s time. Slurry technology is i n its infancy and very considerable development e f f o r t is required before adequate information is obtained. It i s doubtful whether s l u r r i e s can have cycle times of the order of 10 years. I n the LKE’R the s l u r r y p a r t i c l e s


i \

t h a t are added t o the system consist of 30-50$ of U02, the balance Th02. For a long cycle time up t o 50% of these metal atoms will f i s s i o n , destroying completely the p a r t i c l e as such. It cannot even be guessed what this degradation will mean. The r a t i n g f o r the LMFR i s 4.

c

I n the AHR, a s e r i o u s problem i s a l s o encountered with the uranyl sulphate solution, Extreme care must always be exercised t o keep the solution within proper chemical control. Even a s h o r t time i n s t a b i l i t y i n the solution may r e s u l t i n burning a hole i n the Zircaloy core tank with the consequence of mixing the s o l u t i o n and s l u r r y fuels. "his s e r i o u s problem influences the r a t i n g number of both the core vessel and the f u e l ,

i

3.

Alternate AHR

Also included i n the t a b l e is a rating f o r the non-reference design, i n which t h e core f u e l is a slurry. The corrosion problem f o r Zircalvy seems t o be s u b s t a n t i a l l y reduced i n changing t o this concept although data are q u i t e meager. The b e t t e r r a t i n g of 2 i s assigned. Similarly, t h e corrosion resistance of stainless s t e e l i s improved, leading t o a r a t i n g of 2, s l i g h t l y b e t t e r than t h a t given f o r 2.4 Croloy i n LMFR. Elimination of the s e r i o u s d i f f i c u l t i e s of t h e s o l u t i o n h e 1 makes possible a r a t i n g of 5 f o r t h e f u e l , s t i l l somewhat worse than that for the LMFR because of the f a s t e r slurry s e t t l i n g rate. Considering only f u e l and materials, the s l u r r y core has s i g n i f i c a n t l y less d i f f i c u l t technical problems than the s o l u t i o n core. B.

Primary System and Components 1. Reactor

The differences i n t h e r e a c t o r concepts are d i c t a t e d by the physical and nuclear properties of t h e f l u i d fuels. The AHR uses heavy water a s the c a r r i e r f l u i d because of its good nuclear prope r t i e s b u t the system must be operated at r e l a t i v e l y high pressure t o maximize the thermal efficiency, The s i z e of t h e vessel i s limited by t h i s high pressure requirement and, f o r l a r g e power s t a t i o n s , multiple r e a c t o r units a r e used. For a single-region r e a c t o r , this s i z e l i m i t a t i o n also introduces thermal neutron shields. "he LMFR and PER u t i l i z e a l o w vapor pressure c a r r i e r but require the addition of a moderator. The HRE-2 experience has shown t h a t the hydrodynamic and heat t r a n s f e r conditions a t the surface of the container i n contact with


- 15 uranyl sulphate solution must be c a r e f u l l y controlled t o prohibit phase separation and eventually burnout of the i n n e r container. In t h e reference concept the temperature of the i n n e r container i s controlled by removing heat with the slurry f l u i d flowing on the outsi.de and with the i n l e t solution f l u i d flowing along the i n s i d e of t h e w a l l . The eventuality of credible accidents (e.g. l o s s of slurry flow) d i c t a t e s t h a t the solution f u e l be capable of removing a l a r g e p a r t of the heat generated i n the inner w a l l ; t h i s requirement probably decreases the allowable t e m p r a t u r e of the solution f u e l and may f u r t h e r decrease an already u n a t t r a c t i v e thermal effici-ency. A considerable development e f f o r t has already been expended on t h e problems associated with the hydrodynamic and heat t r a n s f e r condi.tions of s o l u t i o n f u e l s i n spherical configurations. It has been shown t h a t an acceptable design f o r a s p h e r i c a l container t o accommodate s o l u t i o n f u e l can be developed f o r steady-state operation,, Higher t o t a l power a t given maximum power density i s a t t a i n a b l e i n t h e c y l i n d r i c a l configuration suggested f o r t h e two-region, solution-slurry reactor. Very l i t t l e development work has been done on c y l i n d r i c a l configurations u t i l i z i n g s o l u t i o n f u e l s and much more remains t o be done.

In addition t o the problems associated with the t r a n s i e n t operations, i t is highly probable t h a t t h e i n n e r container made of Zircaloy-2 w i l l have t o be replaced 10 t o 15 years after s t a r t u p as a r e s u l t of the high corrosion r a t e by the s o l u t i o n fuel. This likelihood makes the use of the reference design reactor questiona b l e f o r c e n t r a l power s t a t i o n i n s t a l l a t i o n s .

The s p e c i f i c a t i o n of a single-region o r two-region r e a c t o r u t i l i z i n g s l u r r y f l u i d s only, cannot be made a t t h i s time. Satisf a c t o r y materials for the long-term containnent o f s l u r r y fluids i n a r e a c t o r vessel, assuming proper design, are now available. However,, the hydrodynamic and heat t r a n s f e r c h a r a c t e r i s t i c s of s l u r r y f l u i d s i n several configurations are needed before the r e a c t o r vessel can be specified. The LMFR and MSR concepts have problems which are s i m i l a r t o each other. The foremost of these i s t h e requirement t h a t the vessel and i t s i n t e r n a l s be heated t o a temperature i n excess of the melting point of the f u e l c a r r i e r i n an i n e r t atmosphere. The s e v e r i t y of the preconditioning problem i s g r e a t e r i n t h e MSR because of the higher temperature and g r e a t e r need t o minimize oxygen contact with the f u e l f l u i d . However, t h i s difference appears t o be i n s i g n i f i c a n t . Careful design of the heating system i s d i c t a t e d t o insure s u f f i c i e n t heating and acceptable thermal s t r e s s e s i n t h e r e a c t o r vessel and its i n t e r n a l s .

t

k


- 16 The moderator chosen f o r t h e U l F R and I4SR i s high density, impervious graphite. It i s necessary t o provide support f o r the graphite cylinder and means by which the blocks are held firmly; these problems may involve considerable design e f f o r t and an acceptable design i s expected to evolve from r e a c t o r experiment experience. Fuel and f i s s i o n products may be deposited on and absorbed by the graphite i n both the L J P R and fSRm The magnitude of t h i s problem has been roughly defined f o r both r e a c t o r s but the r e s u l t s obtained t o d a t e do not permit f i r m conclusions. The extent of the phenomenon and its possible consequences need t o be examined experimentally and a n a l y t i c a l l y . Several i n - p i l e loops a r e indicated; the cost of these i n v e s t i g a t i o n s i s expected t o be of the order of two million d o l l a r s ,

E

The physical properties and heat transfer c h a r a c t e r i s t i c s of the solution f u e l suggested f o r the PER have been determined experimentally. Xoweves, i t appears prudent t o determine the hydrodynamics of the p a r t i c u l a r r e a c t o r design i n mockup s t u d i e s , This work need not be extensive and can be accomplished for a modest expenditure. Hydrodynamic i n v e s t i g a t i o n s a r e required f o r the bismuththoria-uranium oxide s l u r r y suggested f o r t h e L%.IFia. Extensive design work f o r the INiQ r e a c t o r is indicated because of the unknown properties of t h e slurry. It i s conceivable t h a t s l u r r y hold-up i n localized areas within t h e r e a c t o r may cause overheating a d s p a l l i n g of the graphite,

It appears t h a t both the Lsy";FR and IilSR may require control rods f o r successful operation. The AHR does not require them. The need f o r control rods i s a disadvantage but not a s e r i o u s one. It i s apparent t h a t additional development work must be performed f o r a l l three concepts. Since the f u e l f l u i d f o r the 6 R is a solution, i t appears t h a t design o f a vessel f o r FfiR w i l l be most e a s i l y accomplished. 2.

Primary System

i

F

The r e l a t i v e merits of the primary systems can be evaluated on t h e b a s i s of the properties o f t h e f u e l c a r r i e r s . From this s t a d point the FIR f l u o r i d e salts have the advantage of high volumetric thermal capacity combined with a r e l a t i v e l y low density and low vapor pressure

.

The UIFR uses bismuth as the basic f u e l c a r r i e r . The vapor pressure i s low; however, the density is approfimately 2??t h a t of the fluoride salts and t h e volumetric heat capacity i s approximately l'3 t h a t of the fluoride salts,

L

i


- 17 Because of t h e high density of the f u e l c a r r i e r and the low thermal capacity, the LEviFR design i s highly vulnerable t o changes i n allowable temperature difference i n the loops, The r e s u l t i s a design based on a high difference. The consequence of decreasing t h e temperature difference i s t o increase the pumping horsepower requirement, This increase rapidly pushes the pumps beyond the present and a n t i c i p a t e d f u t u r e available canned motor c a p a c i t i e s and would require a d d i t i o n a l loops The MSR on tize other hand uses t h e high thermal capacity t o obtain a low temperature r i s e across the reactor and hence a high mean temperature difference i n the primary heat exchangers, There still. remains f l e x i b i l i t y i n t h e design parameters within the l i m i t s of p r a c t i c a l i t y should i t prove necessary t o drop the mean temperat u r e difference i n the exchangers and r a i s e the r e a c t o r temperature difference The s o l u t i o n s o r s l u r r i e s used f o r the aqueous concepts have roughly the same volumetric thermal capacity as t h e f l u o r i d e s a l t s but s u f f e r from the high vapor pressures of water a t the temperatures required f o r power production. I n addition, t h e use of s l u r r i e s i n t h e two-region breeder concepts involves a d i f f i c u l t technology and presents a s e r i o u s b a r r i e r t o the development of components, p a r t i c u l a r l y f u e l c i r c u l a t i n g pumps. The need f o r smaller piping i n t h e case of t h e MSii tends t o compensate f o r t h e higher temperatures i n the thermal stress analysis. There i s l i t t l e t o choose between the two concepts on t h i s b a s i s although t h e LlylFR designers have apparently given more thought t o t h i s p a r t i c u l a r phase of the concept. This smaller piping a l s o o f f s e t s t h e higher melting temperature of the molten salt, Thus the o v e r a l l system preheating requirements a r e about equal f o r the two high temperature concepts. The major Components of all three concepts are pushing the f r o n t i e r s of t h e technologies involved. !?he f a i l u r e t o solve any of the s p e c i f i c problems therefore jeopardizes t h e technical feasib i l i t y of t h e program, To d a t e none of t h e concepts have demonstrated technical f e a s i b i l i t y of u t i l i t y - s t a t i o n - s i z e d primary system components, however, the s i z e problem can be minimized by using a d d i t i o n a l l o o p s at some s a c r i f i c e of economic p o t e n t i a l , It is believed t h a t , i n general, t h e technical f e a s i b i l i t y of t h e components of t h e primary systems of t h e three concepts depends on matters of engineering ingenuity and a r e therefore evaluated as e s s e n t i a l l y equal,

3.

Primary Auxiliary System

The success of any of these concepts n e c e s s i t a t e s the performance of functions by equipment not d i r e c t l y a p a r t of the primary


- 18 heat t r a n s f e r loop. Some of t h i s hardware i s i n d i r e c t contact with highly radioactive f l u i d s ; other p a r t s of i t may become radioactive under some conceivable circumstances. -411of t h i s equipment must be considered t o be highly contaminated, and w i l l require remote maintenance, Both t h e LfG!i? and PER concepts s u f f e r a s e r i o u s disadvantage as a r e s u l t of the r e l a t i v e l y high melting point of the carrier materi a l s . All of the primary system must be heated by an e x t e r n a l source. The magnitude of t h e external source i s of t h e order of 3 MW f o r both systems and w i l l be about the same f o r both since t h e LMFR and MSR require temperatures of about 6 5 0 0 ~and 1050째F, respectively, but t h e ISQR volume is about 1.7 times t h a t of t h e MSB. H e l i u m gas c i r c u l a t i n g i n annuli about a l l primary equipment is used as a heating medium f o r both r e a c t o r systems; some other gas may be acceptable and an e f f o r t t o eliminate the use of helium should be made because of its high p r i c e (&44/1000 c f at STP) and l i m i t e d supply, The quantity of helium c i r c u l a t e d is of t h e order of 3000 pounds per minute and moderate s i z e d blowers a t about 75 psig pressure are the prime movers. The blowers, furnaces, and piping must be c a r e f u l l y designed and constructed t o eliminate wide temperature gradients i n the primary equipment and leakage of t h e helium, It is l i k e l y t h a t t h e c i r c u l a t i n g gas w i l l become contaminated during operation, Therefore, provision f o r t h e remote maintenance of t h e blowers, furnaces and piping, and valves must be made; t h i s provision complicates the design considerably. Design of t h i s heating system i s not impossible but i t i s expected t h a t an acceptable design will only evolve from the design and construction of the a c t u a l equipment, Either a corrosive s o l u t i o n o r an erosive slurry-thoria i n heavy water--or both flow through t h e primary a u x i l i a r y equipment of t h e AHR. Equipment f o r handling uranyl s u l f a t e s o l u t i o n has been found. Some o f the equipment must be titanium-lined t o withstand the corrosive s o l u t i o n ; t h i s equipment has not been fabricated commercially i n l a r g e s i z e s but such p r a c t i c e s can probably be developed. The experience with the s l u r r i e s has shown t h a t i t Will be d i f f i c u l t t o obtain t h e operating l i f e of some of the equipment ( p a r t i c u l a r l y check valves) which i s acceptable f o r use i n a c e n t r a l power s t a t i o n , D i f f i c u l t problems have been encountered and t h e r e s u l t s from some development work have not been encouraging. Addit i o n a l development work i s required and t h e i n d i c a t i o n s are t h a t t h e susceptible equipment w i l l have t o be eliminated i n s o f a r as i s possible by desigx and by changing and complicating t h e mode of operation, The primary a u x i l i a r y systems of t h e IYlFR and MSR a r e characterized by l a r g e , but r e l a t i v e l y simple, equipment whereas the AHR system involves s e v e r a l items of r e l a t i v e l y small s i z e and of s p e c i f i c function. Remote maintenance procedures f o r the LMFR and MSR have

t L


- 19 n o t been demonstrated adequately and f u r t h e r development work is required. The problems of remotely maintaining t h e primary auxili a r y systems f o r the D P R and MSR are e s s e n t i a l l y t h e same as those involved with the primary heat t r a n s f e r loop. The AHR primary auxi3.iary equipment can be maintained, as has been demonstrated by HRT experience, The advantages and disadvantages are about equal for the t h r e e concepts. The distinguishing f e a t u r e is the ease or d i f f i c u l t y with which the d i f f e r e n t problems may be solved. In t h i s respect it i s expected that t h e AHR s l u r r y will use a low pressure system which i s operated batchwise i n an e f f o r t t o minimize t h e problems caused by erosi.on. It is expected t h a t t h e development of remote maintenance techniques will be successful but a l s o d i f f i c u l t . It appears t h a t t h e r e is l i t t l e t o d i s t i n g u i s h t h e d i f f e r e n t concepts with regard t o the primary a u x i l i a r y systems.

4.

Instrumentation

The nuclear instrumentation associated d i r e c t l y with r e a c t o r operation i s e s s e n t i a l l y the Same for a l l three concepts. The s t a t e of t h i s technology is highly developed and additional work d i r e c t e d t o a i d the operation of these t h r e e r e a c t o r s does not appear t o be required. The primary loop and primary a u x i l i a r y l o o p instruments for the Pfeasurement of a nuniber of temperatures and several l i q u i d l e v e l s appear r e q u i s i t e i n the MSR and IslIFR while i n t h e AHR fewer of these measurements and flow rate and pressure measurements are needed. All of these i n s t r u menta are exposed t o intense neutron and gamma f i e l d s but t h i s is not an insurmountable problem. All instruments must be designed t o permit remote replacement; the instrument design for t h e MSR and LMFR i s more d i f f i c u l t s i n c e the outer containment, heating and cooling shell. must be penetrated.

MSR and Wl2 d i f f e r from those needed f o r the AHR.

Jacketing of t h e primary loop and some primary a u x i l i a r y equipment complicates t h e l e z k detection system, and i a s t k m e n t a t i o n for t h i s purpose is deemed necessary f o r t h e MSR and LMFR. The i n s t r u mentation t o be used for this purpose has not been designed, but t h e r e are s e v e r a l p o s s i b i l i t i e s . The need f o r t h i s equipment i s a d e f i n i t e disadvantage

.

Some instrument development f o r both t h e MSR and LMFR i s necessary and t h e extent of t h i s development will become evident during design and operation of a r e a c t o r experiment. Development work has been needed for t h e HRT and i t has been q u i t e successful. Similar experience with t h e other r e a c t o r concepts can be expected.


- 20 Instrumentation requirements of the AHR are most e a s i l y met but mainly as a r e s u l t of a successful development e f f o r t i n connection The instrumentation requirements f o r the MSR and with the -2. LgiIFR a r e nearly the same.

C.

Operation

k

The primary purpose of any power producing scheme i s the output of e l e c t r i c a l power at t h e generator terminals, I n a t t a i n i n g this objective, r e l i a b i l i t y and ease of operation of the plant a r e two of the most important factors. Both r e l i a b i l i t y and ease of operation depend on the design of the plant. Because a given plant is only a p a r t of a much l a r g e r system, its operation will be affected by the operation of t h e e n t i r e system, From the point of v i e w of operating personnel, the people who will l i v e with the plant a f t e r i t i s constructed, the r a p i d i t y and simplicity of start-up and shut-down operations i s very important, Squally important i s the ease with which the plant w i l l respond t o changes i n e l e c t r i c a l load. During the e a r l y l i f e of t h e plant the necessity of shut-downs and start-ups w i l l be d i c t a t e d l a r g e l y by t h e need,s of maintenance, During low system demands, at night, and weekends the plant would carry a somewhat reduced load. As the plant becomes older and more e f f i c i e n t p l a n t s are added t o the system i t w i l l no longer be base loaded. A t this point i n the p l a n t g s l i f e , perhaps as l i t t l e as ten rears, i t may be necessary t o shut down during low demand periods, The molten salt and l i q u i d metal systems have the same basic problems i n s t a r t i n g up and s h u t t i n g down, both need a preheating and cooling system. For both t h e molten salt and l i q u i d metal system the time required in a s t a r t - u p operation will be determined by t h e t i m e required t o preheat the systems, It may be found i n f u t u r e operations of these p l a n t s as p a r t of a l a r g e e l e c t r i c a l system t h a t i t c o s t s more t o shut down and start up than t o operate at a very low load. This would be a disadvantage f o r e l e c t r i c a l systems with very low periodic loads. A possible a l t e r n a t i v e would be t o hold the r e a c t o r j u s t c r i t i c a l so the decay heat is equal t o t h e system heat losses, All t h a t would be required t o put t h e plant For t h e back on system would be t o start up the turbo-generator. aqueous homogeneous r e a c t o r system, the shut-down and start-up procedures do not present any g r e a t problems, The r e a c t o r plant is made s u b c r i t i c a l by d i l u t i n g the f u e l and t h e rate at which the temperature can be reduced i s determined by t h e permissible cooling rate o f the system components; likewise, t h e start-up time is determined by t h e permissible heating r a t e of t h e components.

In comparison of the three systems from the standpoint of shutdown f o r maintenance and start-up, the aqueous homogeneous is by f a r

P

t

t


- 21 t h e simplest; the degree of d i f f i c u l t y is the same f o r both t h e mo1te:n salt and l i q u i d metal r e a c t o r concepts.

Two of t h e proposed concepts have had the advantage of conducting r e a c t o r experiments, the Aircraft Reactor Experiment (ARE) and t h e aqueous homogeneous r e a c t o r s (HE-1 and HRE-21, both using s o l u t i o n fuels. From t h e standpoint of o p e r a b i l i t y t h e experiments were a success. The operation of the aqueous homogeneous r e a c t o r is complicated by the need of very extensive a u x i l i a r y and supporting systems. "he r e l i a b i l i t y of r e a c t o r operation w i l l therefore depend on the proper functioning o f these systems, Control rods a r e not necessary on t h i s type r e a c t o r due t o the l a r g e negative temperature coefficient. A s e r i o u s problem with the solution type aqueous homogeneous r e a c t o r has been t h e phenomenon of l a r g e power surges during operation. This is due t o f u e l coming out and going back i n t o s o l u t i o n i n the core.

.

The operation of a s l u r r y type r e a c t o r has yet t o be demonstr a t e d The A i r c r a f t Reactor Experiment demonstrated the f e a s i b i l i t y of operation of a high temperature molten salt reactor. Although the r e a c t o r w a s equipped with control rods i t was found t h e r e a c t o r would follow changes i n the heat extraction system due t o t h e negative temperature c o e f f i c i e n t of r e a c t i v i t y . For any large-scale plant, control would be a combination of the negative temperature c o e f f i c i e n t , movement of control rods, and v a r i a t i o n of f l u i d flow. The l i q u i d m e t a l f u e l r e a c t o r is a t a disadvantage by not having However, the operation of the r e a c t o r would probably be very similar t o t h a t of the molten salt,

had a r e a c t o r experiment.

The three concepts should be compared on the basis of reliable on-system operation i n which t h e plant is expected t o meet load changes and t o operate f o r extended periods of time. A l l three conc e p t s could equally w e l l meet changes i n load on the turbogenerator. The differences, i f any, between the t h r e e would be the r a t e a t which t h e changes could be met. I n the case of r e l i a b i l i t y of operation f o r extended periods, the aqueous homogeneous i s at a disadvantage because of its extensive a u x i l i a r y systems which a r e more complex than e i t h e r of the other two concepts. Minor operational problems i n these a u d l i a r y systems could lead t o a shut down of the plant.

D.

Maintenance

Mai ce w a s ized as one of the most important f a c t o r s influencing the p r a c t i c a b i l i t y of any of the three concepts. The combination of unknown s e r v i c e l i f e of components,


- 22 s e r i o u s consequences r e s u l t i n g from release of r a d i o a c t i v i t y due t o a f a i l u r e , and extreme d i f f i c u l t y of hot maintenance d i c t a t e t h e p l a n t arrangement, containment systems, and component design, The molten salt and l i q u i d metal systems o f f e r t h e same basic problems i n regard t o plant layout and maintenance, with t h e possib i l i t y t h a t t h e l i q u i d metal system may be somewhat more d i f f i c u l t t o maintain due t o problems created by s l u r r i e s i n t h e system. This could be balanced out i f appreciable air were t o leak into the molten salt system, causing t h e p r e c i p i t a t i o n of UO from t h e fuel. 2 The maintenance concepts proposed f o r t h e above two systems can be used interchangeably i n t h a t the same fundamental problems and conditions are prevalent i n both systems. Major equipment components, i.e. pumps and heat exchangers, appear t o have been a n m g e d s o t h a t replacements can be made without undue d i f f i c u l t y ; howzver, i t i s f e l t t h a t problems of instrumentation f o r pressure, temperature control, and l e a k control of t h e e n t i r e system have not been adequat e l y solved from the standpoint o f maintenance of this part of the systern.

I n layouts evaluated f o r LMFR and MSR, equipment has been stacked on s e v e r a l l e v e l s , one above t h e other, This, i n turn, l e a d s to dependence upon t h e e f f e c t i v e performance of remotely operated robot vehicles, manipulators, t e l e v i s i o n , and o t h e r controls t o accomplish maintenance i n the lower l e v e l s of the plant, This concept r e q u i r e s considerable spacing of t h e equipment t o make c e r t a i n t h a t t h e robot c a r t s have complete access t o a l l parts. I f this concept does not work, t h e only apparent s o l u t i o n s would be t o spread t h e p l a n t out on a horizontal plane s o t h a t complete v i s i b i l i t y and access could be achieved from overhead o r t o i n s t a l l heavy viewing windows a t each l e v e l and manipulators t o provide access t o each segment of the equipment. In e i t h e r case the c a p i t a l cost of t h e p l a n t would be increased considerably. ' f i l e i t i s expected t h a t t h e i n t e g r i t y of the piping systems i n these concepts will be such t h a t the frequency of remote weld-ing will be low, t h e l i f e of t h e 2lant could w e l l depend upon t h e a b i l i t y t o remotely weld and inspect pipe of l a r g e size. A remotely operated welding machine h a s been designed, b u i l t , and has undergone t e s t s , Considering the r e l a t i v e l y s h o r t time t h e machine has been under development, the progress has been encouraging. Methods f o r remote inspection of welded j o i n t s are under investigation but no satisfactory solution has been demonstrated. The l i q u i d metal, molten s a l t , and aqueous homogeneous concepts are not f e a s i b l e unless remote welding o r s u b s t i t u t e methods of closure are perfected, On t h e p o s i t i v e s i d e , the molten salt and t h e l i q u i d metal systems have some advantage over the aqueous homogeneous system.


While at operating temperatures they have a l l the c h a r a c t e r i s t i c s of t h e l i q u i d systems, When the salt and metal are allowed t o cool down and freeze, they take on the c h a r a c t e r i s t i c s of a s o l i d f u e l i n t h a t most of t h e radioactive f i s s i o n products will be frozen i n t h e metal o r salt. This w i l l minimize t h e problem norma l l y expected i n aqueous systems, t h a t f i s s i o n products w i l l gradually contaminate l a r g e areas of the plant including t h e areas above t h e c e l l blocks, t h e maintenance tools, t h e operating c e l l s , and, ultimately require elaborate f i l t r a t i o n systems t o prevent radioactive p a r t i c u l a t e matter from being discharged from t h e s t a c k t o t h e surrounding environs, The aqueous homogeneous system requires more equipment, due t o t h e necessity of having three r e a c t o r systems t o achieve t h e same power production as one reactor i n t h e other two systems, It a l s o has the necessity of doing more frequent reprocessing o f the f u e l , with associated on-plant reprocessing f a c i l i t i e s . The aqueous system has one simplifying feature i n t h a t pipe and vessel jacketing are not needed as i n the other two concepts, This jacketing increases several-fold the magnitude of the maintenance job .in the l i q u i d metal and molten s a l t systems. The aqueous homogeneous reactor concept has other basic requirements necessitating additional maintenance: gas recornbiners, D20 recovery, d a i l y (or frequent) operation of reprocessing and waste systems, and the much higher operating pressure, Based on present knowledge from t h e two reactor experiments of t h e aqueous homogeneous concept, i t has been demonstrated t h a t maintenance is possible, a l b e i t time consuming, while t h e molten salt and l i q u i d metal systems have yet t o be demonstrated and, i n fact, depend upon t h e successful completion of extensive development programs. I n conclusion, based on today's knowledge and t h e experience of a properly designed aqueous homogeneous system can be maintained by t h e use of w e t maintenance, although a t high cost and considerable d o m time, The use of a dry maintenance scheme f o r t h e molten salt, l i q u i d metal, and aqueous homogeneous s y s t e m depend upon the s a t i s f a c t o r y development of remotely operated maintenance equipment. The degree of d i f f i c u l t y and cost of maintenance between t h e three systems w i l l depend upon t h e plant layout and the maintenance scheme adopted. EIBE-2,

E.

Control

All t h r e e r e a c t o r concepts r e l y upon a gross negative temperature c o e f f i c i e n t of r e a c t i v i t y as the primary means of controlling t h e r e a c t o r power. Through this inherent feature some load changes are r a p i d l y and s a f e l y accommodated, The W R concept has included


- 24 variable speed primary pumps t o minimize wide and r e l a t i v e l y rapid changes i n r e a c t o r temperatures, 'While conservation of design may have been t h e reason f o r including t h i s feature, i t appears t h a t t h e need f o r variable speed control i s more acute i n the I"2 than i n the MSR. The temperature r i s e across the r e a c t o r at normal operating conditions is 30째F i n LMFR VS. 150째F i n FGR and the magnitude of t h e thermal shocks t h a t could be im2osed upon t h e system are larger. The temperature c o e f f i c i e n t i n the AHR i s about 100 times g r e a t e r than t h a t i n e i t h e r t h e LMFR o r MSR. Reactivity changes i n t h e order of 2 percent A k/k per second can be controlled s a f e l y i n AHR, Control o r s a f e t y rods are not included i n AHR but are indicated f o r UIFR ' and MSR. For concepts u t i l i z i n g s l u r r y f u e l s i t has been calculated t h a t i t i s possible t o operate ( a t constant temperature) with s l u r r y concentration such t h a t any v a r i a t i o n i n the concentration w i l l cause s u b c r i t i c a l i t y . These calculations have not included t h e e f f e c t s of l o c a l i z e d v a r i a t i o n s of s l u r r y concentrations; these e f f e c t s can be ascertained by r e a c t o r experiments only, There are s e v e r a l unanswered questions with regard t o control of t h e blSR and they, too, w i l l require i n v e s t i g a t i o n by operating a r e a c t o r experiment.

I n t h e operation of the AKR, f u e l f l u i d s may be continuously removed and added to t h e primary heat exchange loops. T h i s arrangement permits rapid v a r i a t i o n of f u e l concentration t o accommodate wide changes i n load and i n s u r e s m a x i m u m thermal e f f i c i e n c y during peak load periods. The high temperature reactor concepts do not have as rapid f l e x i b i l i t y i n accommodating load changes and i t may be desirable t o use control rods f o r t h i s purpose. The aqueous homogeneous r e a c t o r experiments have demonstrated t h a t r e a c t o r s of t h i s type can be controlled easily. The FIR and LMFX have not had t h i s opportunity. F.

Xazards

The f3IR i s p o t e n t i a l l y the most hazardous of t h e three f l u i d f u e l concepts. The primary reason f o r t h i s is t h a t it i s a high pressure system. Additional contributing f a c t o r s are t h e radiol y t i c gas explosion hazard, and the f a c t t h a t i t s f u e l s t a b i l i t y i s probably t h e most precarious of t h e t h r e e concepts. The two non-aqueous systems are q u i t e s i m i l a r i n t h e i r s a f e t y c h a r a c t e r i s t i c s , Neither Po-210 build-up i n t h e LMFR nor t r i t i u m production i n the MSR are considered overly s e r i o u s problems. There

i

F


- 25 i s a possible complication r e s u l t i n g from the use of a s l u r r y i n the

urn.

A possible design l i m i t a t i o n from t h e point of view of hazards is operation at very ‘low values of the e f f e c t i v e delayed neutron f r a c t i o n , beta e f f . It remains t o be demonstrated t h a t s a t i s f a c t o r y opera.tion can be obtained with extremely low values of beta eff. The homogeneous core, as opposed t o the core with coolant passages, is more susceptible t o nuclear i n s t a b i l i t i e s a r i s i n g from hydrodynamic f l o w i r r e g u l a r i t i e s

The AHR, though i t seems t o have more p o t e n t i a l hazards, has the advaxtage of being able t o point t o e x i s t i n g experience, which often i s reassuring

Discussion -: Fluid f u e l r e a c t o r s are c h r a c t e r i z e d by the requirement t o pump, sampl-e, and otherwise handle q u a n t i t i e s of highly radioactive l i q u i d s (or Ejlurries). This means t h a t standards of i n t e g r i t y must be quite high and t h a t precautions t o prevent leakage o f t h e f u e l s should be taken t o t h e m a x i m u m extent commensurable with t h e requirement t o operate (from the points of v i e w of physical operation and of keeping the c o s t s t o l e r a b l e ) The hazards analysis i s complicated by the many possible variat i o n s within each of t h e f l u i d fuel concepts. The variations within each concept often d i f f e r considerably among themselves i n some of t h e most important c h a r a c t e r i s t i c s concerned with reactor safety. For example, they may d i f f e r i n such respects as:

(a)

Solution vs. s l u r r y fuel.

(b)

Continuous o r frequent removal of f i s s i o n products VS. long-term accumulation of f i s s i o n products within the system.

(c)

Control rods

(d)

The magnitudes and time constants of the various components of the temperature c o e f f i c i e n t of reactivity.

(e)

!Be e f f e c t i v e delayed neutron fraction.

VS.

no control rods.

As a consequence of these variations, it is often d i f f i c u l t t o make generalizations applicable t o each of the concepts as a whole.


- 26

1, Nuclear S t a b i l i t y Hany t h e o r e t i c a l analyses of the s t a b i l i t y of c i r c u l a t i n g f l u i d f u e l systems have been made. The r e s u l t s have invariably been t h a t the systems would not s u f f e r s e r i o u s l y from this cause, I n those cases i n which t h e power a t which the system would become unstable w a s determined t h e o r e t i c a l l y , i t w a s found t o be man3 times t h e normal power, The t h e o r e t i c a l s t u d i e s have t o some extent been v e r i f i e d by experimental i n v e s t i g a t i o n s i n t h e HRE-1, H E - 2 , and ARE, Even though these i n v e s t i g a t i o n s have been e s s e n t i a l l y reassuring s o f a r as inherent s t a b i l i t y is concerned, they have had t h e i r disquieting 1 had r e a c t i v i t y v a r i a t i o n s due t o flow p a t t e r n features, v a r i a t i o n s i n t h e s i z e of c e n t r a l flow vortex. I n HRE-2, t h e r e have been r e a c t i v i t y o s c i l l a t i o n s , due i n all p r o b a b i l i t y t o i n s e r t i o n of p r e c i p i t a t e d f u e l i n t o the system (and hence, not a proper mechani s m within t h e d e f i n i t i o n of inherent nuclear s t a b i l i t y ) , The ARE showed good nuclear s t a b i l i t y c h a r a c t e r i s t i c s ,

t

From the above t h e r e i s l i t t l e reason t o believe t h a t inherent i n s t a b i l i t y w i l l become a s e r i o u s technical f e a s i b i l i t y question, However, t h e t h e o r e t i c a l s t u d i e s have been d e f i c i e n t i n t h a t t h e e x t e r n a l feedback mechanisms were o f t e n ignored o r t r e a t e d only s u p e r f i c i a l l y . The analyses which did consider e x t e r n a l feedback showed no s e r i o u s i n s t a b i l i t y . O f the t h r e e concepts, the AHR i s probably t h e most susceptible t o nuclear i n s t a b i l i t y . This is a consequence of its l a r g e negative temperature c o e f f i c i e n t which, due t o c i r c u l a t i o n effects, has a delayed component which may be a contributing f a c t o r t o i n s t a b i l i t y , Also, hydrodynamic f l u c t u a t i o n s a r e most l i k e l y i n t h e AHR because t h e flow p a t t e r n encompasses t h e e n t i r e core. I n t h e o t h e r systems t h e flow proceeds i n channels through the core, minimizing t h e p o s s i b i l i t y of important fluctuations. Proper hydrodynamic flow design will be an important problem i n the development of f l u i d f u e l reactors.

The experimental and t h e o r e t i c a l s t a b i l i t y i n v e s t i g a t i o n i n f u t u r e r e a c t o r experiments should be a c e n t r a l p a r t of the program, Experimentally, t h i s should probably include o s c i l l a t o r programs s i n c e t h e t h e o r e t i c a l a n a l y s i s i s then more amenable t o q u a n t i t a t i v e conclusions. ‘There should a l s o be a more extensive t h e o r e t i c a l program for r e a l i s t i c a l l y Considering e x t e r n a l feedback, 2.

Nuclear Accidents

The main categories of nuclear accidents t o which these systems are subject are: (a) f u e l accidents, (b) f l u i d temperature accidents, ( c ) structural i n t e g r i t y f a i l u r e s , and (d) c o n t r o l system mismanagements.

k


- 27 a.

Fuel Accidents

These accidents arise from abnormal q u a n t i t i e s of f u e l i n the coreu This condition could arise from f u e l accumulation i n the core; from sudden l o s s of such accumulation; from deposition i n the external system, with subsequent return of t h i s f u e l t o the core, Examples of possible d i f f i c u l t i e s of this type are:

-

L

AHR

I

(1) Formation of uranium-rich l i q u i d phase by l o c a l overheating; ( 2 ) p r e c i p i t a t i o n of uranium, containing mixed sulphate brought on by build-up of dissolved nickel (most probable i n h o t t e s t region); (3) i n inclusion of uranium i n (or absorption on) the corrosion product on zirconium a l l o y core vessel, o r s i m i l a r association with external system s t a i n l e s s s t e e l s o l i d corrosion product ; (4) deposition of uranium-bearing residue by l o c a l boiling; (5) s e t t l i n g of s l u r r y i n r e a c t o r ; ( 6 ) s e t t l i n g o r caking of s l u r r y i n external system,

(1) Formation of s o l i d 3LiFeThF4 (including some uranium), a consequence of too low a temperature; ( 2 ) p r e c i p i t a t i o n of mixed trifl.uorides, when t h e sum of the concentrations of rare earths, t r i v a l e n t plutonium and uranium (produced by reaction of the f u e l with the nickel-bearing container material) exceeds the s o l u b i l i t y l i m i t ; ; ( 3 ) p r e c i p i t a t i o n o f UOz9 U03, o r U02F2 by the inleakage of oxygen o r water; (4) association of f u e l with deposited f i s s i o n products such as molybdenum and ruthenium ( t h i s seems unlikely, judging from t h e limited information available) ; ( 5 ) association o f f u e l with INOR-8corrosion products ,

m (1) P r e c i p i t a t i o n of uranium from t h e solution o r p a r t i a l freezing of the bismuth with possible g r e a t e r than normal uranium content i n the s o l i d f o r s o l u t i o n f u e l , t h e increased uranium cantent is expected; f o r s l u r r y f u e l i t can be t r u e ; (2) accumulation of m a s s of s l u r r y p a r t i c l e s (presumably by f l o t a t i o n ) i n t h e reactor; ( 3 ) accumulation of m a s s s l u r r y p a r t i c l e s i n external system; (4) possible contribution of above items by enough oxygen inleakage t o prevent wetting of the s l u r r y by bismuth; ( 5 ) association of f u e l with i r o n and chromium deposited on cold areas (mass transfer).

--

In addition t o accidents a r i s i n g from f b e l inhomogeneity, there is a l s o the p o s s i b i l i t y of a f u e l accident a r i s i n g from a concentrat i o n error. Such wrong concentrations could arise from mismanagement during start-up.

i

i


- 28 I n order t o minimize the consequence of f u e l accidents, i t i s . desirable t o operate a t f u e l concentration such t h a t e i t h e r increasi n g o r decreasing t h e concentration reduces r e a c t i v i t y . I n principle, t h i s i s sometimes possible, but i n many cases such operation would too s e r i o u s l y compromise t h e design from o t h e r considerations. I n addition, changes i n f u e l composition during operation seem t o preclude t h e p o s s i b i l i t y . A d i f f i c u l t y analogous t o the f u e l accident i n its e f f e c t i s t h e accidental change of flow rate through t h e core. A r e a c t i v i t y excursion can r e s u l t from t h e changed number of delayed neutrons emitted i n t h e core, A complicating f e a t u r e and a possible contributory f a c t o r t o t h e f u e l accident i s t h e d i f f i c u l t y associated with an accurate accounting of the f u e l inventory i n t h e primary system.

b,

Fluid Temperature Accidents

Such accidents can arise from mismanagement or misoperation of t h e heat t r a n s f e r system. Due t o its l a r g e temperature c o e f f i c i e n t , the AHR is most susceptible t o the cold s l u g accident. t

Thus, the l a r g e negative temperature c o e f f i c i e n t i n the iixR has both advantages and disadvantages. It makes possible t h e introduction of l a r g e amounts of r e a c t i v i t y through cooled f l u i d entering t h e core. On t h e other hand, t h e l a r g e negative r e a c t i v i t y c o e f f i c i e n t can a c t t o quickly compensate l a r g e r e a c t i v i t y i n s e r t i o n s , C.

S t r u c t u r a l I n t e g r i t y Failures

F a i l u r e of a core tank could allow core material t o e n t e r the blanket and thus give r i s e t o a dangerous r i s e i n r e a c t i v i t y . Other examples of accidents due t o s t r u c t u r a l f a i l u r e are the breaking o f f of a graphite s e c t i o n i n t h e core of the non-aqueous systems or the t r a n s f e r of secondary coolant material i n t o the primary f u e l stream. Tfie l a t t e r accident is e s s e n t i a l l y impossible i n t h e AHR due t o t h e pressure d i f f e r e n t i a l of t h e primary and secondary systems, d.

Control System Mismanagement

A t l e a s t t h e MSR and U F R may require control rods. The main reason f o r this i s t h a t i t may be awkward i n these systems t o reduce r e a c t i v i t y . Although the control rods provide a d d i t i o n a l protection i n the event t h a t the negative temperature c o e f f i c i e n t is n o t s u f f i c i e n t t o prevent a dangerous excursion, t h e r e is now t h e associated amount of r e a c t i v i t y a v a i l a b l e i n the system and thus a possible source of d i f f i c u l t y .


- 29 A contributing f a c t o r t o t h e likelihood of nuclear accidents is the small value of b e t a e f f , the e f f e c t i v e delayed neutron fraction. Since the excursions t h a t cause the most s e r i o u s damage have rea c t i v i t i e s far above prompt c r i t i c a l , there i s not too much importance i n the very s e r i o u s cases attached t o the exact value of t h e e f f e c t i v e delayed neutron fraction. This i s not t r u e f o r t h e l e s s s e r i o u s excursion, ??he two-region systems have the lowest e f f e c t i v e delayed neutron fractions. T h i s r e s u l t s from t h e l a r g e r r a t i o s of external t o core volume. A f u r t h e r contributing f a c t o r t o a low beta e f f i s t h e low value of the U-233 delayed neutron portion (.0026). Under some cond i t i o n s beta eff nay be as low as a few hundredths of a percent.

There have been a considerable number of c a l c u l a t i o n s of the consequences of assumed nuclear accidents i n the AH2 and, t o a l e s s e r extent, i n t h e LMFR and MSR. These calculations, done on p a r t i c u l a r designs, have f o r t h e most p a r t , indicated consequences t h a t are n o t too s’erious.

However, i n none of the concepts can t h e negative temperature c o e f f i c i e n t of r e a c t i v i t y be counted on as a completely c e r t a i n prot e c t i v e device against dangerous r e a c t i v i t y (and hence temperature) excursions. Because of t h e many v a r i a t i o n s of the systems under consideration, i t i s d i f f i c u l t t o make any generalizations as t o the r e l a t i v e r a t i n g s of t h e three concepts on t h e i r s u s c e p t i b i l i t y t o nuclear accidents.

3.

Chemical Accidents

As i n other r e a c t o r systems, there are p o s s i b i l i t i e s of various chemical r e a c t i o n s i n the f l u i d f u e l systems. Some of these are: a.

Aqueous Xomogeneous Reactor

(1) A r a d i o l y t i c gas explosion hazard exists, cont r i b u t e d t o by possible f a i l u r e of t h e c a t a l y t i c recombiner.

-

A reaction between t h e core ( 2 ) Zr-D20 reaction tank wall material and water is, i n p r i n c i p l e , possible, Conceivably i t could be i n s t i g a t e d by t h e two-liquid phase separation o r uraniumr i c h s o l i d deposition, and consequent formation of a hot spot and melting through of a hole i n t h e core tank. The design provided where a should make t h i s l e s s l i k e l y than w a s t h e case i n H-2, melt-through occurred (without major reaction), In t h e event i t does occur, the probability of a s e r i o u s metal-water reaction is UnknOldn,


- 30 b.

Molten S a l t Reactor

-

This probably is not (1) Stored energy i n graphite a serious consideration due t o t h e high operating temperature, (2) Molten salt reactions with 1~02-8. T h i s has been indicated t o require quite high temperatures and, hence, unlikely as long as the f u e l i s uniformly d i s t r i b u t e d (dissolved),

( 3 ) Molten salt reaction with water does not l i b e r a t e a l a r g e quantity of energy, although heating the water could c r e a t e pressure. C.

Liquid Metal. Fuel Reactor

(1) Stored energy i n graphite (same as MSR). (2)

Na-H20 reaction, with l i b e r a t i o n of hydrogen and

heat ,

( 3 ) Contact of bismuth with sodium o r container materials should not l i b e r a t e much heat; reaction with air i s slow a t conceivable r e a c t o r temperatures. 4.

Primary System I n t e g r i t y Failure

The most c h a r a c t e r i s t i c s e r i o u s accident i n the f l u i d f u e l systems i s a f a i l u r e i n t h e i n t e g r i t y of t h e primary system, W n t a i n i n g the i n t e g r i t y of the primary system i n t h e f l u i d f u e l systems can be made more d i f f i c u l t by reaction between t h e f l u i d s and t h e i r containers. As an example there can be s e r i o u s e f f e c t s due t o mass t r a n s f e r i n the LMFR, stress corrosion cracking i n t h e AHR. No indication of similar r i s k has been found for molten fluorides i n IN OR-^. I n the AHR, the consequences of any break o r l e a k are far more serious, since the spread of the f i s s i o n products will be more widespread.

5.

Containment

Among the systems under consideration, the AHR has proposed a l a r g e s t e e l sphere o r sealed c e l l f o r containment while the two non-aqueous systems have proposed c l o s e r f i t t i n g containment systems. The sealed c e l l of t h e AHR is designed t o the requirement that i t contain the pressure produced by accidental discharge of the f l u i d s of the r e a c t o r system, The required volume within the containment


- 31 s y s t e m f o r t h e other two cases is not s o g r e a t , since the vapor pressures of t h e f l u i d s are very low.

It i s not deemed possible t o assign a higher probabrility of penetration of a c t i v e material ( t o t h e outside of t h e containment system) i n t h e event of breaching t h e primary system, f o r one r e a c t o r type over the probability f o r the others, Consideration of t h e maximum credible accident f o r each of these systems ultimately depends on the question of how l a r g e a f r a c t i o n of t h e f i s s i o n products (or other toxic materials, e.g. Po210 i n LNl?R o r T, i n PER) i n the system might pessimistically be assumed t o be released outdoors i n an accident. Such considera t i o n s thus are strongly dependent on which v a r i a t i o n of each of the systems i s being considered. Some may have continuous removal of f i s s i o n products while others l e t them accumulate f o r say, 20 years.

In general, however, due t o the presence of high pressure, t h e AHâ‚Ź? would have t o be assigned a l a r g e r "maximum credible" accident

than would t h e non-aqueous system. I n the event its outer containment is breached, t h e f i s s i o n products would then c o n s t i t u t e a public hazard, I f the close f i t t i n g containment. of t h e non-aqueous systems i s breached, t h e r e s t i l l remain s e v e r a l r e l a t i v e l y l e a k t i g h t b a r r i e r s , Furthermore, the c a r r i e r would freeze, imprisoning t h e f i s s i o n products. G,

Chemical Reprocessing

*

The l i q u i d f u e l r e a c t o r systems are distinguished from other perhaps a l l the r e a c t o r types i n having most of t h e f a c i l i t i e s facilities found i n a t y p i c a l radiochemical processing plant. Numerous samples of radioactive f l u i d s must be withdrawn, transported, and analyzed t o follow and c o n t r o l plant operations. Radioa c t i v e f i s s i o n gases must be t r e a t e d , aged, sampled, analyzed, and i n some instances b o t t l e d and shipped off s i t e ,

--

--

There are o t h e r points of s i m i l a r i t y t o radiochemical processing plants, Equipment w i l l have t o be removed from contaminated areas, decontaminated and repaired, o r decontaminated and buried on t h e s i t e , These operations require waste f a c i l i t i e s which are s i m i l a r t o , although much less extensive than t h e f a c i l i t i e s found a t a radiochemical processing plant.

I


- 32 1. Minimum Reprocessing Requirement For two of the three r e a c t o r concepts, i t has been calculated t h a t minimum f u e l c o s t s would be obtained with long cycle times, up t o 30 years, without any reprocessing except f o r removal of f i s s i o n gases. U-235 would be added continuously t o compensate f o r poison build-up and depletion of f i s s i l e inventory,

It i s anticipated t h a t i n a c t u a l p r a c t i c e some f r a c t i o n of the f u e l w i l l become contaminated with impurities caused by such things a s air and water vapor inleakage, l e s s than p e r f e c t purging o f the system o r s e a l i n g of equipment, corrosion products, f i s s i o n products, r a d i a t i o n damage, change i n s l u r r y c h a r a c t e r i s t i c s , e t c , The a c t u a l extent t o which build-up i n these systems may occur w i l l not become known u n t i l a r e a c t o r experiment is operated f o r an extended period of time, I n the meantime, i t seems prudent t o assume t h a t impurities will form t o a minimum extent o f 5%of t h e t o t a l inventory, i.e, the e n t i r e inventory will be contaminated once i n 20 years, I n addition, i t is f e l t t h a t no u t i l i t y would be w i l l i n g t o build and operate a f l u i d f u e l r e a c t o r plant without a means of cleaning up the f u e l should it become contaminated, e i t h e r by gradual degradation or by accidental means, The reprocessing p l a n t s provided i n the reference design meet, o r exceed t h e 5% l i m i t a t i o n described above. 2,

Off Site

VS.

On S i t e Reprocessinq

*

For chemical reprocessing of t h e f u e l , t h e r e i s a choice of doing this e i t h e r o f f s i t e o r on s i t e . O f f s i t e reprocessing involves many problems t h a t are not generally recognized. I n order t o process the high melting f u e l s ( l i q u i d metal o r molten s a l t ) o f f s i t e , the reactor operator must can h i s f u e l i n sealed, dissoluble cans, I n addition, the cans must be small, preferably no more than 2" i n diameter and 4' long, o r constructed i n such a manner t o a i d dissolution; otherwise t h e capacity of the very expensive d i s s o l u t i o n equipment will be s e r i o u s l y reduced, These cans must a l s o be capped, seal-welded, inspected, decontaminated, handled remotely much as s o l i d f u e l elements are handled, and shipped t o a processor i n heavily shielded casks. ' h e n t h e produce i s returned by the processor, some months l a t e r , i t must be returned i n shielded casks. The one operation of f u e l canning f o r shipment o f f s i t e could require about the same investment and involve about t h e same annual expense as processing the f u e l on s i t e .


- 33 It a l s o should be noted that any centralized plant w i l l hold a customer's material f o r a year o r perhaps several years i n order t o accumulate enough f u e l f o r a production run. ?hen a s u f f i c i e n t l y l a r g e inventory i s accumulated, t h e f u e l w i l l be processed as consecutive batches a f t e r flushing out t h e plant. It i s only i n this fashion t h a t the f l u i d f u e l operator can be c e r t a i n of receiving h i s product back uncontaminated by tramp diluents, or i f material i s t o be sold, of receiving proper c r e d i t f o r t h e valuable isotopes. The problems involved i n shipping aqueous f l u i d f u e l s o f f s i t e are complex a l s o . The f u e l must first be evaporated, calcined, and probably vacuum purged t o recover s u b s t a n t i a l l y a l l t h e contained heavy water, The operation must be done i n s p e c i a l shipping cont a i n e r s o r t h e residual oxides must be dissolved i n strong acids o r s l u r r i e d i n water and transferred t o shipping containers, After s e v e r a l months cooling the shipping containers must be sealed, transferred t o large casks and transported t o t h e processing plant. Although shipments of d i l u t e , neutralized f i s s i o n product wastes have been made on a limited experimental s c a l e , t h e transport across country of the d a i l y e f f l u e n t from a 1140 MWT aqueous homogeneous r e a c t o r plant poses questions of hazard and l i a b i l i t y too involved t o be considered f u r t h e r here.

I

i

A comparison of t h e equipment involved will show t h a t t h e operat o r of a f l u i d f u e l s power plant has available on h i s s i t e most of the a u x i l i a r y f a c i l i t i e s required t o process his fuel. Idhen i t is considered t h a t the cost of a loading f a c i l i t y t o can f u e l f o r o f f site shipment appears t o be as expensive as a f a c i l i t y t o process t h e f u e l on s i t e , i t becomes evident t h a t on s i t e processing of f l u i d f u e l s is an economic necessity which i s not l i k e l y t o change with technological progress.

3.

Seprocessing Concepts of Reference Reactors a.

AHR

- Thorex Process

Reprocessing of AHli f u e l will be by t h e conventional thorex process, t h e technical f e a s i b i l i t y of which has been adequately demonstrated by t h e a c t u a l operation of such a plant a t Oak Ridge. Some additional equipment i s required, however, t o provide f o r t h e routine recovery i n high y i e l d and purity-particularly purity--of t h e heavy water associated with t h e fuel. I f the p u r i t y of t h i s recovered heavy water is not up t o specification because of contami n a t i o n with l i g h t water, considerable added expense will be incurred. The fact t h a t t h i s heavy water i s heavily contaminated with t r i t i u m (DTO) contributes added problems of v e n t i l a t i o n , personnel protection, protective clothing, t o o l decontamination, etc.


- 34 b.

EISR

- Fluoride V o l a t i l i t y Process

The most economical process f o r recovery of uranium from molten salt appears t o be t h e f l u o r i d e v o l a t i l i t y process. Although t h i s process is simple i n concept, i t i s more complicated than one might suspect. Corrosion i s severe at the high temperatures involved, materials handling i s complicated, batch f r a c t i o n a t i o n and post treatments of the d i s t i l l a t e UF6 f r a c t i o n s are required t o secure a s a t i s f a c t o r y product. Also, the process involves a secondary high temperature f l u i d i z e d bed reduction of gaseous UF6 t o s o l i d UF4 using hydrogen as reducing agent. TIe f l u o r i n a t i o n s t e p appears t o involve more c e l l decontamination work and personnel access problems than the o t h e r processes. Access i s no problem when processing low i r r a d i a t i o n l e v e l fuels. Access and maintenance i n this application where the c e l l w i l l handle kilogram q u a n t i t i e s of f i s s i o n products each day is a problem which m a y prove t o be of considerable magnitude.

One major disadvantage of t h i s v o l a t i l i t y process i n a nuclear breeding economy is t h a t f o r the one-region reference r e a c t o r thorium i s not recovered but is run t o waste along with t h e i s o t o p i c a l l y separated lithium f u e l d i l u e n t and the accompanying f i s s i o n products. If thorium i s t o be recovered l a t e r , additional thorex f a c i l i t i e s on a s u b s t a n t i a l scale w i l l have t o be provided. c.

DER

- Thorex Process

*

Although a flow sheet has not been developed t h e conventional thorex process probably can be used f o r processing the LMF% s l u r r y fuel. The bismuth does have, however, s i g n i f i c a n t e f f e c t on dissolving problems as w e l l as requiring process d i l u t i o n t o hold solvent degradation within reasonable working limits. Furthermore, t h e bismuth cannot be recovered without the addition of f u r t h e r equipment.

It has been assumed t h a t about 150 kg per day o f B i containing only about 4 kg of thorium, and about 200 grams of f i s s i o n a b l e uranium, and up t o about 0.6 kg of f i s s i o n products w i l l require f a c i l i t i e s equivalent i n c o s t t o those required t o process about 90 kg per day of normally i r r a d i a t e d thorium, The equipment required f o r processing f a c i l i t i e s which recover bismuth reflects t h e same maintenance-access problems involved i n t h e v o l a t i l i t y process except t h e problems probably are much worse due t o aerosol generation and ruthenium v o l a t i l i z a t i o n which a r e t o be expected. This process i s t o t a l l y undeveloped. It contemplates blowing bismuth with oxygen a t elevated temperatures t o slag and

-_-__-

~

*

For divergent view, see project statement on page 95.


- 35 f l o a t off oxides. Aqueous caustic may be used instead o f oxygen but t h i s will not mitigate the radioactive aerosol generation problem A s a minimum figure, t h e recovery of bismuth can add $1million investment ( d i r e c t materials and labor basis) and may very well make t h e process unworkable. In view of the 5193,000 annual saving r e a l i z e d by recovering bismuth, and t h e nebulous saving involved in shrinking t h e s i z e of an already undersize thorex f a c i l i t y , there seems t o be l i t t l e incentive i n pursuing t h i s method.

d.

Throughput Rates

The throughputs l i s t e d f o r each of t h e processing p l a n t s are based on an a p p r a i s a l of t h e l i m i t i n g f a c t o r s involved, assuming no d r a s t i c shutdowns f o r cleanup o r maintenance. Throughput d a t a f o r t h e well-developed thorex plant used on the aqueous homogeneous system is based primarily on O a k Ridge experience with t h i s process. The yearly throughput rate of the reference design is s l i g h t l y g r e a t e r than the t o t a l system inventory and, theref o r e , well above t h e minimum reprocessing requirement mentioned above. Throughput figures f o r t h e fluoride v o l a t i l i t y process are based primarily on t h e experience o f Argonne. The throughput rate of t h e reference design is 20% of the t o t a l system inventory. Throughput figures f o r the LplFR bismuth-thorex (no bismuth recovery) process represent an educated guess as t o t h e e f f e c t of bismuth on dissolving problems and the process d i l u t i o n required t o hold solvent degradation within reasonable working limits. The throughput r a t e of the reference design i s based on 5% of t h e t o t a l sys t e m inventory e The g r e a t e s t uncertainty i n throughput c a p a b i l i t i e s of these plants is not s o much t h e name-plate throughput t h a t can be squeezed through these s m a l l p l a n t s under t h e most favorable circumstances, but i n t h e number of days per year such d i r e c t maintenance p l a n t s can be kept operating, S p i l l s , breaks, o r mishaps can, and occasi o n a l l y do, put p l a n t s of t h i s type out of commission f o r weeks o r even months a t a time i f a bad s p i l l occurs.

4,

Fuel Reconstitution

An area of processing which has received l i t t l e i f any consideration i n any of the t h r e e systems is the r e c o n s t i t u t i o n o f f u e l from t h e thorium and uranium recovered by t h e thorex or other processes used. I f recovered uranium and thorium a r e t o be re-used


- 36 i n these plants, a considerable investment will be required t o prepare t h e appropriate thorium and uranium compounds t o exacting reactor requirements under extremely high r a d i a t i o n conditions which characterize such recycled material. This tail-end processing, which may involve precipitation, calcination, p a r t i c l e s i z e adjustment, recalcination, sample analysis, physical t e s t i n g , and other s t e p s , must be carried out completely remotely behind heavy shielding. Although f u e l reconstitution is common t o all three reactor concepts, i t is noteworthy t h a t the product from t h e MSR fluoride v o l a t i l i t y process only requires control over chemical p u r i t y where a l l other f l u i d concepts require close control over physical chara c t e r i s t i c s such as p a r t i c l e s i z e , p a r t i c l e density, and even p a r t i c l e shape. For the MSR system, probably t h e only operations which might be required here would be the blending of diluent s a l t and makeup thorium fluoride with t h e recovered uranium fluoride product, standardizing t h e mixture chemically, and p e l l e t i z i n g t h e standardized blend. All these operations would have t o be c a r r i e d o u t behind shielding i n a dry box f i l l e d with an inert argon o r nitrogen atmosphere t o avoid moisture absorption and subsequent hydrolysis of t h e fluorides (and p r e c i p i t a t i o n of U02 i n t h e r e a c t o r circuit1

.

A question which is asked i s t h e reason why tail-end process shielding i s necessary when working with freshly separated U-233 which is not p a r t i c u l a r l y radioactive, t h e a c t i v i t y from U-232 only growing i n a f t e r a prolonged t i m e . Experience shows t h a t complete decontamination of t h e processing c e l l and equipment i s not f e a s i b l e a f t e r each use. P a r t i c u l a t e matter becomes d i s t r i b u t e d , equipment gradually becomes more radioactive as time goes on even though t h e Sooner o r later, batch being processed may be q u i t e non-radioactive. however, t h e e n t i r e c e l l and equipment comes t o its equilibrium l e v e l of r a d i a t i o n i n t e n s i t y . It is a t t h i s time t h a t the shielding i s needed.

5.

Investment Bequirements

The c a p i t a l investment requirements and estimated throughput c a p a b i l i t i e s f o r on s i t e processing f a c i l i t i e s f o r the reference plants are shown i n the following table:

t #


- 37 Fhel Cycle Total Investment Summary

-

Concept Process Throughput B i recovered Li recovered Th recovered U recovered

AHR -

-

LMFR -

MSR

Thorex

Fluoride Volatility

300 kg/day Th

60 kg/day

Bi-Thorex SB

150 kg/day B no

no no Yes

Yes yes

Direct Materials & Labor Recovery process Fuel. reconstitution Laboratory Gas F a c i l i t y Stack X e , K r recovery Waste F a c i l i t y

TOTAL M U

4,330,oOO

39 600,000

1,100,oOO

100,ax) 700 S60,000 250,000 1,OOo,~

700,000

560, oo0 250,o

1,000,OOO 560,000

8,5 ~ , ~ 0 6,770,000

Lumped Dist, Items Q

74%

.

6,290,

Approx Total Investment

560,000

oa,

14,790, ~ 0 0

2,600,000 1,100,000 700,000

560,000 250,000

1,~,OOo

560,000

6,770, OOO

5* oao,oo0

5,010,000

1~,780,000

11,780,000

The investment f o r t h e AHR process r e f l e c t s a conventional thorex operation capable of handling t h e capacity shown, For t h e IslIFR and MSR, these are considered minimum f a c i l i t i e s t h a t ' c o u l d be bui1.t and represent what are considered t o be minimum investment costs, The higher investment f o r the MSR recovery process r e s u l t s from t h e f a c t t h a t t h i s p a r t of the chemical processing is more cornplicated, Equipment required f o r MSR f u e l reconstitution i s , on t h e o t h e r hand, much simpler, as described above and therefore l e s s costly, Bismuth could be recovered i n t h e MFR process but t h e value of the bismuth recovered would not pay f o r the added investment cost required,

6.

Annual Operating Requirements

It has been found from experience t h a t the annual operating


- 38 cost of l a r g e radiochemical processing i n s t a l l a t i o n s amounts t o 15% of the t o t a l f u l l book investment. This does not include charges on B i , D20). It does c a p i t a l and the c o s t s of s p e c i a l materials (E, include a l l operating expense items such as supervision, s a l a r i e s , wages, maintenance material, maintenance labor, general supplies, u t i l i t i e s , and appropriate associated d i s t r i b u t i v e items. Operating c o s t s f o r small d i r e c t maintenance p l a n t s which are heavily weighted with hot laboratory f a c i l i t i e s , s k i l l e d personnel, and s u b s t a n t i a l health physics requirements, amount t o much more than 15%. In the absence of a b e t t e r f i g u r e for t h e type operation considered here, i t i s suggested t h a t t h e 15%f a c t o r be used.

I


- 39 -

VI.

!J%X"NCAL FEASIBILI!CY OF BREEDING

I n appraising the technical f e a s i b i l i t y of breeding, the Task Force was requested t o make a survey of available c a l c u l a t i o n s and estimate t h e breeding r a t i o s a t t a i n a b l e i n t h e various f l u i d f u e l systems. I n t h i s estimate consideration w a s given t o any f l u i d f u e l r e a c t o r concept deemed t o be a "reasonable extrapolation" of technology,

Also included a r e estimates of the additional cost t o a t t a i n breeding, developments necessary t o achieve specified amounts of breeding, and a discussion of the s t a t u s of the basic physics data, A.

Breeding P o t e n t i a l

Since the need f o r breeding w i l l i n all probability arise, the question of the breeding p o t e n t i a l of t h e f l u i d f u e l system when operating on the thorium U-233 cycle is of g r e a t importance,

It i s convenient t o d i s t i n g u i s h between the following systems: Converters '%old ownf1 breeders Doublers The converters encompass all systems with e f f e c t i v e conversion r a t i o s less than, say, 0.95, The %old own" breeders are systems with conversion r a t i o s around unity. Their main f e a t u r e i s t h a t they do not have s u f f i c i e n t l y s h o r t doubling times t o make considerations of t h a t quantity of any significance. They do not, however, require any s u b s t a n t i a l amount of f u e l a f t e r they are supplied with t h e i r i n i t i a l inventory (which may be very l a r g e ) . I n principle, such systems permit f u l l u t i l . i z a t i o n of nuclear f u e l ,

s h o r t enough (under Doublers are breeders with 20 years) s o t h a t r e a c t o r capacity based on bred f u e l can expand with load growth, Insofar as breeding is concerned, the major inherent difference between the three concepts under consideration is i n the p a r a s i t i c l o s s e s t o t h e f u e l c a r r i e r and t o the graphite moderator i n the cases where i t is used, These l o s s e s can be minimized by a higher f u e l


- 40 loading. I f t h i s i s done i n LMFR and MSR i t tends t o equalize the breeding gain i n the t h r e e systems (but s t i l l leaving a l a r g e disp a r i t y i n the doubling time unless IJIF'R and MSR go over t o i n t e r n a l l y cooled systems)

.

I n p r i n c i p l e i t i s possible t o obtain a breeder with a one-region system, i f i t is s u f f i c i e n t l y large. It i s possible t h a t f u t u r e c e n t r a l s t a t i o n requirements might favor power outputs corresponding t o such l a r g e s i z e s . Even then, however, i t i s probable t h a t i t would be advantageous t o s a t i s f y t h i s l a r g e r power requirement with a tworegion system. The following "optimum" breeders a r e c h a r a c t e r i s t i c of t h e best breeders t h a t can reasonably be expected t o be obtained i n t h e order of 25 years. The f e a s i b i l i t y of these systems is dependent on t h e successful solution of problems l i s t e d l a t e r . The optimum breeders for each of t h e concepts have the following

characteristics:

AHR

The AHR optimum breeder i s a two-region system with a s o l u t i o n core, s l u r r y blanket with breeding i n t h e blanket only. There would be continuous Xe removal, l i m i t i n g t h e Xe l o s s t o approximately .01 per neutron absorbed i n fissionable material. I n t e r n a l recombination of the r a d i o l y t i c gases woulc be used provided a s u i t a b l e c a t a l y s t can be developed which w i l l not r e s u l t i n too l a r g e a p a r a s i t i c loss. Otherwise, e x t e r n a l recombination w i l l be used. The neutron leakage would be limited t o approximately .O3. There would be frequent processing f o r removal of the non-volatile f i s s i o n products and corrosion products. It is expected t h a t the l o s s e s t o the non-volatile f i s s i o n products and corrosion products would be i n the range .O3 t o .06. The frequency of processing would be determined by a compromise between c o s t against the doubling time. A breeding r a t i o of approximately 1.07 t o 1.10 i s expected. The t o t a l f u e l inventory i n both t h e r e a c t o r and processing f o r 1140 thermal MW(333 MltE gross) would be approximately 600 kg. The corresponding doubling time would be 15 t o 20 years.

The MSR optimum breeder is a,two-region system with a graphite s t r u c t u r e i n the core but not i n the blanket. It has a heavily loaded core. Breeding is accomplished i n both t h e core and blanket. There would be continuous Xe removal. The


- 41 There neutron leakage would be limited t o approximately 0.03. would be frequent processing f o r removal of the non-volatile f i s s i o n products

.

A breeding r a t i o of approximately 1.05 i s expected at a s p e c i f i c power of approldmately 1 thermal MW per kilogram of fuel. Preliminary s t u d i e s have been on a 100 thermal MW system. It is assumed t h a t the s a l t i s retained a f t e r processing s o t h a t the Li6 i s s u b s t a n t i a l l y burned out.

The LMFR optimum breeder is a two-region system. It has a. graphite s t r u c t u r e i n both core and blanket. It uses a solut i o n f o r the core and s l u r r y i n the blanket. There is breeding i n the blanket only. There is continuous removal of Xe. !i'Iie neutron leakage would be limited t o 0.03. There would be frequent processing f o r removal. of non-volatile f i s s i o n products. A breeding r a t i o of approldmately 1.05 could probably be obtained a t a s p e c i f i c power of approximately 1 thermal M1J per kilogram of f i e l . Preliminary s t u d i e s have been on a 800 thermal MW system.

Depending on the successful s o l u t i o n of c e r t a i n technical feasi b i l i t y questions, the AHR has a chance o f having a reasonable doubling time. On t h e o t h e r hand, the present concepts f o r MSR and LMFR, although they can lead t o "hold own" breeders, have no possib i l i t y of being good doublers. The only p o s s i b i l i t y f o r MSR and LiWR t o be good doublers would be by the i n t e r n a l cooling concept. W s , however, introduces a new development program with many d i f f i c u l t and complicating features. If solved successfully the p o t e n t i a l i n terms of doubling time would then be roughly equivalent t o t h a t of the presently envisaged AHR. I n evaluating t h e three concepts as "hold owntt breeders, t h e difference between the systems is n o t large. The ALQ has a breeding r a t i o advantage of about -04 over t h e non-aqueous systems. I n all probability the breeding p o t e n t i a l of the two non-aqueous systems are very close. The higher p a r a s i t i c capture of the molten salt c a r r i e r r e l a t i v e t o bismuth i s compensated by t h e smaller f l u i d volumes (both i n t h e core and i n t h e e x t e r n a l system) needed i n the molten salt case. A t this time i t i s not possible t o make a r e l a t i v e r a t i n g on t h e breeding potential. of the two non-aqueous systems. I n both of these systems t h e breeding r a t i o , when around unity, can be improved s l i g h t l y at t h e expense o f fuel inventory. The molten salt r e a c t o r has a higher s o l u b i l i t y l i m i t and i s consequently more


- 42 f l e x i b l e i n this respect.

I n any event, when operating on a "hold

own" basis, t h e r e i s probably l i t t l e advantage t o push t h e breeding

gain up s l i g h t l y at t h e required c o s t i n f u e l inventory. B e

Additional Cost t o Attain Breeding

Present design s t u d i e s lead t o t h e conclusion t h a t converter systems are now the most economival. It i s of i n t e r e s t then t o estimate t h e additional c o s t t o make these systems breeders. In the case of the AHR with 15 year doubling time, t h e main a d d i t i o n a l cost above t h a t o f the design operating i n t h e most economic fashion is t h e cost involved i n processing more frequently. A rough estimate of t h e t o t a l a d d i t i o n a l c o s t (including fixed charges, opera t i o n and maintenance, and f u e l and processing c o s t s f o r AHR as a i 5 year doubler compared t o t h e AHR minimum c o s t r e a c t o r ) i s approxFor the Wti'R and MSR there a r e two separate imately 1 mill/kwh. cases t o consider. First, t h e r e are t h e modifications of these systems t o make them "hold own" breeders. The cost of t h i s above t h a t of t h e LMFR and MSR reference designs i s of the order of 1 m i l l / kwh. The second case i s the far g r e a t e r extrapolation t o i n t e r n a l l y cooled systems s o t h a t they operate as doublers. It i s e s s e n t i a l l y impossible t o make any estimate of the a d d i t i o n a l c o s t involved since a t t h i s stage comparatively l i t t l e i s known of t h e technical feasib i l i t y of these systems. Among the problems t o be solved are the development of adequately impermeable graphite (or possibly beryllium) tubing and extremely r e l i a b l e graphite-to-metal seals.

same

C.

b

Necessary Developments

The m a i n questions associated with t h e technical f e a s i b i l i t y of breeding (more s p e c i f i c a l l y , of obtaining a reasonable doubling time) i n the AHR are:

le 2,

3. 4. 5.

-

removal of Xe s l u r r y core; solutiol: core removal of non-volatile f i s s i o n products and corrosion products core tank r e l i a b i l i t y economic reprocessing s t a b i l i t y of s o l u t i o n and slurries under concentrations needed f o r breeding.

The main questions associated with t h e technical f e a s i b i l i t y of breeding i n t h e non-aqueous system on a "hold own" breeder basis are: 1. removal of X e removal o f non-volatile f i s s i o n products (and corrosion products i f non-negligible

2.

c


- 43 f e a s i b i l i t y of adequately blanketing systems on t o p and bottom k,, f e a s i b i l i t y of graphite o r possibly beryllium as the core tank material 5 , core graphite r e l i a b i l i t y 6,. economic reprocessing s t a b i l i t y of s o l u t i o n s and s l u r r i e s under concentrations needed f o r breeding.

3.

Tlne main f e a s i b i l i t y question with respect t o t h e non-aqueous systemj as good doublers is t h a t of developing an i n t e r n a l l y cooled system, D.

-

S t a t u s of Basic Physics Data

Due t o u n c e r t a i n t i e s i n some of the basic data, primarily eta (number of neutrons emitted per absorption) of U-233, precise predict i o n o f the breeding p o t e n t i a l i s uncertain. The s t a t u s of t h e more important c r o s s s e c t i o n data leading t o u n c e r t a i n t i e s i s as follows: The current best estimate f o r the value of eta of U-233 a t thermal There a r e a number of measurements centered around energy is 2.28. this v;?lue. There are, i n addition, some B r i t i s h measurements yield"he conclusions i n t h i s i n g a subst.antially lower value of 2.18. evaluation are based on a value o f 2.28 i n the b e l i e f that t h i s is t h e most probable value. It, of course, must be borne i n mind t h a t the estimates of breeding r a t i o and, more importantly, doubling time are s e n s i t i v e t o u n c e r t a i n t i e s i n t h i s quantity. Not only t h e thermal value of eta of U-233 is important but a l s o its energy v a r i a t i o n i n t h e epithermal region. This i s p a r t i c u l a r l y true w i t h reference t o those v a r i a t i o n s of MSR and LMFR which have s u b s t m t i a l amounts of epithermal fissions.

+

It seems reasonable t o assume an uncertainty of the order of 0.05 i n t h e value of eta both i n the t h e m a l and epithermal range,

Kotwithstanding the f a c t t h a t this uncertainty does not overlap the low B r i t i s h thermal value. A corresponding uncertainty i n t h e breeding r a t i o o f 2 0.05 r e s u l t s . Even with' the uncertainty i n the e t a value, t h e conclusion t h a t all the systems can be made t o be "hold own" breeders i s probably safe. This follows from t h e f a c t t h a t breeding gains based on the most probable values are estimated t o be somewhat over unity. Furthermore, within the meaning o f "hold ownf1 breeder, a value s l i g h t l y less than unity is e s s e n t i a l l y as good.


-44-

There i s considerable uncertainty i n t h e long-term f i s s i o n product cross sections, This is of g r e a t importance i n t h e long burnup, low conversion systems, However, i n the systems optimized as breeders, the uncertainty i s not too important; t h e main e f f e c t is then i n determining the necessary frequency of reprocessing.

There is also considerable uncertainty i n the Pa-233 absorption cross section, p a r t i c u l a r l y i n its dependence on p i l e spectrum. However, it is not too important i n any of t h e optimum breeder systems since i n none of them a r e the Pa-233 l o s s e s large.

VII, POWER COSTS A.

Preparation of the Comparative Power Cost Tabulations 1. Purpose

“he cost estimates were prepared t o provide a b a s i s f o r comparing the r e l a t i v e power cost p o t e n t i a l i t i e s of the molten salt, l i q u i d metal a e l , and aqueous homogeneous r e a c t o r concepts f o r l a r g e c e n t r a l power p l a n t s , assuming them t o be developed i n accordance with t h e reference designs submitted by t h e projects,

The AEC assignment t o t h e Task Force did not request comparison with estimates f o r any other r e a c t o r s , o r f o r conventional plants. Ehphasis was placed on estimating the r e l a t i v e c o s t s , t o r e f l e c t as nearly as possible the e s s e n t i a l differences between the concepts; there w a s less emphasis on determining absolute cost l e v e l s , Hence, the plant and power cost estimates should not be considered as being comparable on an investment o r mills/kwh b a s i s with cost estimates for other p l a n t s o r r e a c t o r concepts, and t h e Task Force study does not attempt t o evaluate the r e l a t i v e merits of these and other r e a c t o r types o r conventional power plants. 2.

T i m e Perspective

The tabulations a r e not intended t o represent the c o s t s of building f l u i d f u e l power r e a c t o r s now; n e i t h e r do they predict t h e cost of building a first plant at some f u t u r e date, It is assumed t h a t when the p l a n t s a r e b u i l t , concepts and designs w i l l have been f u l l y developed; t h a t experience will have been obtained i n the operation of a prototype and a t l e a s t one large-scale p l a n t ; and t h a t components will be obtainable from commercial sources. On the other hand, the estimates a r e not considered as r e f l e c t ing the long-term p o t e n t i a l i t i e s of the concepts. They do not take

k


- 45 i n t o account economies t h a t might be expected from larger plant c a p a c i t i e s , l a r g e r on s i t e chemical process operations, and more economical chemical processes.

3.

Procedures Followed i n Making t h e Estimates

The comparative estimates were based on a review of the concepts f o r which conceptual designs and cost estimates were submitted by the respective projects. Comparative plant investment and power cost estimates were prepared by the Task Force on the basis of these data, taking i n t o account judgments of Task Force members i n r e l a t i o n t o t h e i r several specialized technical f i e l d s .

’

Slnce t h e designs and cost estimates submitted were i n d i f f e r e n t s t a g e s of development and had been prepared by d i f f e r e n t groups employing various design assumptions and sources of data, i t w a s considered necessary t o a d j u s t them t o a consistent cost estimating basis. The Task Force considered t h a t some f e a t u r e s of the p l a n t s were common t o a l l three concepts, i , e . that the c o s t s of some f e a t u r e s did not depend upon the c h a r a c t e r i s t i c s of t h e p a r t i c u l a r reactors; and t h a t f o r other f e a t u r e s i n the present s t a g e of development there was no way of evaluating t h e differences. For plant items i n these categories, i d e n t i c a l c o s t s were assumed f o r a l l three concepts, For most of the plant items, e s s e n t i a l c o s t differences were i d e n t i f i a b l e . These items included the reactor buildings, r e a c t o r , heat t r a n s f e r loops, steam generators, power equipment, and many of the a u x i l i a r i e s , Calculations were made t o check the reasonableness of the design assumptions and cost estimates t h a t had been presented. Data w e r e adjusted t o a common basis, i n order t o minimize t h e e f f e c t of non-essential v a r i a t i o n s , The estimates were checked against d e t a i l e d estimates for conventional plants, against the conceptual plant description, and i n r e l a t i o n t o each other i n an e f f o r t t o correct omissions and assure reasonable comparability, The e f f e c t s on construction c o s t s of unusual construction materials such as INOR-8,Zircaloy, and graphite, and unusual requirements f o r leaktightness, r e l i a b i l i t y , and a c c e s s i b i l i t y f o r remote maintenance were considered, E f f o r t s were made t o reconcile the widely-varying judgments of the Task Force members regarding c o s t estimates i n f i e l d s where experience on f l u i d f u e l r e a c t o r s is limited and sometimes t o t a l l y lacking. (Examples a r e remote maintenance f a c i l i t i e s and spare parts.

ts w a s somewhat d i f f e r e n t from t h e above, w i ven t o the d a t a submitted by t h e projects. The i n i t i a l analysis submitted by the p r o j e c t s i n regard t o chemical processing were not considered by t h e Task Force

e


- 46 t o be s u f f i c i e n t l y developed f o r use i n the o v e r a l l cost comparisons. The Task Force adopted a general s e t of cost estimates f o r chemical plant investments, fixed charges, and operating c o s t s covering the minimum on s i t e processing f a c i l i t i e s t h a t were considered f e a s i b l e f o r each of the three reactors. The cost p i c t u r e thus evolved w a s considered by two of the project leaders (MSR and Lt4F.R) t o be incons i s t e n t with the assumptions on which t h e i r design concepts were based, These project leaders have prepared new f u e l cycle analyses based on f u e l cycles and processes which they consider t o show promise f o r future cost reduction, These a l t e r n a t i v e analyses are included i n the respective project statements, pages 76 and 95. Operating cost estimates f o r the power p l a n t s were based upon estimates submitted by the LMFR project f o r the operating personnel required f o r the various plant functions. Pay r a t e s assigned were analogous t o those i n coal-fired plants, The annual maintenance c o s t s were assumed t o be 3% of the plant investment. This percentage is within t h e range of experience i n maintaining various radiochemical f a c i l i t i e s . The annual c o s t s of operating and maintaining the f u e l processing f a c i l i t i e s were taken as 15%of t h e investment i n these f a c i l i t i e s .

4.

Ground h l e s

In determining t h e annual fixed charges corresponding t o the plant and inventory investment cost estimates, t h e following ground rules were applied: fixed Charge Rate for Depreciable Plant and Inventories

14%

This w a s arrived at as follows: Bond Interest (3.5% on 50% of capital Preferred Stock Mv. (5%on 15%of capital) Common Stock (10% on 35%of capital) Overall Return Federal Income Tax (52/48 of return on common and preferred stock) Other Taxes (real estate, e t C . ) Insurance (other than 3rd party l i a b i l i t y ) Sub-total Depreciation (sinking fund Q 6% 30 yrs.) Total Fixed Charges ,

1.75

%g 4.60

70 2.00 0.10 12

1.30 14.004:

'Jhorium and fuel-carrier materials were considered as depreciable inventory.


- 47 F'ue

Use Charge for F i s s i l e Materials (applied to average inventories of U-235 and U-233 i n the reactor during a complete fuel cycle 1

Plant Operating Factor (the equivalent of 7000 hours/year at f u l l load t h i s is based on the very low incremental operating cost, consistent with the fuel cost estimates and adequate avaiLability)

4%

80%

-

B.

gonclusions

A review o f the o v e r a l l summary of power c o s t s , Table V I I - 1 which follows, shows s i g n i f i c a n t differences i n estimated c o s t among tilt? three concepts i n these major categories: Reactor and Steam Generating Plant Turbogenerator Plant Chemical Processing and Waste Disposal Plant Fuel Burnup, Use, and Inventory Charges

As can be seen from t h e d e t a i l e d t a b l e s supporting the summary, these differences are accounted f o r primarily by the p a r t i c u l a r design features, and f u e l cycle assumptions discussed below : 1. The high pressures and low temperatures i n the primary system of aqueous homogeneous r e a c t o r s are r e f l e c t e d i n higher c o s t (of the steam generators; and, because of t h e r e s u l t a n t lower conversion efficiency, i n higher c o s t of the turbogenerator and p o w e r system components.

2. The indicated cost disadvantage of t h e molten s a l t system (compared t o LMF'R) r e s u l t s f r o m the c h a r a c t e r i s t i c s of t h e particular heat transfer 'system chosen f o r t h e reference design: t h e combination of a secondary f l u i d (barren s a l t ) with a high-melting point and a Lieffler b o i l e r system t o permit a steam-cooled seconda r y heat exchanger, Not only does t h e L i e f f l e r system increase t h e investment i n equipment and buildings; it a l s o r e q u i r e s development of a s t e a n pump.

Other possible systems t h a t have been proposed, such as one employing a low-melting-point salt i n t h e secondary loop, are being investigated. I f such a system proves f e a s i b l e , there should be no g r e a t difference i n t h e investment requirements between K R and LMF'R.

3* The aqueous homogeneous r e a c t o r reference design i s f o r a breeder r e a c t o r ; the other two are f o r converters.

This difference


I

- 48 is r e f l e c t e d i n ( a ) more frequent f u e l processing f o r t h e homogeneous r e a c t o r , which r e q u i r e s a l a r g e r and hence more c o s t l y chemical plant; and (b) a n e t c r e d i t f o r f i s s i l e material produced, i n s t e a d of a n e t cost f o r f u e l burned. Further, t h e homogeneous r e a c t o r shows a higher f u e l c a r r i e r investment because of the l a r g e volume of heavy water required. It i s i n t e r e s t i n g t o note t h a t these f u e l cost differences between t h e homogeneous breeder reactor and t h e non-breeder r e a c t o r s approximately o f f s e t each other, with a s l i g h t c o s t advantage i n favor of the breeder ( f o r f u e l processi n g investment and f u e l cycle c o s t s combined: 2.74 mills/kwh VS. 2.84 mills/kwh). This r e s u l t obviously does not support the conclusion which has been sometimes expressed, t h a t breeding necessari l y i n c u r s a f u e l c o s t penalty. However, these r e s u l t s are not considered t o support d e f i n i t i v e conclusions, i n view of the present s t a t e of development of t h e concepts, designs, and f u e l processes. Overall, t h e differences i n power cost estimates among the three concepts are s o small i n r e l a t i o n t o t h e p r o b a b i l i t y accuracy

of t h e estimates t h a t they are not considered a determining factor. C.

Presentation of Cost Information

The o v e r a l l summary of power c o s t s , Table V I I - 1 , which follows, is based upon more d e t a i l e d information presected and discussed as noted below: Descriptive Data Power P l a n t Investment Chemical aeprocessing Plant Investment SUrmnary

of Fuel Costs

Chemical Reprocessing Operation and Maintenance Power Plant Operation and Maintenance

- Sections Project Reports, I X , X, and XI - Table VII-2 and notes - Section V-G

-

Table VII-3 and notes

- Section V-G - Notes on Table VII-1 and Sections V-C and V-D

E


a

I

i

I a

N

g

rq

n

o

N

"3.0:

N

r-

r l r -

.';?rl

u \ W

'r 0l .mI

rl

"3.

r l r l

N r ! i

m

$ 5

m

'10.

%

N

N, N

2

tc

p: rl

o

N

m

a/ x


T a b l e VII-2

COF2.IZRISOM 0% IWICAT% E V E L OF WWR P W N T INVES"iEI?T FOR B E ~ L E I J C ECONCEPTS OF bCTUR3 STATIONS WITH FLUID FUEL R&XTOiG Note:

figures below a r e not t o be considered as r e l i a b l e cost estimates.

The figures represent adjusb.ents

m i y r e s presented by project representatives i n those categories where various disinterested t a s k force The figures represent investment levels members have experience and judge such adjustments as necess-. judged t o be somewhat i n line 55th costs t h a t might be obtainable i n the d i s t a n t future ( i n terms of 1959 d o l l a r s ) and a f t e r all necessary research and development has been completed, all technical problems r e m k e d ( i f pobsible) and after experience with dezicn and operation of a few generations of full-scale power stations. It i s judged impossible t o provide any of t'nese pomr p l m t concepts today for figures an7ywhez-e app-OaChin~.theSet o t a l Is?els,

Em

-m 310.

WJD & UXG RIGHTS

3%,

STRUCTLi-R c t IIGROVEXXE S i t e Improvenents $2 S i t e F a c i l i t i e s .3 Station Buildings .31 &actor 8C Steam Gen Bldg 032 Turbine & Lux Bldg *33 S e m L! ; h i n t Bldg 034 C h a C \"Ja~te Bldgs 035 Other Station Bldgs

2.co

.1

312,

-

___

-pJ-

--v

1,60 27.50 -

1.60 -

24&

21090

21.00

18.00

1.30

1.90

1.03

1000

2,40 1.00

..._

_.,_~, ,~

1 .nr ' Y

I. I \

0.50

6 .lo

66.70 -

,>

0.50

0.50

5.00

11.20 21.70 s030

17 70

2C 70

l.LO

3000

11.60

1&30

none none

3.60 10.60 2030 none 3.70

13*8O 4.10 3.10 3.70 0.30 0.90

0.30 0,30

3.70 4.00

Sob

4.00

1-70

16.00

18.00 2.10

4070

1.50

0,t.O

2.80 0 !rO

0,2c

L50

0.20

6.70

4080

9.50

0.70

4.50

4050

0.50 4.50

4.60

5

Solo

~~

DI2ECT C 0% E U C T I O X

Overhe& Cons'c (%) f.llow f o r m i s s Conting (2%;) Ckp Train k Prep f o r Reg Gpr

TOTAL COILSIRUCTION COST Spare P a r t s Inventory IIeat Trans LFLuids & I n e r t Gases

I

en

0 I

n

37050

37-50

o,ao 38 30

0060

0.60

0.60

5.50

4.50

0.20 6.00

0.20

10.70 0020 10.70

6.00

0.20 0070 1.70 0.90 0.20

0.20

1.50 2030 1.10 0020

1.70 0.90

0.20

0.60 -

316.

~

,.

0.50

ACCZSOEY EI.,XCmI,CIL CZJJIPMZXT Foundations & Structures Pwr & Conversion Equip Conduits,Conducirs %. Insuis Syitch, Cont & Protec Equip Station Grounding SystCm

ILO

IC'C

.%

315,

-I\

10,OO

10.00

10.00

06.00 -

13.b0

1.50

30 3.00 13040 2.50 none 3.70 0.30 0020 0

26.50 -

2000 2.60

2,oo 2.60

\, >,,\ .--

'lurbineGen & E x c i t e r Equip !turbine ?l A m ( i n c l Crane) Condenser & Auxiliaries Turbine P1 piping & Insul Circuh t i n g Water System

--.

19.50

,1

---.

333-R

2.60 22,?0

r:tucm

?: STELA GEN. PLUYT %actor Lquipment .2 %at Transfer Equip .21a & i m y Sgs E l .ci Stg Q u i p .21b ?rimy SJSb o p Squip .22a I n t Sys E l L Stg Lquip .22b I n t Sys Loop Equip .3 Zeactor Plant l i d l i a r i e s .3l Xadioactive Cont Zquip Let down 8C Recomb Equip .32 .33 Reactor Off-gas Equip .& I n e r t G a s squip 035 h r g e Cquipment -36 sys iIeat,Co.ol,& vent quip S t e m Generator Equipment 5. Steam Generator Plant dux. .51 Boiler Feed biater ?quip --& ?e& t!a-kr ~ r e c t i n g& p i p 053 Serv Boiler & Fuel O i l Equip .6 b a c t o r & stm Cen 11 ~ p sys e Reactor & Stn Gen F 1 Insul .7 .O 'controls Imtr m u i p Iiot C e l l L L3emo% Xaint q u i p .9

AIB

VSR

333%

3 3 m

--

152.00

162,bO

180,i'O

22.80 45.60

48 70

SL.20

47.10 6 . "C

5'2.30 6.00

44.10 6,oo 270,;O

15,OO .60

27.10

286,60

320.16 15.00

1S.CO

&.to

.lo

3:-

B l i s t i n g of principal i t e m s considered t o be included under each i t c n i n t h i s 'trible is contained i n Appendix B.

++x

Investment f o r plant f a c i l i t i e s 'to handle waste and reprocess f u e l ( i o e . Stack, Laboratory, Gas, f u e l recovery, f u e l reconstltution, Xe and Yi' rccovery and waste f a c i l i t i e s ) i s not included i n this Table but i s shown i n section V 4 Chemical Processing, page 31.

-


Certain elements of plant investment cost need not vary t o any significant degree f o r purpose of r e l a t i v e evaluation of these reactor concepts, although such elements may vary considerably f o r specific l,ocations and, for s p e c i f i c i n s t a l l a t i o n s f o r d i f f e r e n t u t i l i t i e s . For purpose of r e l a t i v e comparison of the plant investments, the concepts were a l l assumed t o be suitable f o r some i d e n t i c a l s i t e and t h e following elements of investment cost were considered t o be t h e same f o r a l l of the concepts: Land and Land Rights S i t e Improvements

S i t e (General Use) F a c i l i t i e s Service Ee Yaintenance Bldgs. Ffisc. Station Bldgs.

Turbine Plant Auxiliaries Turbine Plant Piping Systems ilccess. 4 k e ::quip. Pdns. 8c Structures Station Grounding System PHsc. l’ower € l a n t auipment

i

Other elements of plant investxent cost concerning which l i t t l e or no d e t a i l variances could be ascertained because of lack of specific ideas having been developed pertaining thereto, but f o r which provisions rather common t o each of the concepts must be made, were considered t o be the same f o r all of the concepts. These are as follows: Reactor Off-Gas Quip, I n e r t Gas Equipment

Control & Instru. Equip, Mot Cell Zr iiemote Kaint, Equip.

Elements of plant investment cost where distinguishing differences are judged t o exist and b r i e f explanations o f these differences are as follows:

a.

Figures presented by the project representatives h c l e a r Steam Generator Building: are considered t o be of proper order f o r the NJS, and iER concepts, Building f o r t h e ,E€? and Ul~l?a r e considered- essentially the same, except f o r additional requirements of the i4SE concept essontial f o r extra loop system i n the heat transfer f a c i l i t i e s between the reactor and the superheated-reheated steam system.

b.

Figures used f o r the IiZR and :SR concepts a r e Turbine and Auxiliaries Building: considered t o be e s s e n t i a l l y equivalent and similar t o cost f o r this eknent taken from a f o s s i l f u e l plant having l i k e turbine and a u x i l i a r i e s characteristics. iLdditional space required due t o much greater condensing capacity and related much larger size of lotr pressure, saturated steam, non-reheat, tandem compound, double flow turbine-generator equipment required f o r the AHR accounts f o r the difference reflected i n the figures f o r t h i s eknent,

C.

Figures f o r a l l systeris were developed fron material requirenents Xeactor Jquipment: and weights f o r the respective specifications presented, considering graphite internals f o r both the U;FTr and KSR concepts and core vessel and thermal shield internals f o r t h e IGITL concept. Also considering control rods and operating mechanisms f o r the i2dX and i%i; concepts and supports and miscellaneous hardware f o r all of the vessels,

d.

Heat Transfer Equipment and Reactor Plant Auxiliaries: AFi.prespresented by project representatives were used exceDt i n those cases where disinterested task force members had experience and basis f o r judging t h a t adjustments were necessary, iilso, fi,ures judged t o be reasonably representative were inserted f o r elements of cost there omissions were evident. I n order t o insure reasonably complete coverage and as much accuracy as possible with respect t o the necessary principal equipment requirements f o r the heat transfer and i e a c t o r plant auxiliaries, these features of the plant equipment requirements were broken down i n t o approximately 100 item f o r the investment cost evaluation.

e,

S t e a m Generator Lquipment and Steam Generator Plant Auxiliaries: Mgures presented bv wro.iect renresentatives were used f o r the steam wenerator eauiumnt. However. ”

f i f i w e s f o r a h systems were developed f o r feedwate; heaters, doiier feed pumps,deaerator and auxiliaries, condensate and heater drain pumps, condensate storage tanks, reducing and desuperheating s t a t i o n (except f o r the iiKR concept, r e q u i r i w none), and fecd1:ater treating equipment. The figures developed f o r these elezents r e f l e c t t h e s p e c i f i c differences s e t b> the reqtrirements of the steam generator characteristics of the respective conceptse

0

I.

Q*

€ % w epresented by the project representatPJes Xeactor Plant F5ping and Insulation: f o r thc CEil concept >ias ciieckod and used, and approximate nei,hts of piping were ascertained f o r a l l of the concepts and figures were dcvelopcd ;or the iX3R and X I 2 concepts by factoring rrith r e l a t i o n t o the U i 1 1 figure. LFizures f o r insulation vere judzed t o be equivalent t o 10,:of the piping figures used, i35;’iC;ures f o r the IiLT and ?SE Turbine P l a n t Poundations and Turbo Generator Lquipment: concepts axe considered t o be essentially equivalent and simi.1a.r t o c o s t f o r t:hLo elcncnt taken fro:% a fossil 1\01 p h n t having like turbine characteristics. Higher fiii;;urcs f o r the ikL1 concept a r e due t o much larger size condensin,r capacity and related much lwk,cr s i z e uf the low pressure, saturated stem, non-reheat, tandem compoun6, double flow turbine-generator equipment and the much Crcater s i z e condensi.ng capilcity required f o r t h e A I Z concept. Turbo generator equipricnt costs f o r each of t h e concepts rrere taken from current catalog prices t o which I r a added an amount of 10;J f o r i n s t a l k tion,

I _

5.

7igwes f o r a l l systems r e r e develoyed.based upon relatiCondcnscr and es: difference i n r e q u i r a s f o r condensing capacitj set b> requireiieiits of t l e charact e r i s t i c s of the respective concepts.

i.

PXgures used f o r the W: ‘and 753 concept are considered a fossil Cue1 p l a n t having l i k e condensing water requirements. ilcqnirci.icnts f o r additional condensine water due t o characteristics of the NIR concept accounts f o r difference reflected i n the fi[>ures f o r this element.

GirculatinL Water *sten:

%o be essential17 eqcLvalent and similar t o cost f o r t h i s element taken from

jo

1 0 e r and Conversion klquipment, Conduits, Conductors and Insulators, Switching, Control and Protective 2quipent: Figures f o r all systems were developTfrom an estimate, based on specifications presented by project representatives, of horsepower f o r plant auxiliary equipment and factoring v l t h r e l a t i o n t o requirements of a fossil f u e l plant of the same comparable output capacity,

k.

Indirect construction costs were judged t o be 1.5’7;of Direct Distributive upenses: Construction Cost and Overhead Construction Costs were judged t o be 3C$ of iKrect Construction Cost based upon experience of Architect 6ngineer representatives building f o s s i l f u e l installations. An allowance f o r Ommissions and Contingencies of 25);; of Direct Construction Cost was judded t o be a minimum prudent allowance considering the s t a t u s of the art a t this time,

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- 53 -

VI11

.

RESEARCH AND DEVELOPMENT PR(jG2AMS

The accompanying Table VIII-1 summarizes time schedules and funds required f o r development programs a s presented by the Project Directors i n Sections I X , X, and XI. These estimates indicate t h e i r appraisal of the e f f o r t required t o develop the s t a t e of the a r t f o r each concept t o a l e v e l suitable f o r designing and constructing commercial power plants. The Task Force has not considered these programs i n d e t a i l nor put them on a comparable basis. To carry through t o a practical design of a commercial power plant, each program requires research and development, design, construction, and operation of a reactor experiment and of a prototype reactor. The cost of the e n t i r e development program f o r any one concept would be something above one hundred million dollars. Ten years would be required. There is no indication that any one would be e;ubstantiallg less costly than the others.

All of these estimates are, of course, based on the assumption that, the problems on which technical f e a s i b i l i t y depends will be successfully resolved. The programs should be reviewed periodically t o appraise t h e i r progress. The Task Force believes t h a t i n these programs major emphasis should be placed on extensive design and cost studies and evaluation of i n d i r e c t maintenance schemes suitable f o r large-size radioactive equipment. Subsequent research and development, reactor experiments, and prototype operation should be guided by the results of these studies

.

In the event t h a t strong emphasis i n the Commission's development program i s placed on breeder systems: the AHR program would be essentially unchanged; t h e E4SR and LMFH would place greater emphasis on the development of two-region graphite moderated reactors and on reprocessing system.


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- 55 IX.

MOLTEN SALT REACTOR

Table of Contents A.

B.

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

Page No.

The Objectives of the Molten S a l t Reactor Program 1. Low-Cost Power 2. Breeding The Approach e 1. Summary of Concept e a. Basic Reactor Concept b. L o w Cost Power Reactors e C. Breeding Reactors e d. The Special Merits of the Molten S a l t Concept. e. Limitations of the Molten S a l t Concept 2. Development o f Low-Cost Power P o t e n t i a l i t i e s a. The Reference Design for the Economic Plant (1) Description (a) Proposed Fuel System and Cycle e Fuels h e 1 Problems

.. . .

.. ..

.. .. .

... .. ..

.. .. . . . .

.. .. .

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

. ..

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

Materials (b) The Reactor e (c) Primary Heat Exchange System Heat Exchangers e e Fuel Circulating Pumps piping primary Auxiliary Systems (d) Secondary Systems and Steam Gen. (e) Control, Instru., & O p e r e Charac. Facil. & Problems (f) Special h i n t . Inherent Problems e Methods Employed to Solve Maint. Problems 0

.

0

. .

-

. .. . .. . Hazards . Maintenance Equipment . . Major Equipment Requirements . Reactor Operating Characteristics . . Potentially Achievable Cost Reduc. . The Development Program . . . Development the Breeding Potentiality . .. Two-Region Homogeneous Breeder . . One-Begion Graphite Moderated Breeder . . Two-Region Graphite Moderated Breeder e

0

Spare

(g)

(2)

3.

b. The a.

b. C.

0

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56 56 56 57 57 57 57 57

58 59 59 59 59 60 60 61 62 63 64

65

65 66 66 68

69 70 70

72 73 74 74 74 75 78 82 82

83

84


- 56 -

IX - MOLTEN SALT REACTCR

1

A.

L

The Objectives of the Molten S a l t Reactor Program

1. Low-Cost Power Assuming success of an orderly development program it should be possible t o s t a r t construction of a 333,000 KWE e l e c t r i c a l gross p l a n t of the type discussed herein with these power costs: Klls/Kwh Gross Power Plant Investment ($31l/KWE) Chemical Processing Plant Investment ($35/KbE) Fuel Inventory, Use Charges and Burnup Chemical Processing:Cperation & Maintenance Power Plant Operation & Nairitenance

6.24 7x4'. 1*3? 76% 1.5'7

-

Total Power Cost

- Gross

10.7

Total Power Cost

- Net

11.1

FJith reasonable developments and design refinements, t h e c a p i t a l In t h e opinion of the proc o s t could be reduced t o perhaps $30b/FX. j e c t director, the f u e l c o s t could be reduced t o 1.7 m i l l s / ! S W H by o f f - s i t e processing-%, including f u e l charges and chemical processing operation and maintenance

.

2.

Breeding

Present conceptual designs would be break-even breeders a f t e r equilibrium uranium isotopic balances a r e attained. These designs include a two-region homogeneous r e a c t o r and a single-region graphite moderated r e a c t o r with a 201 diameter c y l i n d r i c a l core. A two-region externally cooled graphite moderated breeder with a graphite core s h e l l would have a conversion r a t i o of 1.05 a f t e r an equilibrium d i s t r i b u t i o n of uranium isotopes is attained. This yields a doubling time of 44 years on the r e a c t o r design postulated. It is possible t h a t with improvements i n means of power removal, s h o r t e r doubling times a r e a t t a i n a b l e The two-region homogeneous break-even


-

5f

-

breede:r, a f t e r it is supplied with U-233, would have power costs about equal t o those listed f o r the economic reactor. The one-region break-even breeder might a l s o have no higher power costs than the economic reactor. It does, however, require development of a s a l t p u r i f i c a t i o n process t h a t can handle the e n t i r e charge of s a l t about once a month.

.

No estimate has been made f o r the two-region graphite moderated breeder

B. 4

The Approach 1. Sumnram of Concept a.

Basic Reactor Concept

The Xolten S a l t Reactors u t i l i z e molten f l u o r i d e s a l t s a s the solvents f o r both the f u e l and the f e r t i l e material. The f l u o r i d e s a l t s themselves have about half t h e slowing down power of graphite, and s o r e a c t o r s may be homogeneous with only s e l f moderation by the s a l t , o r may use graphite a s the p r i n c i p a l moderator. The container material, a nickel-molybdenum alloy, is compatible With both t h e s a l t and graphite. A large v a r i e t y of reactor types can be constructed using these basic materials of construction and u t i l i z i n g U-233, U-23s o r Pu f u e l and Th or U-238 a s a f e r t i l e material. b.

Low Cost Power Reactors

(1) For low c o s t power, a s i n g l e region e x t e r n a l l y cooled graphite moderated reactor u t i l i z i n g thorium a s the f e r t i l e material and U-235 a s the added f u e l is being given primary consideration. U-238 can be used a s a f e r t i l e material i n the same reactor a t s l i g h t l y higher power costs. (2) Another r e a c t o r type given considerable study is a two-region e x t e r n a l l y cooled homogeneous reactor. This reactor type avoids the complication of a graphite moderator, b u t has a lower conversion r a t i o and consequently a s l i g h t l y higher f u e l cycle cost. C.

Breeding Reactors

Three types of molten s a l t reactors a r e considered f o r breeding. I n order of increasing d i f f i c u l t y of construction, they are: the tworegion e x t e r n a l l y cooled homogeneoils reactor, t h e one-region externall y cooled graphite moderated reactor, and t h e two-region externally cooled graphite moderated reactor. The f i r s t of these, r e l a t i v e l y


.

easy t o construct, t r i l l j u s t barely breed, a s w i l l the one-region The two-region graphite moderated regraphite moderzted reactor a c t o r i s 'oet'cer a s a breeder but has a doubling t h e of about L O years

.

d.

The Special Merits of the Molten S a l t Concept

Tne molten s a l t reactor is Yne only one of t h e three f l u i d f u e l reactors t h a t does not require a slurry.. The use of solutions f o r both f u e l and f e r t i l e material simplifies mechanical design, avoids erosion problems, s i t i p l i f i e s some s t e p s i n chemical processing, reduces concern o w r nuclear i n s t a b i l i t i e s , avoids s e t t l i n g o r drainage problem t h a t may occur when a f u e l pmp stops, and decreases the recpired amowit of development work by a large f a c t o r . Corrosion is not a problem i n the molten s a l t reactor system f o r The basic mechanism of t h e corrosion t h a t temgeratwes r p to l3O!l(?F. does occur a t higher temperatures i s well understood; the s t a b i l i t y of the IXOR-8 a l l o y t o fluoride s a l t s does not depend. on a s w f a c e film or on additives. The s t a t e of t h e n a t e r i a l s and basic coxponent developnent program is sach t h a t a design and development program f o r an edxperimental reactor r e a c t o r could be i n a q g g a t e d a t once. The ultimate econonic p o t e n t i a l of the system can be properPy appraised within a ten-year period, s o t h a t t h i s reactor, i f successful, would j o i n t h e c l a s s of reactors llachisving econonic power within t e n yearst1 a s defined by the Ad Hoc Commitbe. The wide range of s o l u b i l i t y of U, Th and. Pu i n the f l u o r i d e s a l t s and Kne low thernial neutron cross-section of the Li7, 3e and F a t o m used i n t h e s a l t s makes them v e r s a t i l e for a v a r i e t y of reactors. The fluoride s a l t s (mixtures of L%, BeF2 and f i s s i o n a b l e and f e r t i l e f l u o r i d e s ) are excellent heat t r a n s f e r agents. The v o l m e t r i c heat capacity is high, kending t o y i e l d conpact h e a t t r a n s f e r system. 'l3e high r e a c t o r mean temperature (lls'O?F) allows g r e a t e r temperatwe d v f e r e n c e s i n heat exchangers and thus smaller heat exchangers. The OW p r e s s w e of the l i q u i d flJel reduces heat exchanger header thickness, pipe wall thickness, and reactor v e s s e l wall thickness ovcr the aqueous reactor. These f a c t o r s tend t o reduce t h e dir"ficu1t;y of naintenance operations. The low pressure a l s o reduces the vol-me required within the secondary containment vessel, and should s i n p l i f y p l a n t construction.


- 59 The high melting point of the f u e l i s an advantage from the polnt of view t h a t i n case of a s p i l l of f u e l it w i l l tend t o s o l i d i f y i n t o a re]-atively water insoluble s o l i d , thus tending t o prevent spread of f i s s i o n product, contamination. e.

Limitations of the P?olten S a l t Concept

The high melting point of t h e f u e l s a l t requires equipment f o r preheatirg the system t o the reactor temperature. It a l s o requires conti-ol of the coolant temperatwe s o t h a t it does not freeze t.he f u e l i n the heat exchanger, S o d i m r e a c t s with the f u e l t o p r e c i p i t a t e Uranium. I f sodium is used a s a primary coolant, the f u e l s a l t nust be pressurized w i t h respect t o the sodium s o t h a t any f u e l leaks w i l l be i n t h a t d i rec-t,%on

.

Oxygen i n excess of limited amounts w i l l r e a c t w i t h the f u e l t o p r e c t p i t a t e uranium, s o t h a t the f u e l system must be buffered t o prevent contamination by a i r o r moisture. The maximm conversion r a t i o and minimum doubling time a r e dict a t e d by those a t t a i n a b l e i n a graphite noderated system. These a r e not as good a s those obtained i n the aqueous homogeneous reactor. The a l l o y used f o r construction and the high p u r i t y f u e l s a l t s w i t h t h e i r Li-7 con-lent a r e expensive materials, 2.

The Development of Law-cost Power P o t e n t i a l i t i e s a.

The Reference Design f o r the Economic Plant

(1) Description The reference design reactor* is e s s e n t i a l l y a c y l i n d r i c a l graphi t e assembly 12.25 f e e t i n diameter by 12.25 feet high containing moft e n s a l t f u e l i n v e r t i c a l channels c o n s t i t u t i n g s W e n percent of the volume of the reactors. The f u e l s a l t is circulated from the reacto]? vessel t o four primary h e a t exchangers by sfour f u e l pumps. An i n e r t coolant s a l t i s used a s the intermediate heat exchange f l u i d , which is used t o superheat and reheat steam i n a Loeffler b o i l e r system. The reactor develops 760 14W of heat. By using 2000 p s i steam a t lOOO?F with lOOO* reheat, the net e l e c t r i c a l output i s 318 1W e l e c t r i c a l . -I--.-�.-I--

4:

For d e t a i l see OFtNL-CF-59-1-26

J


- 60 (a)

Proposed Fuel System and Cycle

The f u e l s a l t has an i n i t i a l composition of 0.3 mole percent percent "4, 16 mole percefit BeF2 and 70.7 mole percent L' There a r e 900 cubic f e e t of f u e l s a l t i n the system, and the The f u e l s a l t i s withdrawn i n i t i a l inventories a r e 829 kg of Li'7. from t h e r e a c t o r p r h a r j c i r c u i t a t a r a t e equi-valent t o 20 percent of t h e f u e l inventory p e r year and placed. i n 150 day hold-up tanks t o await chemical processing. A s the f u e l is withdrawn it i s replaced with f r e s h f u e l of the o r i g i n a l composition except t h a t the WL cont e n t is adjusted t o keep the r e a c t o r c r i t i c a l , The uranium add-itions would. comprise a l l of t h e mixed U-233 and U-235 recovered from the chemical p l a n t together with f r e s h highly enriched U-235.

, 13 mole

md.

The equilibrium inventories of uraniwrt isotopes i n the reactor The average primary system a r e 51C kg of U-233 and 430 kg of U-235. conversion r a t i o i s 0.67 s o t h a t 92.5 kg of U-235 per year m u s t be a6ded. t o the system. Thorium i s burned a t a r a t e of 168 kg per year.

Fuels The above f u e l solution would operate a t a r e a c t o r i n l e t tempera-

ture of 107!??F and an o u t l e t temperature of 1 2 2 5 9 (580째C and 663OC). !!&e liquidus temperature f o r this system is 521i째C, t h e f i r s t s o l i d W k . If t h e composition were changed precipitatifig being 3 LW by addition o r removal of one constituent a t 500째C, the following a r e the l h i t s outside of which p r e c i p i t a t i o n would occur: ThFk, 8-21 mole $; BeF2, 11-40 mole $; L i F 53-74 mole %. Eo complications occur from heating. The s o l u b i l i t i e s of a l k a l i f l u o r i d e s and a l k a l i n e e a r t h f l u o r i d e s a r e q u i h high; t r i v a l e n t fluorides, l i k e those of plutonium, uranium and t h e r a r e e a r t h s w i l l cop r e c i p i t a t e when the t o t a l concentration exceeds a c r i t i c a l limit. The s o l u b i l i t y of r a r e e a r t h f l u o r i d e s has not been measured i n t h i s p a r t i c u l a r s a l t composition; from measurements on the s i m i l a r system 70 LiF 10 BeF2: 20 UF4 the s o l u b i l i t y a t 565oC is estimated t o be 1.0 percent, or s u f f i c i e n t t o accommodate t h e 0.k mole percent of r a r e e a r t h f i s s i o n products t h a t would build up during a nine-year f u e l l i f e . Certain heavy metal f i s s i o n products, such a s Xo, Ru and Nb perhaps form a s elemental metals on the walls of the system. Heat or other e f f e c t s from these might be a consid-erable problem. The t o t a l q u a n t i t i e s of t h e three formed i n nine years operation i s estimated t o be about 430 kg. If uniformly d i s t r i b u t e d , with a density of 7 g/cc, a l a y e r of about 1 m i l thickness would form.

-

An a l t e r n a t e f u e l c o n s i s t s of 20 mole $ UF4 (1.3% enrichment), 10 mole % , BeF 70 mole 5 LiF. The liquidus temperature f o system is 500%C, the first s o l i d p r e c i p i t a t i n g being 7

,

LiJ'*G?


- 63. If the composition were changed by addition or removal of one cons t i t u e n t a t s80째C, the following a r e the l i m i t s outside of which p r e c i p i t a t i o n would occur: UFb, 12-28 mole $; BeF2, 3-30 mole $; LiF, 61-76 mole d e PUF3 is formed i n t h i s f u e l a s w e l l a s the r a r e e a r t h f l u o r i d e s , For pure PuF3 i n t h i s f u e l , the s o l u b i l i t y l i m i t i s 1.38 mole $ a t 5 6 5 O c . Thus p r e c i y i t a t i o n appears t o l i m i t the unprocessed operating period t o about 15' years. Fuel Problems Although many aspects of the molten s a l t f u e l system appear promising, there a r e a number of uncertainties which might have a serious e f f e c t on the cost o r operating procedure, These a r e a s follows : Possible reaction of gases i n the graphite with W),, w i t h f o r pation of U02F and U02, during i n i t i a l charging of the r e a c t o r core m m s ? i l h h a reaction w i l l occur unless the graphite is corrp'letely f r e e of a i r , moisture, e t c , The graphite m u s t be pret r e a t e d t o remove a l l residual oxygen. Although no method of pretreatment of the graphite has been demonstrated, the p r o j e c t people a r e optimistic t h a t treatment with HF o r pretreatment w i t h a spare charge of s a l t w i l l reduce the oxygen t o s u i t a b l e levels, Accidental a i r leakage. A s described i n t h e previous paragraph, U02 would be precipitated. if a i r (containing moisture) leaks i n t o the f u e l system. Some U02F2 and/or UF6 might a l s o be formed with r e s u l t a n t corrosion problems. If UO2 is precipitated, processing of the f u e l out of p i l e w i t h H2 and HF may be required which would be d i f f i c u l t on account of high r a d i o a c t i v i t y , This might require a s long a s two months, since it would not be p r a c t i c a l t o i n s t a l l s u f f i c i e n t capacity f o r doing the job i n less t i m e ,

-

Burnable poison vs control rods. With %he U-238 U-23s f u e l cycle, there is a s i g n i f i c a n t increase i n r e a c t i v i t y a s Pu239 s t a r t s t o build up, Poison must be added e i t h e r a s a burnable poison t h a t i s soluble i n the f u e l , o r a s control rods, For the ultimate r e a c t o r design, a preference has been s t a t e d f o r t h e former. No s u i t a b l e poison has y e t been demonstrated, Change i n oxidation s t a t e s i n f u e l due t o f i s s i o n , When each U or h i atom f i s s i o n s , two nuclide species a r e formed, each of which (except f o r t h e r a r e gases) requires fluorine atoms t o stay i n solut i o n , (If a s u f f i c i e n t number of f p l e f t t h e f u e l solution a s Ronfluorides, it i s conceivable t h a t an excess of f l u o r i n e atoms would be present, with the r e s u l t t h a t v o l a t i l e and corrosive uF6 would be formed.) A t t h e end of 9 years continuous operation without reprocessing, about 0,6 m/o of the t o t a l metal atoms i n t h e f u e l solution will


-

62

-

have fissioned, Discounting the iodine and bromine formed there i s a nearly exact f l u o r i n e balance i f 1) the r a r e gases (22% of t o t a l f p ) a r e removed as elemental species, 2) the r a r e e a r t h s (265 of t o t a l f'p) form s t a b l e trifluorid.es, 3) Ru, XO and. Nb (22% of t o t a l fp) p r e c i p i t a t e a s metals and. 4) other f i s s i o n products (30$, 1/2 of which i s Z r ) a r e i n solution a s t e t r a f l u o r i d e s .

One l i m i t t o continued burn-up without processing i s the cop r e c i p i t a t i o n of the t r i f luorides. An e a r l i e r l i m i t , however, may be the p r e c i p i t a t i o n of Ru, Mo and Nb which may cause plugging of h e a t exchanger tubes, If such p r e c i p i t a t i o n occurs (afid t h i s is the current view) it might l i m i t the l i f e of the heat exchangers with o r without reprocessing, since reprocessing would not remove t h e p r e c i p i t a t e d metals. Some system f o r removal of these metals should be developed. 14a$erials The basic material being considered a s a container f o r the molten s a l t f u e l system is a nickel-base a l l o y designed as INOR-8 (Hastelloy N, Inconel 806). It is a s o l i d solution type a l l o y not subject t o The only metallurgical i n s t a b i l i t y is soxe carblde age-hardening, p r e c i p i t a t i o n a t long exposures t o high temperahre with some I w r e a s e i n strength 'out Tdth no reduction i n d u c t i l i t y . Corrosion and Xass Transfer, This a l l o y appears t o have no basic problem from a corrosion and mass t r a n s f e r standpoint. No measurable a t t a c k has occurred during one year exposure i n l a w uranium s a l t a t 677OC ( l e s s than .$ m i l ) . A t 899'C about one m i l deep p i t t i n g a t t a c k was observed i n b2 days exposure, I r r a d i a t i o n , per se, i s not expected t o have any s i g n i f i c a n t eff e c t on these properties since the a b i l i t y of INOR-8 t o r e s i s t corrosion and mass t r a n s f e r does not depend on "surface films", The corrosion and mass t r a n s f e r t h a t does OCCUT is a consequence of the s a l t leaching out chromium f r o n the surface of t h e container i n t h e hot l e g (Cro -i- UF4 +b CrP2 i UF3), thereby decreasing t h e C r content of t h e a l l o y u n t i l an e q u l i b r i o n concentration of about 500 ppm i s a t t a i n e d is the s a l t . I n t h e cold leg, C r is deposited on the metal surfaces; it there d-iffuses inward and increases the surface C r contznt. The process will continue u n t i l steady s t a t e r a t e i s s e t up, t h e lower s o l u b i l i t y of CrZ'2 i n the cold Leg being compensated by the higher C r a c t i v i t y i n t h e metal. Weldability and f a b r i c a b i l i t y of INCR-8 i s comparable t o Inconel and the standard Hastelloys. r n z t gas shielded tungsten a r c weldi3g i s used, No s p e c i a l d i f f i c u l t y has been encounAwred, S u f f i c i e n t q u a n t i t i e s have been produced t o indicate t h a t no problea w i l l be encountered with respect t o procurement.


Graphite. Penetration of graphite by nolten s a l t s i n the L S , R e F 2 , - ~ y s t e mhas not been successfally accorrplished, These t e s t s were ~ u n ,however, wi+A e i t h e r poorly outgassed or as-received graphite. Unfms further t e s t s , including long time i r r a d i a t i o n , show otherwise, it must be assuned t h a t penetration W i l l occur. Since t h e s a t i o of graph?-te t o f u e l volume i n the core i s nuch higher f o r the molten s a l t r e a c t o r than f o r the LXFR, a greater f r a c t i o n of the f u e l would be t i e d up i n the graphite. Other 7ilaterials. For valve s e a t application, a combination of Pi0 again& a Go bonded T i c cermet has been s a t i s f a c t o r i l y used a s a hard contact s e a t , and copper against molybdenum appears s a t i s f a c t o r y a s a sof t-seating conbi2ation. (b)

The Fteactor

This vessel must completely and i n d e f i n i t e l y contain a high k m p e r a t w e f l u i d (12509), which a l s o has a r e l a t i v e l y high melting point (about 9 3 0 9 ) , a t a low pressure (about 100 psig). The v e s s e l provides support f o r the graphite moderator and the means by which the molten, homogeneous s a l t rnixture i s admitted and exited from the c r i t i c a l region. Ibe reference design suggests a r i g h t cylinder about 1 2 f e e t i n diameter and about 1 2 f e e t i n height. This DJOR-8 v e s s e l is about 1.5 inches thick. A multiple loup pipe header is welded t o t h e vessel a t both poles providing the entrance and. e x i t f o r the fluid. The molten s a l t flaws upward through v e r t i c a l channels constituting 16 p e r cent of the volume of the graphite moderator. A'oout one-half of the s a l t leaving the c r i t i c a l region flows upward t o an expansion tank, o r dome, wherein some of t h e gaseous f i s s i o n products a r e rexoved. A. r e l a t i v e l y close f i t t i n g , l o w chrome steel vessel c o q l e t e l y

H e l i m gas, pumped through the annulus between the outside of t h e r e a c t o r vessel and t h e inside of the contairumnt vessel, is used t o cool or h e a t the r e a c t o r vessel and- its interrtals. The helium gas pressure required is about 75 psig; therefore, a r e l a t i v e l y thick containment vessel need not be designed. The support f o r t h e gas jacket probably needs t o be integrated with t h a t f o r the r e a c t o r vessel b u t present design p r a c t i c e s can be used and no further d e v e l q n e n t appears t o be needed.

s-nroxtnds the reactclr vessel.

The use of t h e graphite moderator within the r e a c t o r vessel appears t o present several design problems. It is desirable t o try t o minlmi.ze the v a r i a t i o n i n temperature of the graphite i n a given horizontal. plane during the period following drainage of the f y e l from the reactor. The h e a t generation r a t e a t t h i s t i n e depends upon the amount of f u e l and f i s s i o n products l e f t on the surface of the graphite and the q u a n t i t i e s t h a t have penetrated i n t o the graphite. Experinental


investigation of t h i s problem, especially i n in-pile loops, i s necessary. It is envisioned t h a t a considerable experimental e f f o r t is dictated with a t o t a l expenditure of the order o f two million dollars. It has been shown t h a t the t h e m a l conductivity of graphite decreases with t o t a l e-xposure t o f a s t neutrons; t h i s e f f e c t is not known i n the 1l00-12004;1 tempera t u r e range and should be investigated edxper-hentally. A conpromise between the most desirable nuclear configuration and the predicted heat t r a n s f e r perfomance is d i c t a t e d i n the select i o n of the hole s i z e s i n the graphite and t h e i r spacing. A reactor vessel mockup i s required t o insure t h a t the desired f l o w d i s t r i b u t i o n provides adequate cooling f o r the reactor v e s s e l a s w e l l a s t3ie graphite moderator. It has been found t h a t the f u e l s a l t does n o t always wet the graphite and consequently the h e a t t r a n s f e r work t h a t has been done on s a l t systems does not apply; a very modest e f f o r t i s required t o deternine the heat t r a n s f e r between the f u e l s a l t and the g a p h it e

.

Heat t r a n s f e r between the metal walls and flowing f l u o r i d e s a l t m i x t u r e s has been defined by experhentalwork. Therefore, the designer is able t o specify the flow r a t e required t o adequately cool the vessel without f u r t h e r experimental work. Very few nuclear instruments a r e required f o r successful operat i o n of t h i s reactor. It i s not necessary f o r t h e sensing elements of these instruments t o reside within the reactor v e s s e l or within the contaiment vessel. It w i l l be necessary t o supply i n e r t gases t o the sensing elements and t o adequately gas-cool them. However, there is no new technology involved. F i e l d f a b r i c a t i o n of reactor vessels of the order of 2 3 f e e t i n diame.%er and 20 f e e t i n height i s f e a s i b l e with present construction practices. The wall thickness required f o r s a t i s f a c t o r y perfomance of such a vessel is about four inches and this thickness i s near the maximum f o r present f i e l d f a b r i c a t i o n practices. (c)

Primary Heat Exchange Sysben

The 318 FNF, n e t reference design proposes t h e use of four parall e l primary loops. The combined core and blanket s a l t flow flows from the r e a c t o r a t an average bulk temperature of 1225% through 18" piping d i r e c t l y t o four priinary loop canned r o t o r sump type c i r c u l a t ing pumps each rated 8,900 gpm a t 75 psi. The f u e l s a l t then circulates through t h e tube s i d e of four 5830 square f o o t p r h a r y h e a t exchangers where it gives up 195 Mbd thermal and r e t u r n s t o tihe rea c t o r a t 107#?F. The loops and a l l p r i n c i p a l equipment a r e f a b r i c a t e d from INOR-8.

c


- 65 The secondary f l u i d is a 65 mol percent Li7F - 35 mol percent BeF2 窶話arren s a l t selected f o r compatibility with t h e f u e l s a l t which i n t u r n generates and reheats the steam used i n t h e e l e c t r i c a l genera t i n g f a c i l i t i e s a r e presented a s a Loeffler tyye b o i l e r , I f t h i s feattcre of the p l a n t should prove undesirable, a l m - n e l t i n g lit!!imn, rubid:ium chloride e u t e c t i c s a l t could be used a s a secondary f l u i d tfiat :is compatible with the f u e l . The corrosion resistance of the INOR-8 a l l o y t o the chloride s a l t has not been tested. Heat ?xchansers

No d e t a i l e d reference design has been presented. Extensive work has been performed on the heat t r a n s f e r c h a r a c t e r i s t i c s of the flimride s a l t s and the design parameters have been optimized. The general description of the reference design is a bayonet type tube bundle mouqted i n a v e r t i c a l s h e l l t o permit tlsemi-direct!r replacement. Although simple i n concept the molten s a l t exchanger will be complex i n d e t a i l because of the need f o r preheating t h e primary and secondary loops, provision f o r remote maintenance, drainabilit-y, etc. These,, however, a r e problems of engineering d e t a i l only and seem t o pose no insurmountable obstacle, although it may be anticipated t h a t extensive design and development work w i l l be required t o produce a The u l t b a t e exhanger design shauld s a t i s f a c t o r y confi,mation. prove q u i t e costly, running i n the neighborhood of $100 per square f o o t of ac;t;ive surf ace. The p r o b a b i l i t y and consequences of leakage a r e not conditioning

on t h i s design since it is possible t o maintain the secondary f l u i d a t s d f i c i e n t over-pressure t o a s s m e leakage of t h e secondary f l u i d i n t o the primary f l u i d a t reasonable r a t e s with no deleterious efforts.. Leakage of major proportions w i l l require shuMown for maintenance and/or replacement of the a f f e c t e d component, The low vapor pressure of the primary f l u i d permits the use of t h i n wall tubes and presumably n o t too heavy tube sheets. The benef i t s accruing from t h i s consideration cannot be evaluated u n t i l a d e t a i l e d reference design is established, Corrosion of the exchanger and f u e l hide-out do not appear t o be serious problems on the b a s i s of work t o date. Pwther, the high thermal capacity of t h e f l u i d permits t h e w e of a r e l a t i v e l y small temperature d i f f e r e n t i a l across the reactor. This should f u r t h e r minimize t h e design problems associated with the t r a n s i e n t thermal s t r e s s e s of the tube sheets.

Fuel Circulating Puprrps No molten s a l t pumps with t h e size, performance and gas leakage requirements of t h e FSR conceFt have been b u i l t . However, molten

E


- 66 -

1

i 2

s a l t pumps up t o 800 g p m have been operated s-ilccessfu7.J-y on nonradioactive loups and pumps rated up t o 1500 gpm a t heads t o 350 f e e t have operated successfully a t temperatures t o 1525?F, These are v e r t i c a l s h a f t centrifugal sump t n e puiis with the impeller on the end of the s h a f t ,

Punrps t e s t e d t o date have been of the oil-lubricated type with elastomer s e a l s , For t h e long l i f e requirenents of pumps f o r the power r e a c t o r concept it i s necessary t o develop s a l t or gaslubricated bearings, and l o w leakage r o t a t i n g mechanical gas s e a l s , Stationary elastomer s e a l s could be replaced with metallic 0 rings. Heutron and gama shielding i s required ’for protection of t h e motor wind.ings and lubricant. The reference design i n d i c a t e s a gascooled s h i e l d of r a t h e r complex design witn a lower salt-lubricated journal bearing, a mechanical gas s e a l , oil-lubricated motor bearings. Pumps of similar mechanical design’have been used f o r molten s a l t s i n connection with the AMP program and it i s believed possible t o extrapolate t h i s experience t o the c a p a c i t i e s required f o r a u t i l i t y s i z e r e a c t o r plant, Piping Although t o date no l a r g e f i t t i n g s o r l a r g e diameter tubing have been procured, p l a t e is available i n IMOR-8, It i s believed t h a t piping procurement w i l l not pose a problem, The p r i n c i p a l d i f f i c u l t i e s associated w i t h the piping a r e the design of loops with s - d f i c i e n t f l e x i b i l i t y t o reduce t h e m a l expansion s t r e s s e s t o permissible levels. The thermal expansion of INOR-8 i s of the same magnitude a s the f e r i t i c s t e e l s , The low vapor pressure of the primary and secondary f l u i d s permits t h e use of t h i n wall piping which i s of some b e n e f i t i n designing f o r mechanical f l e x i b i l i t y . The close coupled system proposed i n the reference design would require f l e x i b l e mounting of the r e a c t o r t o obtain permissible ther3 a l s t r e s s e s i n Vne piping between t h e reactor and the pumps, A d e t a i l e d t h e m 1 s t r e s s analysis of t h e piping has not been made and it is doubtful whether the confi,wations presented would stand up t o such scrutiny. The proposed primary system holdup volume may be uqduly optimistic because of t h i s f a c t o r , pl;imary . \ m i l i a r y Systems The functions performed by the equipnent included i n t h e NSX primary a u x i l i a r y system are: (I) heat adtirlticr: snd removal required during abnoma1 operations; (2) removal of v o l a t i l e f i s s i o n

t


products; ( 3 ) sanrplin~, enriching, and removal of primary f l u i d ; and (4) s u b - c r i t i c a l storage of primary fluid-.

A l l of the primary system components a r e heated t o about 1200% p r i o r t o charging of the f l u i d f u e l because the proposed mixture of The components a r e heated by fluor5.de s a l t s melts a t about 950%. c i r c u l a t i n g hot helium gas through the annulus formed by the priKary b a r r i e r and t h e gas jacket. The heating system i s compartmentalized t o allow rough control of the r a t e a t which the temperat u r e s a r e changed. Some of the equipment, e.g. the d r a i n tanks, w i l l be maintained near operating temperature, although the equipment i s not i n use. The s i z e of the blowers and gas or o i l fired. or e l e c t r i c heaters is determined by the p a r t i c u l a r components t o be heated by them. It is expected t h a t the sum of t h e heating capacit i e s of the furnaces w i l l be about 2 Filw and the t o t a l blower horsepower will be about 1000 HP, assuming an acceptable temperature r i s e of the heliuri of 2OO0F through the furnace and an operating pressure of about 75 psig. The r e a c t o r v e s s e l heating system is interconnected t o a gas-to-water heat exchanger capable of removing about 3 N4. A l l . of t h i s equipment is located i n a semi-radioactive area because it may become contaminated during operation. The f l o o r area required t o house these a u x i l i a r i e s i s about 7500 square f e e t . The gaseous f i s s i o n products a r e collected i n the expansion tank which i s attached t o the top of the reactor vessel and a l s o i n the sunrps of the primary pwis. The pressure i n the gas spaces (volume of about 250 f$) is allowed t o increase until. about 5 p s i g overpressure is evident (occurs i n about one month) and is then c a r e f u l l y vented t o a hold tank of about 1000 f $ volume. The gases a r e held i n the large tank f o r six months t o a year and. then compressed i n t o high pressure cylinders; it i s anticipated t h a t these cylinders w i l l be buried. The s o l u b i l i t i e s of the relevant gases i n s a l t s s i m i l a r t o the proposed f u e l s a l t have been deterrnined but t h e i r s o l u b i l i t i e s i n the p a r t i c u l a r f u e l s a l t chosen have not been examined experimentally. I n addition, it appears t h a t the behavior (e.g. fugacity) of t h e f i s s i o n product halogens i n the f u e l f l u i d should be detemnined experimentally. A modest expenditure is indicated (approximately $200,000)

0

Three batchwise manipulations a r e reqQired t o s u s t a i n steady Daily additions of f u e l a r e required t o r e c o n s t i t u t e the f u e l , and. d a i l y samples by means of a mechanical lock and v e r t i c a l t r a n s p o r t assembly which handles s o l i d cylinders of f u e l s a l t m i x t u r e . The assembly penetrates one primary pump sump. This assembly has not been designed but it does not appear t h a t any development problems a r e involved. Fluid f u e l may be removed from the primary system by controlled, pressurized t r a n s f e r from one purrip sump t o a spent f u e l storage tank; a r i g h t cylinder f i v e f e e t i n

state operation of t h e EGR.

t

i

i


- 66 diameter i s capable of s t o r i n g the spent f u e l f l u i d . This vessel need be heated only during t r a n s f e r from it to the chemical reprocessing p l a n t or t o shipping casks; it i s jacketed and is heated or cooled by means of c i r c u l a t i n g helium gas,

It is necessary t o d r a i n the f u e l f l u i d from the r e a c t o r and primary h e a t exchange loop p r i o r t o some maintenance operations. During t h e course of these operations the f u e l f l u i d is stored i n an a r r a y of inclined pipes contained within a s h e l l ; the u n i t consists e s s e n t i a l l y of a cross-flow h e a t exchanger w i t h cool gas entering a t the top and e x i t i n g a t the bottom. About 2700 square f e e t of a c t i v e surface is required i n each exchanger. This method of temperature control is u t i l i z e d t o eliminate the need f o r remote maintenance of e l e c t r i c a l resistance heaters, bayonet tubes and the l i k e , The helium flaw r a t e needed t o maintain the s a l t temperature a t 16000~is about 3000 pounds per minute, Five blawe r s of about 250 HP each (1 standby) appear to be required. These blowers a r e a l s o those used f o r the heatup cycle. Two atnospheric water b o i l e r s of moderate s l z e a r e suggested t o remove the h e a t from t h e r e c i r c u l a t i n g helium gas; t h e h e a t t r a n s f e r area of each i s about 500 f t 2 and the water r a t e t o each i s about 100 gpm. A l l of these components can be designed from present technology; however, it does appear prudent t o determine the a f t e r h e a t generation. Fuel f l u i d i s transferred t o t h e d r a i n tanks from the r e a c t o r system by pressurizing t h e r e a c t o r s y s t e m with clean helium gas. Fluid f u e l i s pumped i n t o the reactor system from t h e drain tanks by means 6f a v e r t i c a l sump type c e n t r i f u g a l pump s e t i n a closed sump located a t t h e l o w p o i n t of the d r a i n tanks. "he capacity of the ptmrp i s about 200 g p m a t a discharge pressure of about 25 f e e t of f l u i d . During the t r a n s f e r the r e a c t o r vessel i s vented t o the hold-up tank of t h e off-gas s y s t e m ; t h e v o l a t i l e f i s s i o n products i n t h i s gas a r e removed and t h e p a r t l y contaminated helium gas is compressed and stored i n gas cylind-ers f o r re-we. (d)

Secondary Systems and Steam Generators

The secondary s y s t e m c o n s i s t s of an intermediate coolant s a l t t r a n s f e r r i n g h e a t from the primary h e a t exchangem t o steam superheaters and reheaters, A l a r g e portion of the superheated steam is used t o generate steam i n a Loeffler boiler. A steam blower c i r c u l a t e s steam from the b o i l e r s t o the superheaters. Problems of the Loeffler system a r e discussed i n Section ( f ) below, under Maintenance

.

The L m f f l e r b o i l e r s y s t e m imposes a c a p i t a l c o s t penalty t o t o the molten s a l t system. It was chosen s o t h a t a high melt'ng p o i n t f l u o r i d e s a l t t h a t is compatible with the f u e l could be employed a s a s i n g l e h e a t t r a n s f e r loop between the f u e l and steam systems

.


- 09 (e)

Control, Instrumentation and Operating Characteristics

The f e a s i b i l i t y of the operation of a molten s a l t fueled rea c t o r a t high temperatures was demonstrated i n 1954 i n experiments The r e a c t o r was equipped w i t h the A i r c r a f t Reactor ;"xperi.ment (ARE), with a conventional control rod system, but it was demonstrated t h a t C O C ~ P O ~could be achieved by use of the negative temperature coeffic i e n t of r e a c t i v i t y . It was found t h a t the r e a c t o r would follow any changes i n the h e a t extraction system. The r e a c t o r proposed i n the reference design would make use of the negative tenperatwe c o e f f i c l e n t f o r its prinlary control. It may be necessary t o have control rods, a s i n the case of the L?II?R, t o l - i n i t excessive teinperature r i s e s i n the core due t o a temporary excess of f u e l . Only preliminary work has been done on the poss i b l e start-up, shut-down, and system operational procedures, The start-up operations of t h i s reactor system w i l l be complex due t o the need of preheating a l l equipment and piping; likewise, i n shutdown operations, t h i s equipment w i l l have t o be cooled. Neans have t o be provided f o r preheating and cooling of both the primary and secondary systems. The time required i n start-up w i l l be determined t o a g r e a t extent by the time required t o preheat the systems, The start-up of t h i s p l a n t requires e i t h e r steam from an outside sow'ce or e l e c t r i c heaters t o s t a r t the Loeffler cycle, Once the r e a c t o r is c r i t i c a l and t h e secondary system and the i o e f f l e r cycle a r e i n operation, control would be a combination of negative tempera'cure coefficient, v a r i a t i o n of f l u i d flaw i n the secondary system, and v a r i a t i o n of steam f l a w i n the Loeffler cycle. Information on p l a n t s t a b i l i t y and its k i n e t i c behavior f o r disturbances i n flow, power, and r e a c t i v i t y w i l l have t o be obtained from analogue studies and r e a c t o r experiments. This information would a i d i n designing an over-all p l a n t control system. Extensive instrumentation w i l l be needed f o r start-up and shutdown operations p a r t i c u l a r l y with regard t o thermocouples. These w i l l be required t o indicate the thermal cond-itions of a l l piping and equipment. An additional problem during start-up and shut-dawn, when there i s no flowirg molten s a l t i n e i t h e r primary o r secondary systems, is t h e possible d i s t o r t i o n of the pump i n t e r n a l s due t o uqeven h e a t i w or cooling, This d i s t o r t i o n could prevent the s t a r t up of t h e pumps,

I n addition t o operation of the main plant, there is a l s o the operation of supporting systems. These supporting systems w i l l need extensive control c i r c u i t s and instrumentation t o indicate t h a t they a r e functioning properly. The supportins systems would


-

70

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include t h e dunp tanks, c e l l cooling equipment, off-gas system, and eqdptnent f o r maintaining an inert; gas atmosFhere i n the r e a c t o r cell,

'--

An area, which is important i n t h e control of t h e reactor, i s t h e control of f u e l inventory i n the system. This w i l l require some mechanism which would obtain a sample of t h e molten s a l t , the sample would be moved t o a hot c e l l where it would be analyzed fop i t s uranium and t h o r i m content. The control of t h e inventory of rnolten s a l t in the primary sys+ten may require t h a t t h e dump tanks a l s o a c t a s weigh tanks, p a r t i c u l a r l y when f i l l i i g o r dumping the p r h a r y system, An a l t e r n a t i v e f o r weigh tanks would be a system of tank l e v e l indicators. The pruposed reference design is s t i l l i n i t s very e a r l y s t a g e s and a g r e a t deal of work remains t o be done before a d e t a i l e d opera t i n g procedure can be developed.

(f) Special Maintenance lems

-

- Facilities

and Prob-

Inherent Problems Many of the problems i n maintenance associated with the molten s a l t r e a c t o r concept are similar t o those encountered with the molt e n bismuth r e a c t o r concept, The f u e l c a r r i e r (LiF, BeF2) has a very much higher melting point than the ambient tenperatwe, theref o r e preheating of t h e p l a n t equipment p r i o r t o the in-broduction of the molten s a l t is a necessity, Likewise, a cooling system is needed t o cool the r e a c t o r a f t e r t h e nolten s a l t i s removed from the system, This concept has an additional problem i n t h a t t h e secondary systern uses a fluoride s a l t , which has a high melting point. The necessity of t h i s fluoride s a l t i s f o r reasons of compatibility with the f u e l i n case of a leak between the primary and secondary system, The secondary system would therefore have t o be equipped with a heating system.

The proposed heating and cooling system f o r the r e a c t o r vessel, pri-nary pumps and h e a t exchangers, would use helium gas, This necess i t a t e s a gas-tight outer container around each piece of equipment and piping. This type of construction g r e a t l y complicates the problems of maintenance, with the g r e a t e s t d i f f i c u l t y a r i s i n g when piping i n e i t h e r the prixary o r secondary system has t o be c u t t o remove a major component. No methods have been proposed f o r carrying out t h i s type of operation bu-b methods develuped i n t h e molten bismuth concept would be applicable f o r t h i s concept. It has a l s o been indicated i n the design t h a t some p a r t s of the primary system would be heated by the use of f'conventional e l e c t r i c heater-insulation"; the s h o r t service l i f e of this type of equipment nay cause extreme problems i n maintaining such a system remotely,


- 71 A f u e l f i l l and drain system has been provided t o serve a s a molten s a l t storage f a c i l i t y before the plant is s t a r t e d and a s a drain system when the primary system has t o be emptied. The maintenance of the drain tanks may prove very d i f f i c u l t due t o the number of .connections t h a t have t o be broken remotely f o r removing sections of it. An additional problem w i l l be the maintaining of the extensive instrumentation t h a t w i l l be needed.

The reactor vessel with i t s graphite core i s the l a r g e s t component i n the plant. Wo maintenance except on control rods would be performed on the reactor proper, e.g. vessel, graphite, and any repair work, replacement of graphite or work on the reactor vessel is considered a plant modification.

.

Only very preliminary design work has been done f o r the primary pumps and heat exchanger. For the primary pump, the f i n a l design would be one i n which the entFre pump i s removed and replaced a s a unit, with the pump casing a s a permanent p a r t of the piping s y s t e m . For the primary heat exchangers, the design would be of the sump type so only the tube bundle would have t o be removed.

4

i

The auxiliary systems f o r the molten s a l t f u e l reactor include molten s a l t transfer equipment, which includes the dunp tanks, offgas ,system, and preheating and cooling s y s t e m . The f u e l transfer system uses a v e r t i c a l sump type centrifugal pump, together with i t s associated valves which would be remotely maintained. The problems associated with valves f o r molten s a l t f u e l s are those of maintenance of allignment and self-welding of the closure. The off-gas system is used f o r the continuous removal of f i s s i o n product gases. This s y s t e m , because of the high activity, would have t o be maintained remotely. The proposed gas preheating and cooling system would use a package type maintenance, t h a t i s the blowers, heaters, and cooling c o i l s would be i n some type of container which could be removed remotely a f t e r all connections were broken. No detailed work has been done on t h i s containe ion system i s used t o prevent excess temperatures i n Am the reactor cell. This cooling of the c e l l s is done by means of forced gas circulation through radiation type space coolers. A coolLng medium such a s Dawtherm, i n a closed loop removes heat from the space coolers and dumps it t o a water heat exchanger. The gas blower and cooler would be b u i l t so t h a t it could be removed and replaced a s a unit. a l t a s the heat transf e r medium. For this s y s t e m maintenance w i l l be completely contact a f t e r the system i s drained. This secondary s y s t e m w i l l operate a t a higher pressure so any leakage would be i n t o tb primary system.

k


Although the f e a s i b i l i t y of the molten s a l t concept does not depend on the use of the Loeffler steam cycle, some consideration w i l l have t o be given t o the maintenance of t h i s system. A l l maintenance on the system would be d i r e c t contact a s there a r e no problems of radioactive equipment, Two important items i n the Loeffler cycle which may cause major maintenance problems a r e the steam blowers and the steam nozzles i n the boiler section. Much developmentwork i s needed t o obtain a large dependable steain blower. The solution t o the problem of erosion of the steam nozzles may require an extensive research program. The instruiientation and methods of maintenance of these various i n s t r m e n t s has not been gone i n t o i n the reference design. A leak detection system w i l l be needed f o r the primary system since a l l components and piping are enclosed by outer sealed containers. The detection s y s t e m I ~ I U Sserve ~ two functions; show t h a t a leak has occurred and, give the location of the leak.

NFthods Employed t o Solve Naintenance Problems The maintenance concept enployed i n t h i s design would be called "dry semi-contacttl, t h a t is, using both contact and remote methods

f o r the primary system. The layout of the plant consists of rectangular canyons w i t h the equipment located belaw tne f l o o r of the canyon, Located i n the canyon f l o o r are removable plugs over major equipment. The reactor, primary pumps and heat exchangers are located iq a c e l l which i s s t e e l lined and is gas-tight. An i n e r t atmosphere i s maintained i n the c e l l a t a l l t i m e s . The canyon above the biological shield, which forms the roof of the reactor cell, i s sealed and provided with a crane, boon mounted manipulators and viewing windows, This room would be accessible t o personnel t o a l low carrying out certain maintenance operations. I n the operation of replacing a f a i l e d p r h a r j j puxp, a f t e r Yne primary system i s drained, service connections snch a s power leads, instrument leads, gas lines, and prater cooling l i n e s would be removed manually. A l l personnel woald leave the compartment and the removal of the primp i n t o a container would be carried out rernotely wiYn the aid of tne viewing windows, The pump i s transferred t o a storage coffin, and a new pump i s lowered i n t o place. A f t e r the compartment i s decontaminated, personnel w i l l again e n h r and- xake up a l l service connections. The replacenent of a heat exchanger would follow much the saxe procedure with the exception t h a t the secondarj system pipes and a s e a l weld on the head of the heat exchanger would have t o be cut manually. A problem which a r i s e s during the changing of primarj pwrrps and heat exchangers i s the need of preventing large amounts of a i r fron entering the prinapj systen.


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becones c r i t i c a l , a t which point the temperatwe can be lowered no f u r t h e r because of power generatisn. Finally, lR?b w i l l be added t o bring the reactor t o i t s design operating temperature, a s a trinning uperation,

A s long a s no r e p a i r s a r e necessary on the primary f u e l loops, t h e r e a c t o r w i l l be on continuous operation, with both f u e l and coolant s a l t pumps going, A t zero power operation, the r e a c t o r f u e l c i r c u i t and coolant s a l t loop operate isothermally a t 1 1 ~ O ? F , the mean c r i t i c a l temperature, When there is a demand f o r power, steam i s s t a r t e d i n t o t h e superheater and t h e r e a c t o r picks up the load and w i l l automatically a d j u s t t o any load. There i s no r a t e of loss o r pick-up of load t h a t can cause nuclear or temperature overshoot d i f f i c u l t i e s i n the reactor-%. Load change r a t e s w i l l be l i m i t e d by allowable thermal s t r a i n s caused by rapid temperature changes i n h e a t exchangers and piping, When r e p a i r s a r e necessary on t h e p r b a r y loop, t h e f u e l w i l l be drained and the loop cooled. On restart,ing, t h e f u e l and the r e a c t o r should be heated above the c r i t i c a l temperature before t h e f u e l i s brought i n t o t h e r e a c t o r c i r c u i t , since it already contains t h e c r i t i c a l concentration of uranium. Simulator s t u d i e s would. ind i c a t e t h a t no hazard is involved even i f t h i s precaution i s not properly followed, Control of t h e reactor is primarily by thermocouples t o indicate temperatures. Although thermocouples w i l l be located throughout the system, it is intended t h a t routine r e a c t o r control (achieved by UF4 or ThF4 additions) would be based on thermocouples i n the coolant s a l t c i r c u i t where they a r e r e a d i l y available f o r maintenance, Other c r i t i c a l instrument requirements a r e those t o tell whether t h e pumps are running, and l i q u i d level indicators t o determine the amount of f u e l and coolant s a l t present. Hazards The following incidents may be consrdered: A steam leak i n t o a coolant s a l t c i r c u i t would lead t o drainage of t h a t coolant s a l t loop and r e p a i r s t o the superheater. The r e a c t o r need not be drained nor need t h e f u e l pumps be stopped, Afterheat could be d i s s i p a t e d through the undamaged superheaters,

i

Temporary loss of a l l coolant s a l t purrps would n o t cause a sudden temperature overshoot i n t h e reactor. If c i r c u l a t i o n of a t l e a s t one of t h e coolant loops can be established. within 15 minutes, draining of the r e a c t o r would not be necessary, since the temperature rise due t o a f t e r h e a t i s not catastrophic,

t


- 74 The proposed f u e l solution has limlted oxygen tolerances t o prevent the possible precipitation of U02 from the molten s a l t fuel. After a replacement of a pump o r heat exchanger itwi.13.. be necessary t o purge the primary system before the molten s a l t f u e l is introduced. The design of the molten s a l t reactor i s not f a r enough along t o have considered i n d e t a i l the performance of maintenance below the main biological shield i n the reactor compartment. Provisions w i l l have t o be mad-e i n the design t o be able t o replace the p r i mary heat exchanger shells, pwnp casings, primary piping and a l l instrumentation not accessible fron the access canyon. This would. involve tize use of remotely controlled equipment, perhaps similar t o t h a t proposed f o r the LMFR and PAX. Maintenance Equipment There has been no extensive development program of remotely operated t o o l s f o r the molten s a l t reactor. Work has been carried out i n the use of boom mounted manipulators for the disassembling and assernbling of small pumps. However, the use of t h i s technique i n a large reactor plant would not be practicable except i n a hot maintenance shop. Development work has been carried out on designing a flanged j o i q t f o r small pipes. This i s a freeze flange j o i n t t h a t consists of a conventional flanged ring j o i n t with a cooled annulus between the ring and the process fluid. The s a l t t h a t enters the annulus freezes and provides the primary seal. The r i n g provides a back-up s e a l against s a l t and gas leakage. The use of a flange connection may complicate the preheating and cooling system and its use might be a disadvantage. The maintenance equipment being developed f o r use with the ISGR and PAR concepts would be applicable t o the molten s a l t concept with changes i n plant layout. (g)

Na jor Spare Equipment Requirements

Major spare p a r t s carrled on hand would include one each of the removable cartridge portions of the putnps and primary heat exchangers and one each of the superheaters and reheaters. (2)

Reactor Operating Characteristics

The i n i t i a l start-up of the reactor w i l l require preheating both fuel and coolant s a l t loops, and then f i l l i n g these loops with hot s a l t from the drain tanks. Preheating may require several days. After f i l l i n g , circulation of the f l u i d s by the pumps w i l l be started. A f t e r a check-out period, the reactor w i l l be brought c r i t i c a l by additions of UF' When c r i t i c a l i t y i s approached, it w i l l be done by gradually owering the temperature of the f u e l until it

li .


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Loss of one f u e l pwnp would not have any temperature overshoot consequences, but would be cause f o r an orderly shut-down of the plant, draining the f u e l and replacement of the pump,

If e i t h e r a f u e l pump o r a coolant s a l t pump should. f a i l , the stean supply t o the corresponding superheaters m u s t be automatically shut off t o prevent freezifig of coolant s a l t in t h a t circuit. Loss of a l l f u e l pwips a t once would cause a temperature overshoot i n the reactor due t o the sudden gain of delayed neutrons i n the reactor core, The exact magnitude of the overshoot has not been determined for this reactor but it would probably not exceed 200%. The princiFal d.ifficulties would r e s u l t from thermal stresses (not yet calculated). Control or poison rods can control t h i s overshoot. It would be preferable, however, t o have auxiliary power available t o s t a r t one of the f u e l pumps, The principal nuclear hazard anticipated would r e s u l t from allowing s u f f i c i e n t oxygen i n t o the system t o precipitate U02. The UO:, would tend t o s e t t l e out i n quiescent regions of the f u e l c i r cuit,, Conceivably, a c r i t i c a l mass could accumulate, or a slug of held. up U02 could be released and circulate i n t o the core placing it on a f a s t period leading t o a severe temperature overshoot, Measures needed t o prevent these happenings are: great care i n prevention of oxygen access t o the fuel; design of the c i r c u i t so that; U02 cannot s e t t l e out a t any point other than a t a designated one; inclusion i n the c i r c u i t of a t r a p where s e t t l i n g can occur; frequent check on the chemical composition of the f u e l and petroPraphic examinations f o r the U02 phase, If these measures should f a i l , the ultimate consequence would be a rupture of the primary c i r c u i t and a spill of f u e l into the containment space, 0

The volume within a reactor containment structure is usually adjusted t o accommodate the t o t a l gas generated a t a reasonable pressure. Since no gas producing reactions have been discovered i n %he molten s a l t system, the reactor containment structure has no pressure requirements other than those needed f o r t e s t i n g it and the volume i s dictated by convenience. One proposal is t o l i n e the primary system c e l l with two layers of t h i n sheet steel. A t c r i t i c a l places (penetrations) the space between would be used a s ii buffer zone so t h a t an i n e r t gas can be kept i n the primary system cell.

During the ana

apparent t h a t the choice of the Loeffler boiler system was a poor one:, adding substantially t o the c a p i t a l cost and requiring a


- 76 -

h

separate development program. A rough analysis on two d i f f e r e n t s u b s t i t u t e systems indicates savings of about $4,200,000 i n d i r e c t labor and material, or $7,300,000 a f t e r top charges. This charge would r e s u l t i n a saving of 0.31 Awlills/kwhi n power p l a n t investment costs and 0.07 FIillsfiwh in calculated maintenance costs, or a t o t a l saving of 0.38 ??lills/kwh.

INOR-8 prices have been taken from the bids submitted f o r t h e f i r s t commercial production of moderate sized orders of s i n g l e l o t s of material. Low bidder quotations range from $2,65/1b f o r 0.25t1 s h e e t (20,000 l b l o t from Haynes S t a l l i t s ) t o $15/lb f o r x .04S11 wall seamless tubing (from Superior Tube). The average p r i c e ass m e d i s $6 per lb. It may be expected t h a t i f a reasonable market develops f o r t h i s material i t s average p r i c e might drop t o $3/lb. This would e f f e c t a c a p i t a l c o s t reduction of a t l e a s t $3,500,000, equivalent t o 0.21 NillsfKwh.

2

I n the opinion of the p r o j e c t d-irector, t h e b a s i s used f o r evaluating chemical processing costs almost e n t i r e l y obscures t h e value of the f l u i d f u e l r e a c t o r concepts i n achieving low f u e l cycle costs. It does t h i s by assuming Vnat each power s t a t i o n is an i s o l a t e d 333 X N p l a n t , and t h a t t h i s p l a n t must stand alone a s f a r a s reprocessing t h e f u e l is concerned. The approach used a l s o eliminates any chance f o r evaluating differences i n chemical processing c o s t s t h a t may e x i s t between t h e three r e a c t o r concepts. I n t h e case of t h e molten s a l t reactor, t h e choice of a graph-

ite moderated reactor f o r evaluation by the Task Force was o r i g i n a l l y macle on the b a s i s t h a t it required l e s s frequent chemical reprocessing t h a t a homogeneous r e a c t o r without graphite t o achieve a moderate conversion r a t i o , and t h a t therefore the f u e l cycle costs and t o t a l power c o s t would be lower. The i n s e r t i o n of an on s i t e chemical processing p l a n t c a s t s doubt en whether the b e s t molten s a l t p l a n t was a c t u a l l y chosen f o r evaluation, since t h e omission of graphite from the system would eliminate one element of doubtful f e a s i b i l i t y from the nolten s a l t r e a c t o r system. Vith regard. t o t h e necessity f o r an on s i t e chemical processing p l a n t f o r the graphfte moderated reactor, it i s granted t h a t , &esp i t e t h e planned precautions, it i s p o s s i j l e t h a t the f u e l could bec m e contaminated by sone accident. However, the proposed on site p l a n t would provide only for the recovery of %he uranium, throwing away the s a l t c a r r i e r , i2nor.i.m and. lithium-7, and would do t h i s a t a r a t e requiring f i v e years t o reprocess the e n t i r e charge. The saxe operation could be accoxplished more quickly i n a l a r z e off s i t e plant, and a t l e s s cost. The savings would come i n a reduction of the on s i t e c a p i t a l costs and ir: t'ne r e s u l t i n g chemical p l a n t operating costs.

P k t


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c

The ORNL estimate of d i r e c t labor and materials involved i n providing f a c i l i t i e s for canning the f u e l , s t o r i n g it, and providing shipping casks f o r sending it off t o a c e n t r a l processing f a c i l i d y is $1,230,000. This item would replace the $3,600,000 l i s t e d f o r t h e on s i t e recovery process. The chemical laboratory would not need t o be a s extensive a s when on s i t e processing i s required s o t h a t $250,000 is estimated f o r two c e l l s replacing the S?00,000 f o r six c e l l s . The savings i n these two items alone over the proposed on site chemical p l a n t is $4,730,000 c a p i t a l cost a f t e r i n d i r e c t , overhead, and contingency f a c t o r s a r e applied t o direct, costs. I t is believed t h a t there should a l s o be s u b s t a n t i a l savings i n both c a p i t a l and operating expenses f o r the gas f a c i l i t y and waste f a c i l i t i e s since these would be used only i n t e r m i t t e n t l y and need l e s s capacity if off site reprocessing were used. Also 5 - t is not believed t h a t t h e ;Ye and K r b o t t l i n g f a c i l i t y w i l l be a s exp n s i v e a s l i s t e d i n the Chemical Reprocessing section. A reasonable f i g u r e f o r t h e c a p i t a l invxxrtment of t h e remaining on s i t e f a c i l i t i e s of a chemical nature might be $5,000,000, and a reasonable operating c o s t f o r t h e l a r g e l y intermittent operation, $f;OO,OOO a p a r . Thus the on s i t e c a p i t a l and operating c o s t s migh? be $1,200,000/year, or 0.54 MillsRwh instead of the $3,460,@00/year or 1.56 Millsfiwh used f o r power cost estimates.

ti

1 L

With off s i t e processing, the remainder of the costs associated with t h e f u e l cycle a r e a s follows, based on the nine-year reprocessing cycle submitted t o the Task Force. The f u e l cycle c o s t s listed below, are comparable t o the t o t a l of those usually l i s t e d f a r s o l i d f u e l element reactors.

Uranium inventory charges Thorium and fuel s a l t depreciation Burn-up of U-235 and thorium Fuel recovery costs*

$/Year

M i lls/kwh (n e t )

615,000 515,000

0.28

1,436,000 326,000

0.23 0.65

0.15

rn

------.-c.--

* Sinking

fund t o provide: Cans f o r shipment of f u e l $150,000; Cost of shipping 1300 tons 1000 miles round t r i p a t $SOO/ton; Cost of chemical processing and waste storage i n p l a n t des$960,000; Inventory charges during holdcribed i n IDO-a363 up, processing losses, and depreciation of U-235 p r i c e from $17 t o $lS/gm = $1,820,000.

-

_.

!


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The development of a continuous s a l t p u r i f i c a t i o n process t o eliminate f i s s i o n products a s they occur would reduce the f u e l cycle c o s t s by 0.52 Mills/Kwh. From t h i s must be subtracted any additional c a p i t a l and operating costs incurred by the improved .scheme. chemical processing I n the opinion of the p r o j e c t d i r e c t o r , the $11,800,000 chemical processing f a c i l i t y should have a through-put c a p a b i l i t y several times t h a t assumed. The reason f o r t h i s belief is t h a t the half cubic f o o t per day capacity is only two days per week operation of the ORNL v o l a t i l i t y p i l o t plant. If the n e t capacity of the chemical p l a n t were even twice t h a t assumed, a second power p l a n t b u i l t a t t h e same site would avoid the chemical p l a n t c a p i t a l and operating charges, or, if spread over two plants, these charges would be reduced by a f a c t o r of two on each plant.

b.

The Development Program (Project Director's Appraisal)

(1) Some of the important objectives of a development program aimed a t an economic r e a c t o r are: The development of reactor components t h a t a r e r e l i a b l e and easy t o maintain i n s i z e s s u i t a b l e f o r a power reactor. The design and development of r e a c t o r layouts t h a t provide f o r

low c a p i t a l c o s t and inexpensive maintenance. The determination of t h e behavior of f i s s i o n products a s they a r e produced i n a reactor over long periods of time.

The development of the b e s t graphite s t r u c t u r e f o r use a s a moderator i n a molten s a l t reactor, and the determination of t h e degree of penetration of t h e graphite, how t o control it and the e f f e c t s thereof. The development program proposed would involve two stages, centering around the design, development, construction and operation of two reactors. The f i r s t would be an experimental reactor of 30 Thermal MW capacity whose design and construction is estimated t o c o s t $18,000,000. The research and development program associated with t h e experimental reactor i s estimated t o cover $19,000,000 over a f our-year period. Upon the successful completion of the experimental r e a c t o r program, a p r o j e c t f o r t h e design, develapment and construction of a prototype power r e a c t o r would be undertaken. Such a r e a c t o r would probably c o s t $~O,OOO, 000 and a $35, 000, OOO research and development program extending over five years i s proposed t o enable t h e successf u l construction of t h i s reactor.


- 79 A t t h e end of this nine-year program costing a t o t a l of $122,000,000, enough information should be available t o enable the design and construction of a commercial p l a n t and t o estimate its c o s t with reasonable accuracy. The accompanying t a b l e shows a possible problem breakdown of the research and development funds over the next f i v e years.

(2) The p r i n c i p a l develapment requirements f o r breeder r e a c t o r s a r e t h e development of a law-cost continuous chemical reprocessing method and t h e development of a r e l i a b l e graphite core vessel t h a t reduces leakage of neutrons t o a very low value.

i

Of the approaches available f o r chemical reprocessing, the most, i n t e r e s t i n g involves three parts. One is a very f a s t s t r i p p i n g of noble gases from the f u e l as they a r e formed. If they can be stripped on a 30-second cycle, about 34.3% of t h e t o t a l f i s s i o n products can be removed i n t h i s way. The second p a r t u t i l i z e s the fact; t h a t r a r e e a r t h t r i f l u o r i d e s from s o l i d solution with each other, and co-precipitate from s a l t solutions. By adding CeF3, which is not a serious poison, so t h a t it holds a concentration of about 1 mole percent i n the f u e l , and s t r i p p i n g it out by a cold t r a p so t h a t a complete change of CeF3 is made once a month, the r a r e e a r t h f i s s i o n product poisons can be held a t t h e i r 30-day value. This would remove another 26.6 percent of t h e f i s s i o n products. F i n a l l y it is believed t h a t the noble metals such a s Mo and IEu w i l l tend t o p l a t e out of the s a l t . The problem is t h a t of fincling a s u i t a b l e way of making t h i s happen i n a harmless place. This would remuve another 23.3 percent of t h e f i s s i o n products. Law cross section zirconium accounts f o r 80 percent of the f i s s i o n products remaining a f t = these treatments, and there do not appear t o be any high cross section r e s i d u a l elements. The provision of t h i s o r some a l t e r n a t e low c o s t chemical processing scheme is v i t a l t o a breeder r e a c t o r and i f breeding is set up a s a serious goal, a million dollar-a-year program would be proposed a s an adjunct t o the experimental r e a c t o r a s it comes i n t o operation.


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82

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The Development of the Breeding P o t e n t i a l i t y a,

Two-Region Homogeneous Breeder

The conceptual design of a two-region homogeneous reactor p l a n t i s described i n ORNL-2634. It c o n s i s t s of an 8' diameter core surrounded by a 2' thick blanket. The core v e s s e l is 5/16'f thick and made from IMOR-8. The performance of t h i s r e a c t o r when fueled w i t h the s a l t

7 mole $ ThFb: 58/7 mole 8 LiF: 34.7 mole $ BeF2: 0.6 mole $. U233F4 is given by L, G. Alexander i n ORML-2626. It has an i n i t i a l inventory of 1238 kg U-233, b u t the increase i n inventory requirements exceeds t h e production of U-233 f o r t h e first t e n years, s o t h a t it requires 1377 kg of U-233 t o s t a r t the r e a c t o r and carry it through t e n years. A f t e r the f i r s t t e n years, t h e generation of excess U-233 holds i t s own with t h e increase i n inventory f o r a t l e a s t t h e next t e n years. With g r e a t e r elapse of time (more than 100 years) the buildU-235 and U-236 will decrease the conversion r a t i o by 0.07. This r e s u l t s from t h e f a c t t h a t t h i s is an epithermal r e a c t o r (mean neutron energy f o r f i s s i o n = 40 ev), and e t a f o r U-235 is excessively low a t these energies. Two expedients would r a i s e the conversion r a t i o t o 1,000 even i n equilibrium w i t h U-235 and u-236, One of these is a f a s t e r chemical reprocessing cycle than the onceper-year proposed by Alexander, The other is the use of a sandwich construction f o r the core vessel, using either a honeycomb cons t r u c t i o n f o r s t i f f n e s s or a f i l l e r metal of low cross section. This expedient would reduce t h e thickness of high cross s e c t i o n IN OR-^ i n the core vessel. trp of

This r e a c t o r has less inherent d i f f i c u l t i e s of construction than t h e reference design "economict1 reactor, since graphite is not involved. Its c a p i t a l c o s t a t t h e same capacitywould be about t h e same, since the r e a c t o r i t s e l f would be less expensive but there is added t h e c o s t of a blanket c i r c u l a t i o n s y s t e m , The apera t i o n and maintenance would probably be t h e same. The f u e l charges might a l s o be about t h e same, since the savings i n fuel burnup would probably be compensated f o r by t h e c o s t of t h e increased volune of reprocessing (about 2 cubic f e e t per day instead of 0.5 cubic f e e t per day) Thus, t h i s breeder, once supplied w i t h U-233, would probably produce power f o r about t h e same c o s t a s the "economic" reactor.

.

The i n i t i a l supply of U-233 could be manufactured i n the blanket of t h i s r e a c t o r using U-235 f u e l i n t h e core, a s i n case 31, The manufacture of the i n i t i a l inventory page 176 of 0.W-2634. of V-233 would take 18 years of 80g load f a c t o r operation. During

i


-

i

83

-

t h i s period, t h e f u e l cycle costs would be 1.9 Millsfiwh more than those f o r the llecononicll reactor, 3270 kg of U-23s would be burned, representing the diffusion p l a n t output from 675 tons of natural uranium, The equivalent burnup on t h e n a t u r a l uranium involved would be h700 MWD/ton. A f t e r t h i s induction period 1377 kg of U-233 would be available f o r t h i s o r other breeder reactors, t o produce power from thorium. A r e a c t o r of t h i s s i z e and type could be risked a f t e r t h e successful operation of an experimental reactor, "hus design and constmction could s t a r t i n f i v e years, operation i n another four years, and the U-233 charge accumulated i n 18 more years. It could then operate f o r 20 years on thorium without the improvements of b e t t e r processing and laminated core wall and by t h i s time (47 years hence) these improvements can be expected.

b,

One-Region Graphite Moderated Breeder

The reference design rleconomictf reactor, with two modifications, could just barely breed, These modifications a r e (1) increasing the core s i z e t o 20' diameter x 20t high, and (2) development of the continuous s a l t p u r i f i c a t i o n scheme described above. An inventory of 1025 kg of U-233 i s required, "his r e a c t o r i s not capable of producing i t s own i n i t i a l inventory, although i f s t a r t e d with U-235 it would eventually approach self-sustaining operation w i t h only thorium feed. The l a r g e r r e a c t o r would add about $5,000,000 t o the p l a n t c o s t s (over t h e economic reactor). A f t e r U-235 additions become negligible, t h e f u e l charges would be about 0.5 Mills/ICwh less than f o r t h e economic reactor, If the continuous processing p l a n t could be b u i l t f o r $13,000,000 ($1,200,000 more than the present p l a n t ) t h i s breeder o r very high conversion reactor would about break even with the 9xonomic1' r e a c t o r a f t e r conversion of U-233 0perat;ion. It does, however, require development of t h e continuous s a l t p u r i f i c a t i o n scheme. S t a r t h g with U-235, 10 years of operation a t 315 MI? ( e l e c t r i c a l ) would suffice t o reduce the U-235 feed t o less than 5 percent of the f u e l requirements. It would take an investment about 1500 kg of U-235 t o g e t t h i s r e a c t o r over t o the thorium cycle. Construction would be s t a r t e d on t h i s r e a c t o r with the same timing a s for t h e "economic" reactor, o r i n about 10 years if t h e program took t h e orderly steps of an experimental r e a c t o r followed by a prototype reactor.


C.

Two-Region Graphite Moderated Breeder

A t y p i c a l two-region graphite moderated breeder r e a c t o r has a graphite core 4.4' i n diameter by 4.4' long. The core volume i s 85%graphite, 15 percent fuel. The f u e l salt i s t h e same as t h a t f o r the fleconomic'f r e a c t o r , and the blanket salt i s the same, but without uranium. A two-inch graphite w a l l separates t h e core and blanket. The 30" t h i c k blanket i s pure salt containing 13 mole percent ThF4 and no graphite. The volume of f u e l i n the core i s t e n cubic feet and i t is assumed that with 30 cubic feet of external volume, 100 thermal E9W can be dissipated. The core salt i s p u r i f i e d on a once-per-month cycle using the salt p u r i f i c a t i o n process described i n t h e development section. The blanket salt has i t s uranium content removed monthly by the f l u o r i d e v o l a t i l i t y process.

The points of doubt with regard t o t h e technical f e a s i b i l i t y of t h i s r e a c t o r are t h e construction of a graphite core vessel and its connection t o the e x t e r n a l system, t h e achievement o f an e x t e r n a l fuel volume o f 30 cubic feet that delivers 100 M W to a secondary system, and t h e development of the salt p u r i f i c a t i o n system. The graphite core w a l l need not be completely impermeable t o t h e salt, s i n c e salt of the same basic composition is used i n both core a d blanket, Leakage of t r a c e s of uranium t o the blanket would not be h a r m f u l s i n c e t h e uranium i s removed from t h e blanket and returned t o the core once a month. As t o t h e connection of the graphite core t o an external system, E&W have successfully operated such a connection t o a 6" diametar pipe. A s u i t a b l e graphite core vessel has not y e t been designed, however, and t h e development and t e s t i n g of such a s t r u c t u r e t h a t eliminates leakage of neutrons f r o m the blanket could require a million o r more d o l l a r s of development effort

.

A high performance external heat t r a n s f e r system requires extremely close coupling and t h e use of s m a l l diameter heat exchanger tubes. Ekperience i s a v a i l a b l e f o r the development of components f o r such a high performance system. Maintenance on such a closely coupled system would be d i f f i c u l t and might be confined t o the pump.

The neutron balance and inventory of this r e a c t o r a f t e r 20 years of operation i s nearly t h e equilibrium case and i s t h e r e f o r e i n d i c a t i v e of its long term performance. The indicated doubling time is 44 years. An i n i t i a l inventory of U-233 could be accumulated from t h e blanket i n 3-1/3 years of 80% load f a c t o r operation with U-235 i n the core, Such operation would require about 100 kg of

U-235. Information is not a v a i l a b l e at present f o r estimating t h e c o s t of power from t h i s reactor.


- 85 x. A.

LIQUID ~ A FUEL L REACTOR Table of Contents

.

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

Objectives 1. Low Cost Power ,. 2. Breeding The LMFR Approach 1. The Concept m a. Fuel & Moderator of Basic Concept b. Alternative Designs Other P o s s i b i l i t i e s C. Merits & Limitations of the Concept 2. The Development of the Low-Cost Power Potential, a, The Reference Design f o r t h e "Target" Plant (1) Plant Description (a) The Fuel System & Related Container Materials Problems (b) Heactor Plant (c) Primary System (d) Intermediate System (e) Control & Instrumentation (f) Fuel Handling & Processing Sys. ( g ) Special Maint. Facil & Problems (h) Spare Equipment Requirements ( i > Turbine System (j) Electrical System (k) Buildings & Structures (2) Reactor Operating Characteristics (a) Startup, Shutdown, & Load Change

.

B.

.. .

8

8

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

.. .

. . . .

..

Dynamic Characteristics

(c)

Instrumentation Problems

& Design

0

0

3.

The a. b.

110 111 111

(b)

Fuel

C.

91

.. . 1049998 . . 105 107 . .. .. 107 108 .. .. 108 108 . . ... ... 114 114

0

b.

86 86 88 89 89 89 89 89 91 91 91

. . .. .. . . 114 (d) Hazards . . (e) Containment System . .Problems . .. 114 ( f ) Other Hand. 115 115 . ( 3 ) Maintenance . . . . Economic Appraisal . . . . . 118 Development Program (Project Director's Appraisal) . . . . . .. 118 Development of the Breeding Potentiality . 126 Breeder , . . . . . 126 Internally Cooled "Short Doublingtt Breeder LMFR . . . . . . 130 Behavior

"Target"

0

i

Page No.


- 86 -

, 1

X. A.

LIQUID METAL FUEL REACTOR

Objectives

The objective of the LMFR program has always been the development of economical power, For this reason the primary a i m has been t o find a system which would use r e l a t i v e l y cheap, commercially available coolant and container materials and would give good thermal efficiency, high u t i l i z a t i o n of f u e l and t h e p o s s i b i l i t y of integrated f u e l processing. "he development program has been given limited support consistent with t h e f a c t t h a t the concept w a s a very advanced one. Successive evaluations have gone deeply i n t o plant arrangement, a u x i l i a r y systems, remote maintenance, f u e l system s t a b i l i t y , hazards, etc. While the c a p i t a l investment has increased t o some extent as more d e t a i l e d consideration has been given, no unanswerable problems have been uncovered. Therefore, as a r e s u l t of t h e continued favorable findings of the research and development program and t h e continued indication that t h i s concept has application as a l a r g e s c a l e nuclear power plant, the objective of the LMFR program is t o build and t o operate a s m a l l experimental r e a c t o r demonstrating the concept, followed by a prototype experiment t o provide the engineering knowhow f o r construction of a l a r g e s c a l e nuclear power plant on a commercial. basis.

1.

L o w Cost Power

The earliest commercial plant would be of t h e one-region converter as described below. The f u e l would be a s l u r r y of Th02@3%2 in bismuth. A s t h e U-235 burned out and w a s p a r t i a l l y replaced by bred U-233, more concentrated Th02-U23%2 s l u r r y would be added t o maintain c r i t i c a l i t y , "he f i e 1 would be processed on a very slow cycle; v o l a t i l e f i s s i o n products would be removed i n the o f f gas system.

Such a plant of 333,000 KSJe ( g r o s s ) , 825 thermal. MW, could be b u i l t in 1969-1973. The estimated u n i t c o s t s based on gross output would be:


- 87 Mills/Kwh Gross Power Plant Investment ( $ 2 8 6 / ~ w ~ ) Chemical Processing Plant Investment ( 835/KWE) Fuel Inventory, Use Charges and Burnup Chemical Processing: Operation & Maintenance Power Plant Operation & Maintenance

4

*

5.72 -71 * 1.36 075 *

1.46

Total Power Cost

- Gross

10.0

Total Power Cost

- Net

10.7

In the opinion of the Project Director, chemical processing c o s t s would be lower than those shown here. * Cost of power i n Mills/Kwh would be expected t o decrease after f u r t h e r development work. For example, more experience on materials and corrosion i n h i b i t o r s i n t h e f u e l may allow operation a t a temperature r i s e across the reactor g r e a t e r than 300OF with a r e s u l t i n g increase i n power. Although past research has been directed at using low cost s t e e l s , more expensive clad materials (e.g. El0 o r T a ) may more than pay f o r themselves i n higher temperature, 4 T and power, i n t h e reduction of cost of power, Operating expereince will permit improvement i n design of systems and components and reduction i n s a f e t y f a c t o r s , which f o r the first plant will be high, based on very limited information, Variolus a r e a s f o r cost reduction could be: elimination of the i n t e r media.te system, elimination of control rods, increase i n steam temperature, reductions i n the cost of graphite, simplification of both heat exchanger and pump designs which a r e conservatively designed at the present time, These and other changes will r e s u l t not only i n c a p i t a l c o s t reductions but as operating experience is gained and development r e s u l t s improve component r e l i a b i l i t y , maintenance c o s t s should be reduced. Increased capacity is r e l a t i v e l y easy t o obtain i n LMF'R by adding heat exchange equipment since the r e a c t o r i t s e l f does not limit power. Increases would be limited only by the capacity desired i n a s i n g l e unit. A s the s i z e g e t s l a r g e r , on s i t e chemical processi n g also becomes more economical. T h i s would lead t o b e t t e r neutron economy and lower f u e l burnup costs.

-----.. *

See LMFR, page 95.


- 88 i

2,

Breeding

Very e a r l y i n the development of the LMFR concept, BNL recognized t h a t the concept had p o t e n t i a l f o r breeding. The l a r g e s t e f f o r t at BEUL ' has been towards the development of a breeder reactor. There a r e three possible designs which show promise of breeding: the single-region s l u r r y f u e l r e a c t o r presented as the low cost r e a c t o r i n paragraph 1, above; the two-region design which u t i l i z e s a solution f u e l i n the core and a s l u r r y i n the blanket region of the reactor; and the i n t e r n a l l y cooled design i n which the f u e l i s slowly c i r c u l a t e d out of the r e a c t o r t o remove f i s s i o n products and a separate coolant i s c i r c u l a t e d through the r e a c t o r t o remove t h e heat f o r t h e generation of useful power. The chief advantage of the two-region r e a c t o r is t h a t breeding may be obtained with a smaller r e a c t o r which permits higher s p e c i f i c power, I f , i n the long run, single-region r e a c t o r s i n t h e range of 1500 MWE become desirable, then breeding gains comparable t o the two-region r e a c t o r s can be obtained.

The chief disadvantage of the two-region r e a c t o r is increased complexity which adds t o the number of technical problems t h a t must be solved. From the breeding point of view, t h e problems are: a. Adequate blanketing of the core. I n a breeder, t h e blanket must adequately surround not only t h e core, but a l s o the i n l e t and e x i t coolant nozzles. Neutron streaming l o s s e s from the nozzles must be reduced t o a minimum. Radial blanketing of the nozzles involves an awkward mechanical problem. b. The reactor graphite i n t e r n a l s must be designed s o as t o keep intermixing of the two f l u i d s within satisfactory limits established.

From the breeding point of view, the chief advantage of IS.IFR is the low cross-section of bismuth. A breeding gain of .05 can be attained. A t present, the extent of f i s s i o n product and xenon holdup within the graphite s t r u c t u r e is still inadequately known and may a f f e c t t h e achievable breeding gain, While a d d i t i o n a l complexity of design i s c e r t a i n l y inherent i n the i n t e r n a l l y cooled LMFR design, only i n t h a t design does the goal of a 10-year doubling time appear possible,

4


- 89 B.

The LMFR Approach

1. The Concept

a.

h e 1 and Moderator of Basic Concept

The basic LMFR concept i s the use of a l i q u i d metal. as t h e c a r r i e r (solvent o r suspending l i q u i d ) f o r f u e l and f e r t i l e material. Bismtzth i s used because of its low neutron c r o s s section, low vapor pressure, comparatively l o w melting point, and s a t i s f a c t o r y uranium s o l u b i l i t y . Magnesium and/or zirconium may be added t o t h e bismuth t o a i d wetting of s l u r r i e s and t o minimize corrosion mass transfer problems. The moderator i s graphite and can be used uncanned i n d i r e c t contact with t h e fuel. I n t h e primary system t h e f u e l is heated from 750 t o 1050째F within the r e a c t o r ; i t is pumped f r o m t h e r e a c t o r through a heat exchanger where i t t r a n s f e r s t h e heat t o sodium and back t o t h e reactor.

b.

Alternative Designs and Other P o s s i b i l i t i e s

There are many arrangements by which l i q u i d metal f u e l s can be used t o produce power. The two designs under consideration here are:

(1) The one-region s l u r r y fueled reactor designed f o r minimum power costs, and burning U-235 and thorium with a conversion r a t i o of about 0.7. (2)

The two-region solution core, s l u r r y blanket power breeder operating on the Th-U233 cycle,

Other i n t e r e s t i n g p o s s i b i l i t i e s include: a burner f o r recycled plutonium with some conversion of n a t u r a l (or depleted) uranium, and an i n t e r n a l l y cooled breeder with lowered inventory and a s h o r t e r doubling time , C.

Merits and Limitations of the Concept

The special merits of LMFR, i n addition t o those common t o t h e three? f l u i d r e a c t o r s under consideration, are: High temperature and high thermal e f f i c i e n c y but without high pressure; Absence of s t o r e d energy (chemical o r mechanical) which might s c a t t e r f i s s i o n products; No gas production o t h e r than f i s s i o n gases;

I


-90-

(4)

Minimum thermal f l u i n components external t o t h e reactor; Minimum c r i t i c a l i t y problems after a s p i l l ; F l e x i b i l i t y with respect t o neutron energy; The melting point of bismuth i s 525OF, which is s u f f i c i e n t l y low t o present no insoluble engineeri n g problem while r e t a i n i n g f i s s i o n products when i t is i n t h e s o l i d state a t room temperature;

(8) The use of bismuth makes i t possible t o use low cost materials such as unclad graphite and low chrome s t e e l , with i t s long use experience;

( 9 ) Under conditions where corrosion and mass t r a n s f e r o f the container material are acceptable, long

term operation without chemical processing might be possible, U4FR systems have the following l i m i t a t i o n s , i n addition t o those common t o the f l u i d f u e l concepts under consideration :

(1) The heat capacity of t h e f l u i d is lower and t h e density i s greater, r e s u l t i n g i n l a r g e r heat exchangers and higher pumping power requirement; ( 2 ) There has been no demonstrated s o l u t i o n of t h e corrosion-mass t r a n s f e r problem f o r t h e otherwise best choice of container material a t A T = 300OF.

(3) Since t h e f u e l charge i s s o l i d a t room temperature, equipment i s required f o r preheating t h e system before charging. Because i t i s d i f f i c u l t t o remove fissionable material, poison rods a r e probably necessary. The f u e l i t s e l f has no moderating properties and moderator must be supplied, Since t h e s o l u b i l i t y of thorium is low, breeding can be achieved only through t h e use of a slurry. Polonium i s formed along with f i s s i o n products.

e


- 91 2.

The Development of the Low-Cost Power P o t e n t i a l i t i e s a.

The Reference Design f o r the "Targettt Plant

-

(1) Plant Description S t a t u s of technology, problems t o be solved, t h e i r r e l a t i v e d i f f i c u l t y and prospects of solution, f o r the following: (a)

The Fuel System and Related Container Materials Problems Fuel Cycle

The core c h a r a c t e r i s t i c s of this reactor design are established wholly by economic considerations r e s u l t i n g from an optimization study, The conversion r a t i o i s well below the m a x i m u m a t t a i n a b l e i n t h i s type of system and i s an economic compromise of f u e l burnup cost against vessel and inventory charges. The s i z e of t h e core i s a recoxxilation o f neutron leakage VS. vessel and inventory charges, The U-235-thorium-U-233 cycle has been chosen f o r t h i s design. The :reactor is i n i t i a l l y charged with 770 kg of U-235 and 23,000 kg of thorium as urania-thoria p a r t i c l e s . This corresponds t o 30 g. of thorium per kg of bismuth o r approximately a 3 w/o slurry. The feed mate:rial is assumed t o be U-235. The U-233 bred from t h e thorium i s u t i l i z e d i n t h e reactor, k U-233 builds up in t h e f u e l , U-235 required f o r c r i t i c a l i t y decreases t o a m i n i m u m ; then U-235 requirementis f o r maintaining c r i t i c a l i t y increase as f i s s i o n product poisons continue t o build up u n t i l the f u e l is removed from the system f o r decontamination , Table X-1 Fuel Inventories i n LMFR, kg

final -

Average

0

505 656 28

382 542 28

770

1189

952

Initial Mass of U-235 i n System Mass of U-233 i n System Mass of Pa-233 i n System Total

770 0

In t h e o r i g i n a l reference design as proposed by t h e MFR ' project m t h e r e a c t o r f o r chemical decontamination t h e f u e l is not removed from f i s s i o n products u n t i l t h e end of reactor l i f e . Thorium is added t o the system as needed t o maintain a constant thorium concent r a t i o n , and U-235 is added t o compensate f o r f u e l burnup and c r i t i -


- 92 c a l i t y adjustments due t o t h e buildup of U-233 and f i s s i o n products. "he v a r i a t i o n of f u e l inventory with time i s shown i n Figure X-A and i s summarized i n Table X-1. The n e t burnup of f u e l is a function of the conversion r a t i o A mass balance of f u e l added which varies as shown i n Figure X-B. and burned during l i f e is given i n Table X-2, Table X-2 Fuel Feed and h r n u p i n 825 MW Single Fluid LMFR Mass at Startup

Isotope

L

Mass a t

Mass

30 yrs. Burned L K g

505

U-235 U-233 Thorium

0

684

23,000

23,000

4,024 4,894 5,700

Total

23,770

24,189

14,618

770

Mass of Fuel Added Kg

Total Utilization Burnup %

3,759

89 88

5,700

20

38

Chemical Processing

In this f u e l cycle no chemical processing is performed on the f u e l u n t i l 30 years of plant operation have been completed, The i n i t i a l charge c o n s i s t s of f u l l y enriched uranium and thorium contained i n single s o l i d phase p a r t i c l e s as oxides dispersed i n l i q u i d bismuth. The s o l i d s concentration i n the s l u r r y suspension i s approximately 3 w/o, U r a n i u m and thorium are added a t frequent i n t e r v a l s t o maintain t h e proper concentrations f o r c r i t i c a l i t y , nuclear s t a b i l i t y , and economy. After 30 years of operation without processing, the primary system contains Z3,OOO kg of thorium, 505 kg of U-235, 684 kg of U-233, approximately 950 kg of other uranium isotopes and approximately 5,500 kg of f i s s i o n products plus corrosion products. After plant shutdown, the f u e l remains i n t h e p l a n t ' s dump tanks f o r radioactive cooling f o r approximately 200 days, The f u e l is then transferred in s m a l l batches--still i n l i q u i d slurry--to t h e on s i t e oxide slagging plant. Here t h e soluble f i s s i o n products i n the l i q u i d phase are oxidized and, together with t h e uranium and thorium s o l i d s , are separated from all o r most of the bismuth l i q u i d phase, placed i n shielded and cooled shipping casks f o r shipment t o a c e n t r a l thorextype aqueous processing plant f o r recovery of t h e fissionable isotopes, While the bismuth i s not completely free from r a d i o a c t i v i t y , i t should be s u f f i c i e n t l y pure f o r recycling t o another LMFR system.

t

t


/

I

I I I \ \ \

0

0

co a)

0 0

0

a2

0 0 N

r-

*

0 0 \D

0 0 Ln

\o

0 0

co

3 3

+3

3 3 Q r)

3 3

4'

Q

3

> 3 3

> > 3

>


I

I

I

I

I

I

I

I

f

I

'P I

I

0

I 0

0:

\z,

0

ib

0 0

U>


- 95 i

I n contrast t o the above assumpt ons, i t w a s -2cided on the Task Force t h a t t h e f u e l would have t o be processed a t a rate of 5%per year, and the chemical processing c o s t s were estimated on this basis.

t

The chemical process has not been developed o r designed. (This i s One of the important items on the R&D program.) I n the opinion of the LMFR Project Director a s u i t a b l e process would consist of t h e following steps:

i

1.

Treatment of the Th02-U02-fep.-Bi s l u r r y with NaOH and/or air t o remove wetting agents and f l o a t oxides t o the surface of the bismuth.

2.

14echanical separation of the bulk of the oxide from the bulk of the metal by f l o t a t i o n , skinning, etc. (separation need not be complete)

t

i

1

.

3.

Oxide with some residual bismuth is e i t h e r (a) frozen and put i n on s i t e Thorex process, o r (b) c a s t i n t o cans and s e n t o f f s i t e frozen.

4.

Thorex process ( i f on s i t e one cycle gives s u f f i c i e n t decontamination)

5.

Blending of Th(NO3)4 and U02(K03)2.

.

6. Denitration t o form mixed

oxides.

7. Dispersion i n B i (from s t e p

2) with wetting agents, and

r e t u r n t o Reactor System. The s l u r r y must be processed at a r a t e of 340 lbs. o r 0.5 CU. f t . per (lay t o handle 5%of the charge per year. A l l operations with recycled U and Th have alpha and gamma a c t i v i t y and must be handled behind shielding and i n an a i r t i g h t system.

The chief differences of opinion between the project and the Task Force a r e on the probable d i f f i c u l t y of engineering such a process and its cost.

I n t h e opinion of the Project Director, s t e p s 1, 2, 3, and '7 (above) at t h e required s c a l e could be done i n one laboratory hot c e l l s i z e d 10 f t . x 10 f t . with services. ( A spare c e l l would be desirable )

.

Preliminary cost estimates by experienced persons at Union Carbide Nuclear Company agree with the project estimates t h a t this process can be done f o r much l e s s than the Task Force estimates.

1


- 96 mere a r e some a l t e r n a t e s turn out t o be higher than the i n the case of this Task Force economical t o s t o r e the fuel.

available: I f the reprocessing c o s t s value of t h e recovered materials (as c o s t estimate), then i t would seem This can be done s a f e l y and cheaply

ismuth and thorium would not be recovered unless t h e process cost less than the value recovered. Enriched U-235 and U-233 could not be thrown away but it should be allowable t o s t o r e them u n t i l future process development o r l a r g e r batch sizes made the recovery economical. Such process development does not appear ( t o the p r o j e c t ) t o be very d i f f i c u l t . A t any r a t e there would be an economic incent i v e t o develop it.

4

Present Technology

1

The important areas of research and development center around the development of a successful s l u r r y and t h e compatibility of f u e l and container materials. S l u r r y Development The immediate objective i s the development o f an insoluble uranium-thorium bismuth slurry. Current technology leads t o the following summary :

1. Densities of s o l i d and l i q u i d are nearly i d e n t i c a l before i r r a d i a t i o n . During i r r a d i a t i o n t h e density of t h e s o l i d will probably decrease. 2.

Viscosity of s l u r r y i s calculated t o be e s s e n t i a l l y t h a t of bismuth. No measurements have been made a t operating conditions.

3.

Density separation cannot be estimated u n t i l i n p i l e loop s t u d i e s have been made. Caking due t o l o s s of e l e c t r i c a l charge on p a r t i c l e s cannot occur because the metal cond u c t i v i t y prevents such a charge. The reactor design reduces the objectionable f e a t u r e s of separation as much as appears feasible.

4.

The principal problem i n t h i s s l u r r y is proper wetting.

i M l e the s l u r r y program is i n its e a r l y stages, the following experience has been obtained;

1. Argonne National Laboratory has prepared and c i r c u l a t e d Tho2 s l u r r i e s i n sodium, demonstrating a s l u r r y i n l i q u i d metal.


- 97 2.

Knolls Atomic Power Laboratory has success i l l y c i r c u l a t e d an 8.0 w/o U02 s l u r r y i n bismuth i n a 3/4" loop,

:3.

Brookhaven Mational Laboratory and t h e Babcock & Wilcox Company have prepared and c i r c u l a t e d oxide s l u r r i e s i n bismuth on a small s c a l e , and have determined t h a t ( a > ThO2-UO2 powders can be dispersed i n bismuth with proper wetting agents and a g i t a t i o n , (b) dispersions once obtained are s t a b l e f o r many hours without a g i t a t i o n , and ( c ) development has progressed t o the point where very small pumped loops are being operated. Compatibility of Fuel and Container Flat erials

The basic construction materials f o r containing these s o l u t i o n s are low chrome steels. There is a d e f i n i t e problem r e l a t e d t o corrosion. Metal tends t o be removed a t hot regions, where its s o l u b i l i t y i n bismuth i s r e l a t i v e l y high, and t o deposit at cooler regions, where t h e s o l u b i l i t y is lower. This l e a d s t o a mass t r a n s f e r , e f f e c t i v e l y corroding the h o t t e s t regions and tending t o plug chann e l s at coolest regions. The problem i s magnified by increasing A T between h o t t e s t and coldest p a r t s of the system, and by increasing temperature

.

13y c a r e f u l control of s t e e l composition, and by the addition of zirconium t o the bismuth, i t has been possible t o control t h e corros i o n and mass t r a n s f e r of chrome-molybdenum s t e e l s a t a A T of 1800~. Temperatures were 945" and 765째F. Steel containing l%% C r and %%AMo ( t y p i c a l C = 0.12$, N = 0,010-,015%) has shown about 0.1 m i l corros i v e a t t a c k i n 18 months exposure at flow rates up t o 14 ft/sec, There has been some more rapid a t t a c k of welds, which would have t o be prevented before performance could be described as s a t i s f a c t o r y f o r t h i s material a t t h i s AT. The reference material i s Croloy 2%. I n most respects t h e behavior of this a l l o y i s very much l i k e t h a t of the Croloy 1% mentioned above, Nany t e s t s a t r e l a t i v e l y low A T have shown q u i t e s a t i s f a c t o r y performance. One thermal convection loop has been run successfully f o r about 7 months with a A T of 2 6 0 0 ~ . Kowever, only one loop has been run f o r as long as 11 months a t a A T of 1800~ o r above. This t e s t , a f t e r about 10 months a t 18-OF A T showed l o c a l i z e d a t t a c k ; a f t e r another month of operation the a r e a of a t t a c k had increased. Maximum penetrations were about 15 t o 20 m i l s . It i s c l e a r t h a t t h e corrosion of Croloy 2% remains a major worry, and t h a t considerable work remains t o be done before the problem i s solved.


- 98 Graphite gives every indication of being s t a b l e and unreactive with the l i q u i d metal. Absorption of bismuth occurs i n presently It is available graphite t o the extent of about '/2 gram per C.C. possible t h a t pre-impregnation with pure B i would minimize buildup of f i s s i o n product poisons i n the graphite. This would a l s o prevent (less worrisome) f i s s i o n r e c o i l s there. The effectiveness of this treatment cannot be predicted. I

Erosion by s l u r r i e s might not be a problem due t o the density match of s o l i d and liquid. It has not yet been evaluated experiment a l l y . (b)

Reactor Plant

The r e a c t o r vessel contains a core and r e f l e c t o r assembly having a diameter equal t o the height of 1 4 f e e t . The assembly i s cons t r u c t e d of l a r g e graphite pieces machined t o form a cylinder. The desired f u e l content i n the core i s obtained by d r i l l i n g v e r t i c a l holes or channels having the proper s i z e and spacing i n the graphite matrix. The core i s surrounded by a 1.5 f t . r e f l e c t o r containing only enough f u e l channels t o provide cooling of t h e graphite. The l a r g e graphite pieces a r e held t i g h t l y together with s i x circumfere n t i a l temperature compensated metal bands. The fuel-coolant stream (U02-Th02 s l u r r y i n bismuth) e n t e r s the bottom of t h e r e a c t o r a t 750°F, flows upward through the graphite where c r i t i c a l i t y i s achieved, and e x i t s from the top of the core a t 10W0F. The f u e l e n t e r s i n t o and e x i t s from hemispherical plenums located a t the top and bottom of core. Also a c y l i n d r i c a l space above t h e core allows ₏or expansion of t h e f l u i d and a space f o r the expansion o f the f l u i d and a space f o r the control rod drives.

"he r e a c t o r vessel is constructed of Croloy 2% (low chrome a l l o y s t e e l ) with the main s h e l l course having a thickness of t h r e e inches. The design pressure i s 150 p s i and the design temperature i s 1100'F. The r e a c t o r vessel i s surrounded by a c l o s e - f i t t i n g containment vessel 16 f t . 6 i n . I D and l $ i n . thick. This vessel contains any s p i l l of r a d i o a c t i v i t y out of the r e a c t o r vessel and serves as a cont a i n e r f o r gas used i n preheating the vessel during s t a r t u p and removal of decay heat a f t e r shutdown. Gamma and neutron heating o f the r e a c t o r vessel w a l l does not appear t o be a s e r i o u s problem i n t h i s reactor. The i n s i d e w a l l of the vessel i s cooled by the fuel-coolant solution passing upward through an annulus appoximately 1.5 i n . thick e x t e r n a l t o the ref l e c t o r graphite. Thermal s h i e l d s appear unnecessary. This i s due


- 99 t o the excellent gamma shielding properties of the bismuth in the core and the neutron attenuation i n the l a r g e graphite assembly. Thermal s t r e s s e s i n t h e graphite have been roughly evaluated and do not appear t o present a problem. A possible problem i s the regulation of the coolant flow upward d o n g the f u e l annulus between the graphite r e f l e c t o r and t h e r e a c t o r vessel. A hydraulic mockup is being b u i l t and tested.

(c)

Primary System System Design

The f u e l stream e n t e r s the bottom portion of t h e r e a c t o r vessel at a rdlinimum bulk temperature of 750°F, flows upward through the core where f i s s i o n s within the f u e l cause t h e f l u i d t o undergo an average temperature rise of 300OF; t h e m a x i m u m average o u t l e t tempera t u r e i s 1050OF. Upon leaving the core, the f l u i d passes upward t o a degassing a r e a where v o l a t i l e f i s s i o n products a r e removed from the f u e l stream. The reactor discharge is a header which s p l i t s t h e f u e l flow i n t o three primary heat transport loops. Fach primary system loop c o n s i s t s of two 20-inch pipes between the r e a c t o r and t h e pump, one 28-inch pipe between the pump and heat exchanger, and two 20-inch pipes between the heat exchanger and t h e reactor

.

Ii’rom the degassing area discharge header, each f u e l stream flows t o the suction of a variable speed, c e n t r i f u g a l pump designed t o d e l i v e r approximately 17,500 gpm at 20 f e e t of pumped f l u i d . Each pump requires 1250 bhp. The pump is bayonet mounted through a s t e e l s h i e l d plug which is i n t u r n mounted i n the canyon floor. This plug a l s o serves as t h e mount f o r the intermediate heat exchanger. This arrangement permits semi-direct maintenance of t h e Pump To prevent flooding the upper p a r t s of the pump, t h e l i q u i d l e v e l i n t h e pump b a r r e l i s the same as the l i q u i d l e v e l i n the degassing area a t zero flow. A t f u l l flow it is about four f e e t lower than the degassing a r e a level. Pressures i n t h e degassing a r e a and i n the space above the pump l e v e l a r e equalized a t all times, A check valve, i n t e g r a l l y attached t o each pump discharge, is removable with the pump i n t e r n a l s f o r maintenance. Purpose of these valves i s t o prevent thermal shocking i n t h e r e a c t o r o u t l e t with colder back c i r c u l a t i n g f l u i d from the r e a c t o r bottom i n the event a pump stops.

F


- 100

-

CONTROL ROD DRIVES (4)

h SHIELD1

i

CONTROL ROD

1

OUTLET NOZZLE (6)

CONTAINMENT VESSEL

REACTOR VESSEL

CIRCUMFERENTIAL CLAMPS (6)

GRAPHITE

INLET NOZZLE (6)

Fig. X-C-Single

Fluid Reactor for 315 Mw(e) 3-loop Plant.


- 101 r


- 102 From the pump discharge the f u e l stream flows t o t h e s h e l l s i d e of a bayonet mounted, semi-contact maintainable, intermediate heat exchanger. It i s a counterflow, f l o a t i n g head u n i t containing 3000 3/4" OD tubes swaged t o 5/8" OD at each end. The e f f e c t i v e tube length i s 25.5 feet. The primary f l u i d e n t e r s at the middle of the s h e l l , passes downward around the tube bundle and exits at the bottom of t h e s h e l l , and r e t u r n s t o the bottom o f the reactor. The secondary f l u i d , sodium, e n t e r s a t the top of the exchanger, passes down through a pipe centered on the v e r t i c a l a x i s of the exchanger t o a plenum beneath t h e tube bundle, then up through the tubes and exits through an annular space surrounding t h e i n l e t sodium pipe. Each primary loop i s provided with s i x dump tanks equal t o the loop plus a portion of the reactor volumes. The tanks are s a f e l y sized and are provided with a g i t a t o r s t o prevent s e t t l i n g of t h e slurry. One dump l i n e with a bayonet mounted, semi-contact maintaina b l e valve, connects t h e low point o f each loop with t h e loop dump tanks. The primary loops are f i l l e d from t h e dump tanks by means of bayonet mounted electromagnetic pumps. A number of design problems remain t o be investigated, e,g. cavitation, r o t a t i n g seals, bismuth lubricated bearings.

Many t e s t loops have contained bismuth t o bismuth heat exchangers. A f i v e megawatt (design) heat exchanger has been manufactured and i n s t a l l e d i n the IEUL four-inch loop, A bismuth t o sodium heat exchanger i s not considered t o require a l a r g e development program. The development program being conducted i n this area concerns seals

between the inner and outer s h e l l s t o protect personnel during replacement. Such seals are required f o r all semi-contact maintained equipment

.

Heating and Coolin8 Heating and cooling gas systems are provided f o r reactor, primary system, and dump tanks. Each system c i r c u l a t e s helium through the containment annulus. G a s i s heated i n an e l e c t r i c a l furnace and cooled through heat exchangers t o water. The heaters, coolers, and blowers a r e a l l located i n areas adjacent t o the primary system. These components are arranged f o r remote removal i n case of contamination by a primary system leak, but i t i s expected t h a t most maintenance w i l l be performed d i r e c t l y . The t o t a l heating capacity c f t h e e l e c t r i c a l furnace i s about The heat exchangers are capable of removing about 20 MW t o t a l . Different p a r t s of t h e primary system w i l l be serviced by d i f f e r e n t blowers, etc. but the subdivision has not been specified.

3 MW.


- 103 Containment System

All systems and components which are l i k e l y t o contain f i s s i o n products have close f i t t i n g double containment. "he containment i s all welded, and i s helium leaktight. The containment is f i t t e d with expansion j o i n t s . The space between piping o r components and t h e i r containment i s divided i n t o several regions by bulkheads. A helium atmosphere is maintained i n s i d e the containment. The basic function of the containment system is t o contain leakage from the primary system. Close f i t t i n g containment w a s s e l e c t e d f o r t h i s purpose f o r s t r u c t u r a l reasons, and because i t could be used t o heat and cool the system; t o a i d i n detecting leaks; and t o maintain an i n e r t gas blanket around the primary system, O f f G a s System

'The o f f gas system i s based on continuous r e c i r c u l a t i o n of gases (helium mainly) required t o make up seal leakage, f o r operation of samplers and similar mechanisms, and f o r sweep of t h e reactor. The r e c i r c u l a t i o n time i s long enough t h a t gas entering the primary system c o n s i s t s e s s e n t i a l l y of non-radioactive i n e r t gases plus krypton 85. The design of the system i s based on t h e premise t h a t all gaseous f i s s i o n products with h a l f l i v e s longer than 9.8 seconds e n t e r t h e gas phase. The k i n e t i c s of f i s s i o n gas release from bismuth U02Tho2 s l u r r y are not well known.

The d i s t r i b u t i o n of heat i n the o f f gas system is controllable by the! flow rate of r e c i r c u l a t e d gas used i n t h e plant. The t o t a l f i s s i o n gas production rate i s about 7.0 ft3/day a t operating t e m p eratures. The estimated r e c i r c u l a t i n g gas requirements f o r plant operation is about 6.0 ft3/hour, so that t h i s i s controlling and is used t o estimate t h e d i s t r i b u t i o n of the v o l a t i l e f i s s i o n products i n the! o f f gas system,

The o f f gas system c o n s i s t s o f a gas treatment loop, an operat i o n a l gas subsystem, and a vacuum f a c i l i t y . The gas treatment loop c o n s i s t s of a gas cooler, a gas particu l a t e f i l t e r , a charcoal adsorption bed, a surge tank, a compressor, a holdup and decay tank f a r m , instrumentation, and connecting piping. The operational gas sub-system i s composed of a coarse gas f i l t e r , a surge tank, a vacuum pump, a compressor, and such control i n s t m n e n t a t i o n , and piping as required. Gas required f o r reactor plant operation is drawn from t h e reprocessed gas storage tank, and


- 104 recycled i n t o t h e gas treatment loop. The reactor sweep gas i s drawn d i r e c t l y from t h e reprocessed gas storage tank, and is discharged from the reactor i n t o the gas treatment loop. The vacuum f a c i l i t y includes a gas cooler, a vacuum pump, storage tanks, and a stack f o r waste gas atmospheric disposal. This f a c i l i t y i s required f o r t h e i n i t i a l outgassing of t h e primary system, and t h e evacuation and transport of spent gases from t h e primary system during emergency operation. Additional a u ~ l i a r ysystems are: I n e r t G a s System Shield Cooling System Cell Cooling System Raw Water System Waste Disposal Plant Ventilation Storage Pool (d)

Intermediate System

The intermediate system, which also c o n s i s t s of three separate heat transfer loops, u t i l i z e s sodium as the heat t r a n s f e r medium. All material of construction o f t h e intermediate system, except the steam generator, i s 2% chrome 1 moly steel. The steam generator i s constructed of type 304 s t a i n l e s s s t e e l . lfhe intermediate piping 24 in. schedule 30 is sized f o r a maximum sodium velocity of 17 ft/sec.

-

-

Sodium flowing at 22,100 gpm e n t e r s the tube s i d e of t h e i n t e r mediate heat exchanger (which is t h e bayonet mounted, semi-contact maintenance u n i t previously described) at 680째F, flows downward through t h e c e n t r a l downcomer, e n t e r s the bottom tube header, and flows i n t o the tubes. Sodium flows upward through the tubes countercurrent t o the fbel stream and exits from t h e u n i t s a t 1010OF. From the heat exchanger, the hot sodium flows t o t h e suction of a variable speed centrifugal pump. Each intermediate pump i s designed t o d e l i v e r 22,100 gpm at 180 f t . head. Each intermediate pump i s bayonet mounted through a s h i e l d f l o o r i n a manner t h a t permits semi-contact maintenance. Bayonet mounting of t h e intermediate pumps is provided t o minimize pump maintenance time s i n c e the radiat i o n l e v e l s due t o activated sodium i n the v i c i n i t y of t h e i n t e r mediate pumps could prohibit contact maintenance of the pumps f o r a period of about three days.

-


- 105 From ach pump disch the steam generator, The. "once through" type unit. w i t h . an average length of o v e r a l l length i s 68 f t .

rge, sodium flows t o t h e s h e l l s i d e of steam generator is a U-tube, U-shell, The u n i t s c o n s i s t s of 1066% in. OD tubes 65 f t . The s h e l l OD i s 29 i n . and t h e

Sodium flows counter-current t o superheated steam, boiling water, and feedwater i n t h e steam generator and gives up heat which produces 1,100,000 lb/hr o f superheated steam at 2,250 psig and 100C)OF. From t h e steam generator sodium flows t o the intermediate heat exchanger i n l e t t o complete t h e cycle. I n addition t o t h e components l i s t e d above, a u x i l i a r y components are necessary t o obtain proper f'unction of t h e intermediate system, An expansion tank is located at t h e highest point of each i n t e r mediate loop. This tank serves as a cushion f o r pressure surges, a surge vessel f o r thermal expansion of sodium, and suction head f o r t h e pumps.

The lowest point of each intermediate loop is connected by pipe and dump valves t o a sodium dump tank which receives the inventory of the respective loop. Dip tubes, which extend downward through t h e i n l e t downcomer of each intermediate heat exchanger, are provided f o r drainage of t h e heat exchangers. A plugging i n d i c a t o r and a cold t r a p are provided t o determine s0di.m oxide concentration and t o maintain the oxide concentration a t ].OW l e v e l s

The intermediate system will be heated by induction heaters.

( e > Control and Instrumentation A load change w i l l appear i n the steam system as a change i n t h r o t t l e valve position and, therefore, a change i n steam flow and pressure. The feedwater c o n t r o l l e r s a t t h e i n l e t t o the steam gene r a t o r s w i l l sense these changes and operate t o maintain steam pressure constant. The steam flow might also provide an a n t i c i p a t o r y s i g n a l t o the primary and intermediate system pumps t o change t h e i r speed t o s u i t the load.

The prompt and o v e r a l l temperature c o e f f i c i e n t s of t h e reactor are negative. The present control concept u t i l i z e s the selfregulating properties of the reactor t o compensate f o r small transi e n t s and a regulating rod t o compensate f o r l a r g e t r a n s i e n t s .


- 106 Constant core i n l e t and o u t l e t temperatures are maintained between 2096 and 100% of full power by varying pump speed and adjusting regulating rod t o compensate f o r delayed neutron changes and xenon poison e f f e c t s . Below 20% of f u l l power, a constant average tempera t u r e program regulates the reactor. "he r e a c t o r nuclear instrumentation system provides t h e r e a c t o r period and power s i g n a l s necessary t o actuate, overpower, o r s h o r t p o s i t i v e period s a f e t y c i r c u i t s , control s i g n a l s necessary f o r r e a c t o r power control and the operation d a t a necessary f o r the operation of t h e plant. The instrumentation system will consist of conventional s t a r t u p , intermediate and power channels with the d e t e c t o r s placed peripherally outside the secondary containment i n a r e l a t i v e l y low ambient temperature (200OF) The instrumentation w i l l provide nuclear data f o r s a f e r e a c t o r s t a r t u p when an a r t i f i c i a l neutron source is used. Dual channels w i l l be used t o increase the r e l i a b i l i t y of the equipment and t o provide a double check on the neutron power o r period indications.

.

Experience t o date i n d i c a t e s t h a t bismuth has not introduced any unique problems i n the use of non-nuclear instruments f o r l i q u i d metals. The s t a t u s of technology is:

(1) Tests show t h a t Croloy-% C r - 1 Mo sheathed thermocouples should be s a t i s f a c t o r y f o r LMFR service. (2) Diaphragms i n contact with bismuth are used i n both pressure sensing and l e v e l devices. Diaphragms are currently undergoing t e s t s f o r r e l i a b i l i t y and maintenance.

( 3 ) Level i n d i c a t o r s a r e now on t e s t including d i f f e r e n t i a l pressure, J-probe, and f l o a t types. A l l of the f u e l addition t o the primary system of t h e LMFR will be made by way of a s i n g l e f u e l addition mechanism. This mechanism will consist of a one-hundred gallon a g i t a t e d tank i n t o which t h e primary system f l u i d can flow by gravity and be forced out by helium pressure. The U02-Th02 powder will be introduced i n t o this tank by way of a lock which w i l l serve several purposes.

Fuel sampling i n t h e LWR will consist of a mechanism which obtains a sample of the s l u r r y f u e l from a by-pass loop containing an EM pump and a sampling pot

.


- 107 (f)

Fuel IIandling and Processing Systems

The most economical chemical processing cycle time f o r singlef l u i d LMF'R design i s very long, e.g, twenty t o t h i r t y years of continuous operation before any processing. A t t h e end of this period, t h e e n t i r e r e a c t o r charge i s processed f o r recovery of uranium and bismuth on a batch basis, The chemical separation process c o n s i s t s of two major phases;

-

removal of s l u r r y from the reactor, (1) On s i t e plant separation of fuel and thorium from bismuth, storage f o r radioactive cooling and preparation f o r shipping, i

(2)

-

Off s i t e plant chemical separation of the uranium and thorium i n an o f f s i t e , c e n t r a l processing plant using ' aqueous chemistry (thorex process)

For t h e on s i t e plant, batches of slurry a r e taken from the r e a c t o r and are oxidized by passing a stream of oxygen through the melt o r by slagging with caustic, After oxidation, a liquid-solids separation s t e p i s required. (g)

Special Maintenance F a c i l i t i e s & Problems

Major maintenance f a c i l i t i e s i n the LMFR plant a r e t h e following: Overhead crane, storage pool f o r radioactive components, decontami n a t i o n c e l l f o r mobile equipment, hot c e l l f a c i l i t y f o r r e p a i r and replacement of pump p a r t s and repair of other s m a l l radioactive equi:pment, maintenance t o o l s , i.e, s e a l weld c u t t e r s , welders, remote viewing and portable l i g h t i n g equipment and mobile trucks, The design of the overhead crane is such t h a t equipment subject t o r a d i a t i o n damage and with a high r e p a i r frequency i s placed i n a shielded room, t h e crane is remotely operated from a c e n t r a l control room. The major maintenance t o o l s will include s e a l weld c u t t e r s f o r contact welding on the primary heat exchangers and pumps and remotely operated pipe c u t t e r s and welders. A prototype of a remotely operated welding machine has been b u i l t and has undergone tests w i t h encouraging r e s u l t s , Remote viewing will be accomplished by the use of porti3ble l i g h t i n g equipment, TV u n i t s , and periscopes, The mobile equilpment w i l l c o n s i s t of remote controlled mobile manipulators mounted on remote controlled l i f e trucks, and remote controlled t r a c t o r s , The mobile manipulator is undergoing performance s t u d i e s , The :remote controlled t r a c t o r is commercially available.


(h)

*re

Equ-pment Requirements

Spare equipment f o r the reference plant w a s selected based upon expected f a i l u r e frequency and time required f o r procurement, Spares f o r C l a s s I failures (72-hour shutdown) consist of primary and secondary pump r o t a t i n g assemblies, intermediate heat exchanger tube bundles, dump valve operators and i n t e m a l s , i n s t r u mentation, and other items which are designed s o t h a t they can be rapidly replaced. Spares f o r Class I1 f a i l u r e s (1-3 week shutdown) are limited t o a steam generator, sodium piping, and insulation.

Spares f o r Class I11 f a i l u r e s (extremely long down time) include spare casings and containment f o r primary and intermediate p u p s , IHX casing, dump valve, sampler and f u e l addition casing, and spare p r i m a r y piping and insulation. A miscellaneous spare equipment inventory is included f o r such items as pump impellers, bearings, check valve p l a t e s , i n s u l a t i o n , sodium f i l t e r s and e l e c t r i c a l parts.

(i)

Turbine System

A non-reheat, 2000 p s i , lOOOOF u n i t was selected f o r t h i s application. "he non-reheat f e a t u r e was based on economic s t u d i e s performed f o r BAW-2, which indicated lower c o s t s f o r non-reheat, based on constant n e t e l e c t r i c a l output, No attempt has been made t o re-optimize this system based on today's prices; such a study may i n d i c a t e that a reheat cycle may produce s l i g h t l y cheaper electric power. "he temperature f o r t h e system w a s based on once-through steam generator s t a b i l i t y considerations, A heat balance is shown i n f i g u r e X-E.

About 18,000 KW of e l e c t r i c a l power is used f o r the various pump and a&liary systems i n the plant, making t h e n e t output 315,000 KW, Therefore, t h e n e t heat rate i s 8940 BTU/Kwh which corresponds t o an efficiency of 38.2 percent. (j) E l e c t r i c a l System "he primary power supply i s expected t o be 34 KV/incoming t o a 20,000 KVA three winding transformer, 34 KV 4.16/4.16 KV with protective switchgear. The 4160 v o l t main d i s t r i b u t i o n center will supply e l e c t r i c power d i r e c t l y t o t h e large motors of 200 H P capacity and above, through appropriate switchgear. A 480/277 v o l t

-


I

i i

I

I

I

'ii

7

1

1


- 110 power center will be a double end feed u n i t from the 4160 main d i s t r i b u t i o n center t o furnish power f o r motors l e s s than 200 HI'. To provide a r e l i a b l e source o f e l e c t r i c a l power during emergency conditions, a diesel generator unit w i l l be used as t h e a u x i l i a r y power supply. A constant power supply w i l l be used f o r a l l c r i t i c a l instrumentation and control loads. Ehergency l i g h t i n g , a l a r m bus, PA system and b a t t e r y vents w i l l be powered from the constant power supply where s a f e t y of personnel i s e s s e n t i a l . A feeder from the 4160 V main d i s t r i b u t i o n center w i l l be stepped down t o 277/480 V Wye, t o provide 277 v o l t fluorescent l i g h t i n g i n a l l general areas. The high bay o r reactor canyon w i l l have mercury vapor fixtures. The 480/277 V plant d i s t r i b u t i o n i s used t o (1) reduce f a u l t currents, and (2) reduce conductor sizes. A s other voltages a r e required they will be supplied l o c a l l y by dry type transformers.

-

-

Induction heating w i l l be used on t h e primary and intermediate systems f o r heatup. Present estimates i n d i c a t e t h a t 1625 KW w i l l be required. The turbine generator output w i l l o r i g i n a t e from two p a r a l l e l generators each rated at 196,500 KVA with the voltage r a t i n g of 16,000 volts. The excitation voltage is expected t o be 275 v o l t s with a standby e x c i t e r generator. These generators w i l l be the hydrogen cooled type with t h e coolers within t h e generator housing and designed t o operate a t 95째F. The generator output w i l l be transmitted after a stepup transformer t o provide a voltage of 115 KV through t h e usual protective switchgear. A s i n g l e transmission l i n e w i l l be used t o and from the plant s i t e . From the main gene r a t o r bus, power w i l l be used through a s t a t i o n s e r v i c e transformer t o energize a s t a t i o n a u x i l i a r y bus, making the generator output available within the plant.

(k) Buildings and Structures The nuclear portion of the plant will be housed i n a highdensity concrete s t r u c t u r e with t h e radioactive systems below grade where plant s i t e t e r r a i n permits. The above grade "canyon" will be serviced by a remotely opsrated overhead crane. Wherever possible concrete w i l l serve the double function of s t r u c t u r a l support and biological shielding. The plant w i l l be arranged f o r s i m p l i c i t y of system arrangement and economy of piping and space, u t i l i z i n g a semi-contact maintenance philosophy f o r radioactive systems. The non-nuclear portion of t h e plant w i l l be of standard cons t r u c t i o n , housing the turbine generator and aulciliary systems,


- 111 and the supporting f a c i l i t i e s f o r administration, operation and maintenance of t h e plant. Plant arrangement and c o s t s were developed i n conjunction with a consulting architectural-engineering f i r m .

i

(2) Reactor Operating Characteristics (a) In primary primary will be fluid

.

Startup, Shutdown, and Load Change Behavior

the s t a r t u p of the plant it w i l l be necessary t o preheat the system t o 1000OF. This preheating of the r e a c t o r vessel, pumps and heat exchangers, all primary piping, and dump tanks c a r r i e d out p r i o r t o f i l l i n g the system with molten f u e l

A l i m i t e d preheating system which would consist of e l e c t r i c h e a t e r s will be needed on the secondary sodium system. The operat i o n a l procedure f o r the steam portion of t h e plant would be similar t o a conventional f o s s i l f u e l f i r e d plant. The time required t o bring t h e e n t i r e plant from a cold condition t o one of carrying load on t h e turbo-generator w i l l be l a r g e l y determined by the time required t o preheat the primary system. During normal shutdown the r e a c t o r decqy heat i s removed by generating steam and dumping t o t h e condenser. Once decay heat i s less than t h e system r a d i a t i o n l o s s e s t h e tempera t u r e c o e f f i c i e n t w i l l maintain t h e reactor j u s t c r i t i c a l , with t h e slowly c i r c u l a t i n g system at t h e average operating temperature (900OF)

.

The r e a c t o r speed pumps both proposed because of the system t o

is equipped with a control rod system and variable on the primary and secondary loops. This system i s of the limited knowledge of the t r a n s i e n t response various disturbances.

Power i s varied from 0 t o 20% of f u l l load by operating t h e p u m p s at 20% of f u l l load speed and gradually increasing the d e l t a T across The time required t o m a k e t h i s change is deterthe core t o 300'F. mined by the allowable rate of temperature increase i n the container materials. Once 20% of power is a t t a i n e d , the r e a c t o r may be rapidly r a i s e d t o f u l l power by increasing pump speed and adjusting t h e regulating rod t o compensate f o r loss of delayed neutrons.

i

me1 sampling on t h e would obtain a sample of t h e s l u r r y f u e l from a by-pass loop. The samp:Les would be t r a n s f e r r e d t o a hot c e l l f o r a n a l y s i s of uranium, thorium, and additives. The frequency of sampling w i l l be determined by operating experience. I n t h e control of the inventory o f the l i q u i d metal s l u r r y , p a r t i c u l a r l y when f i l l i n g o r dumping t h e p r i m u y system, it may be necessary t o have t h e dump tanks act as weigh tanks. The use of a combination weigh and dump tank would

t

i

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- 112 -


I

- 113 -

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t


- 114 give an indication t o the operator of the amount of l i q u i d metal i n each tank and therefore h i s system inventory. (b)

Dynamic Characteristics

No s p e c i a l problems are foreseen.

(c )

Instrumentation Problems

No s i g n i f i c a n t nuclear o r r e a c t o r instrumentation problems are anticipated.

(d)

Hazards

An investigation of possible methods of i n j e c t i n g r e a c t i v i t y i n t o this reactor has been reported, All nuclear accidents which can be postulated based upon r e a l i s t i c conditions can be handled by the control system without damage t o t h e power plant o r release of radioactive material t o the environment. me maximum credible accident i n t h i s system i s considered t o be a rupture of t h e primary system s p i l l i n g radioactive f l u i d i n t o the c l o s e - f i t t i n g secondary containment. The only f i s s i o n products which can escape are the v o l a t i l e s and t r a c e amounts of v o l a t i l e halides and metallic f i s s i o n products which leak through the containment i n t o t h e c e l l and up t h e stack,

Dose calculations under t h e most unfavorable weather conditions predict lung doses due t o iodine and lifetime dose due t o Po-210 w e l l below t h a t normally considered as a maximum permissible emergency dose. "he absence of s t o r e d energy (due t o low vapor pressure) t h e absence of any known p o t e n t i a l l y s i g n i f i c a n t exothermic chemical reactions (such as metal-water reactions), and the inherent i n a b i l i t y of t h i s system t o i n j e c t sizeable r e a c t i v i t y additions i n t o the core should make t h i s reactor extremely safe and immune t o major accidents which would provide a s e r i o u s radioactive hazard t o t h e environment. This should allow and encourage t h e construction o f t h i s r e a c t o r near l a r g e centers of population. (e)

Containment System

All of the equipment i n the primary system and some of t h a t i n t h e primary a u x i l i a r y system i s enclosed in a c l o s e - f i t t i n g outer pipe o r other metal b a r r i e r . Helium gas is passed through t h i s annulus t o heat o r cool t h e components during abnormal operating conditions. The containment annulus is segmented and p o s i t i v e seals


- 115 1 i

a r e provided between segments i n an e f f o r t t o minimize t h e spread of r a d i o a c t i v i t y i n the event of a leak i n t o t h e annulus. The design conditions f o r the secondary b a r r i e r a r e 35 psig and 1 : L O O O F .

(f)

Other Fuel Handling and Design Problems

Aside from t h e f u e l addition system and t h e fuel dump system described i n t h i s r e p o r t , no other f u e l handling system is needed throughout plant l i f e , "here a r e no requirements f o r deconcentrat i o n of f u e l , the replacement of the f u e l c a r r i e r , o r the chemical processing of f u e l f o r the decontamination from f i s s i o n products. A t the end of plant l i f e (30 years), the f u e l may be handled f o r fissionable isotope recovery as described elsewhere, With regard t o pumping a s l u r r y , the addition of 1.5% t o 3% by weight s o l i d s t o l i q u i d bismuth should not introduce any c r i t i c a l problems. Addition of s l u r r y p a r t i c l e s can conceivably introduce pump bearing plugging and scoring problems where f i n e clearances a r e required. Such problems may be a l l e v i a t e d by using hydrostatic bearings which do not require f i n e clearances, o r eliminated e n t i r e l y by using overhung s h a f t pump designs which do not require a submerged bearing. " h i s question w i l l be l a r g e l y s e t t l e d p r i o r t o LSG'Rl3-11 operation,

(3) Maintenance I n the reference design, the r e a c t o r , primary heat exchangers, primary pumps, and dump tank system a r e located i n a compartmenta l i z e d canyon. A heavy concrete roof covers these compartments and has removable plugs over t h e d i f f e r e n t pieces of equipment. This roof forms the f l o o r of the primary system access room and is at grade elevation. A remotely operated crane s e r v i c e s this area. i n this design can be c a l l e d e type of main dry semi-contact, t h a t is, using both contact and remote methods f o r t h e primary system. Chemical o r other decontamination methods are incapable of removing s u f f i c i e n t contamination t o allow d i r e c t ce on the primar Major components of the plant are considered t o last the l i f e these would include the r e a c t o r vessel, core graphite, of t h e p l a n t primary piping, and dump tanks. Other components i n t h e plant would have less than plant l i f e and would have t o be maintained. Equipment l i k e the primary pumps and heat exchangers would be operated until they f a i l e d , at which time they would be removed and replaced with

-

,


- 116 a new component. Some contact maintenance w i l l be possible on equipment mounted above t h e biological shielding, this would include t h e e l e c t r i c motors f o r the primary pumps, control rod drives, and valve operators f o r the dump valves.

I n the operation of replacing a f a i l e d primary pump, s e r v i c e connections such as power leads, instrument leads, gas l i n e s , and water cooling l i n e s would be removed manually. A seal weld c u t t e r , which is positioned manually, would then c u t t h e seal welds. A component transfer container, which is equipped with a system of g a t e valve locks, would then be used t o remove t h e pump. The a c t u a l removal of t h e pump from its sump would be done remotely due t o t h e s u b s t a n t i a l gamma a c t i v i t y . The replacement of a primary heat exchanger due t o a tube l e a k will be accomplished by contact means e i t h e r a f t e r waiting f o r t h e indPtced Na-24 a c t i v i t y t o decay o r after draining and flushing. The exchanger seal weld i s c u t by positioning a c u t t e r manually, this i s then followed by t h e use of transfer container t o remove t h e tube bundle. The arguments f o r t h i s method of maintenance hinge on t h e f a c t t h a t the f u e l c a r r i e r is a s o l i d at ambient temperature, therefore, t h e r e i s no danger of s p i l l s and i t is f e l t t h a t a s i d e from gamma a c t i v i t y most f i s s i o n products would be r e t a i n e d i n the s o l i d bismuth. However, t h e r e would be a c t i v i t y from f i s s i o n products on the free surfaces of the bismuth. It i s expected t h a t r e p a i r s will be made on f a i l e d primary pumps i n the hot shop. The LWB study has chosen t o dispose of f a i l e d heat exchanger tube bundles r a t h e r than attempt t o r e p a i r them i n a hot c e l l . T h i s choice might be changed i n t h e f u t u r e as more d e t a i l e d information i s obtained on the investment required f o r tube plugging equipment and hot c e l l f a c i l i t i e s needed f o r repair.

The need of a g a s t i g h t outer container around a piece of equipment complicates t h e problems of maintenance. The g r e a t e s t d i f f i c u l t y will arise when piping i n t h e primary system has t o be c u t t o remove a major component. Although this procedure has been studied i n some d e t a i l and methods have been proposed t o carry out t h i s operation, t h e r e are a l a r g e number o f problems t h a t have y e t t o be resolved before p r a c t i c a l i t y i s established.

All maintenance on t h e dump tanks would be c a r r i e d out remotely. The possible w e a k points i n t h e dump tank system are f a i l u r e of the dump valves, electromagnetic pumps f o r f i l l i n g primary system, and overheating of the dump tanks due t o decay heat from within t h e molten fuel. Alternate methods a r e available i f proposed design i s proven impractical.


- 117 The off-gas system, because of the high a c t i v i t y , would have t o be maintained remotely. Because of the p o s s i b i l i t y of f i s s i o n product gases leaking i n t o the system i n the reference design, t h e equipment f o r the helium preheating and cooling systems, e.g. blowers, e l e c t r i c heaters, and cooler would be located i n a limited access area of the p l a n t and sealed i n individual containment vessels. Maintenance on t h i s system's components would be by d i r e c t contact means when the system i s not i n operation, I f t h e components become too radioactive and cannot be deontaminated, they w i l l be destroyed by remote means and replaced.

The fan d r i v e motor f o r the c e l l v e n t i l a t i n g system which would be mounted above t h e plug, would be maintained d i r e c t l y . Work on t h e air fan and cooling c o i l s would be done i n t h e hot shop a f t e r t h e u n i t is removed from the c e l l . The air v e n t i l a t i o n system f o r t h e hot areas of t h e plant would have t o be located i n a c e l l with means f o r decontamination of remote maintenance of t h e fans, motors, and f i l t e r s . I n the reference design, t h i s system i s outside the shielded p a r t o f t h e plant.

The instrumentation and the methods of maintenance have not been studied i n d e t a i l . Some work has been done on a l e a k detection system f o r the primary system. T h i s detection system i s very important s i n c e all components and piping are enclosed by an outer sealed container. The detection system must serve two functions; show t h a t a l e a k has occurred and give t h e location of the leak. A l l maintenance work i n t h e reactor, primary pump, and heat exchanger compartments below the main biological s h i e l d would be performed completely remotely. T h i s work would include replacing the heat exchanger s h e l l s , pump casings, primary piping, dump tanks and t h e i r piping, and all instrumentation not accessible from t h e access canyon. The maintenance would be performed by mobile truck-mounted manipulators and t e l e v i s i o n viewing equipment. The mobile manipulat o r truck, its manipulators and t e l e v i s i o n equipment will be b a t t e r y powered and radio controlled.

Two remote control trucks are under t e s t t o demonstrate the f e a s i b i l i t y of t h i s concept of maintenance. The basic maintenance concept, using mobile trucks, w i l l require extensive t e s t i n g . I n summary, the major probbems of maintenance have been recognized and engineering work i s underway t o solve them.


- 118 b.

Economic Appraisal

The Investment Cost Estimate f o r LMFR i s given i n Table VII-2. The Power Cost Estimate f o r W

R i s given i n Table V I I - 1 . .

Reductions i n investment and power c o s t s might be achieved i n the future by refinements i n design o r as t h e r e s u l t of technical developments. In p a r t i c u l a r , the following f i e l d s seem ( t o t h e project) t o o f f e r opportunities f o r eventual c o s t reduction:

(1) Increase i n e l e c t r i c a l generating capacity. Sharing on s i t e services including chemical reprocessing and spreadi n g cost over a l a r g e r power capacity. (2) Elimination of intermediate system.

( 3 ) Reduction i n graphite cost. Use of U-233 o r plutonium as fuel. ( I f recycled plutonium i s available a t a reasonable price, i t can be burned i n an LMF'R and thus avoid d i f f i c u l t i e s of r e f a b r i c a t i o n of radioA solution of plutonium and uranium a c t i v e s o l i d elements.) i n bismuth gives reasonable conversion without using a slurry.

(4)

( 5 ) Simplification of plant and component designs. (6) Elimination of control rods. (7) Simplification of chemical process. This is probably the most important item since t h e chief motivat i o n of any f l u i d reactor is the opportunity t o cut f u e l cycle c o s t s far below those of s o l i d f u e l reactors. C.

Development Program (Project Director's Appraisal)

The proposed U f F R development program includes two experimental reactors; an LNFRE-1 (Ca. 5 MW) which i s presently being designed t o 1 1(Ca. 50 MW) which w i l l be obtain basic information, and an m directed toward obtaining prototypic o r advanced information requisi t e f o r the f i r s t l a r g e commercial power plant. ~~

~

P r i o r t o t h e fabrication of UFRE-I, pertinent d a t a r e l a t i n g t o materials of construction and external reactor equipment must be obtained. To obtain this data, a comprehensive research and development program was formulated and begun by the Babcock & Wilcox Company


- 119 -

J

i n January, 1957. This program covers physics, materials testing, instrumentation, component testing, and chemistry research. A t the same Lime the Brookhaven National Laboratory i s continuing its basic research on the LMFR and a l s o i n performing research specif i c a l l y required f o r the f i r s t experiment where this work can be performed b e t t e r a t BNL than B&W. Table X-3 lists the principal events f o r the overall LMFR program and the proposed schedule. Table X-3 Program Schedule Begin LMF1IE-I construction Begin Lt.IFRE-I p r e c r i t i c a l t e s t i n g LMFRE-I s t a r t u p & low power operation Begin LMFm-I f u l l power operation Begin advanced R&D f o r DIF2E&II S t a r t design of LMFRE-I1 Begin fabrication of IJlFRE-11 S t a r t p r e c r i t i c a l t e s t i n g LFlFREII S t a r t design LMFR-I Begin hPll-1 pre-commercial operation Begin LMFR-I commercial operation

January 1960 May 1962 September 1962 February 1963 April 1963 September 1963 September 1964

1968 1969 1974 1975

Table X-4 summarizes the estimated costs f o r the complete development program. A more detailed analysis of the program ma;y be found i n Table X-5. Table X-4 cost summary

LMFR Research and Development Program A.

B.

C.

Basic LMFR-I Program R&D required i n order t o build and operate LM.l?RE-I R&D performed on MFREX * Prototype LMFR-I Program R&D required i n order t o build and operate L?@RF,-II R&D performed on -11 * Basic Research on Advanced U4FRS TOTAL R&D Program

36 6,300,000 13,462,000

11,700,OOO 45,200,000 20,000,000

$96,662, OOO

* Includes conceptual, preliminary, final. designs construction and operation of experiment.

as well as


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- 120 -


- 121 The development program w i l l include basic research i n various fields :

:Lo

I-

Slurry

-

5. Make-up Procedures Object of t h i s work is t o determine best procedure f o r preparing U02-Th02 dispersions i n EX, best type and condition of U02-Th02, and best composition of B i phase

-

It w i l l be necessary t o deterb. Chemical S t a b i l i t y mine the eFtent t o which Th i n U02-Th02-Bi dispersions undergoes chemical reduction, t h e objective being t o find a dispersion t h a t i s f o r a l l p r a c t i c a l purposes chemically stable.

-

C. Physical S t a b i l i t y Work w i l l have t o be c a r r i e d out with tge objective of producing a U02-!BO2-Bi s l u r r y which ~3.3.1 have a s a t i s f a c t o r y low rate of phase separation, which w i l l maintain i t s average p a r t i c l e s i z e , and which will not p r e c i p i t a t e U02-Tka02 i n flowing sys terns

.

-

d. Radiation S t a b i l i . t y It will be necessary t o develop a s l u r r y wgch is s a t i s f a c t o r i l y s t a b l e i n a fissioning environment. The influence of f i s s i o n products will have t o be determined.

-

e. Physical Properties Viscosity measurements w i l l have t o be-made; also, t h e tendency t o wet and adhere t o graphite and s t e e l will have t o be determined.

-

f , Sampling Procedures It w i l l be necessary t o do f u r t h e r woFk on the development of s l u r r y sampling~- procedures, p a r t i c u l a r l y on radioactive systems.

-

g. Behavior i n Flowing Systems Both out-of-pile and in-pile loops will have t o be run t o study corrosion, erosion, and p r e c i p i t a t i o n and t o test components, instrumentation, and general handling procedures

.

-

h_. Instrumentation f o r Flowing Systems A certain amount, of research and development work w i l l be required i n t h i s area.

-

i. Small Components f o r Flowing Systems A certain amount, of zevelopment work w i l l be required i n t h i s area also.

f


- 122 2. 0

Removal of Volatile Fission Products

The purpose of t h i s work i s t o determine t h e behavior of t h e v o l a t i l e f i s s i o n products and t h e i r precursors i n LMFR-type systems and, second, t o e s t a b l i s h t h e best method of t h e i r removal from t h e fuel

.

2.

Chemical S t a b i l i t y of Fuel-Graphite-Steel

System

It will be necessary t o determine t h e extent t o which long-term operation w i l l a f f e c t t h e f u e l and t h e r e a c t i v i t y control. Also, the extents t o which uranium and f i s s i o n products penetrate the graphite must be known.

4.

0

Release of Po and Fission Products from Molten and Frozen Fuel

-

This information i s pertinent t o s a f e t y and maintenance considerations.

2.

Equipment Decontamination Studies

It is necessary t o develop s a t i s f a c t o r y techniques and procedures f o r decontaminating s t e e l equipment which has held LMFR f u e l with f i s s i o n products o r has been used f o r maintenance.

-6 .

Recovery of U from S l u r r y Fuel

A s a t i s f a c t o r y method w i l l have t o be developed t h e uranium i n t h e fuel. A s a t i s f a c t o r y method w i l l developed f o r separating the U02-Th02 p a r t i c l e s f r o m It i s presumed t h a t t h e Thorex process would be used t h e uranium i n t h e oxide.

-B.

f o r recovering have t o be the bismuth. f o r recovering

Materials

With the exception of possible e f f e c t s of radiation, a 2ll Croloy f o r t h e =-I s u i t a b l e f o r 13!j째F A T operation f o r a minimum of two years has been obtained. The graphite f o r the core of t h e Experiment w i l l have been completely t e s t e d i n t h e next s i x months. The e f f e c t of i r r a d i a t i o n on corrosion and t h e properties of t h e container material and graphite will be examined i n t h r e e in-pile loops, two of which a r e undergoing out-of-pile checkout at t h i s time. The r e s u l t s t o date i n d i c a t e t h a t the construction of t h e E1 would not be held up by any material problems.


- 123 During constru t i o n f LMFE-I, materi 1 t e s t s will be continued t o demonstrate the p r a c t i c a l i t y of operating a t a higher temperature r i s e across the reactor. It is planned t o operate the f i r s t experiment at a temperature r i s e of a p p r o ~ m a t e l g225OF a f t e r i n i t i a l operation at a A T of 135OF. This i s expected t o be c a r r i e d out some time i n 1963.

The material program will a l s o conduct the prototype t e s t i n g of t h e container material a t the 300OF 4 T t o be used i n the second LMFR Experiment. The i n i t i a l s t u d i e s a t the A T of 3OOOF w i l l have been s t a r t e d before construction of the LMFRE-I. Where s u i t a b l e , improved grades of graphite w i l l be t e s t e d i n the program outlined above. A s soon as the present screening t e s t s at BNL and B&W j u s t i f y it, 3./4" s i z e out-pile l o o p s followed by i n - p i l e loops will be

operated.

Out-pile loops a r e available f o r immediate operation on

a slurry. It i s expected t h a t by the time construction of the LMFRE-I1 is s t a r t e d i n 1964, material s u i t a b l e f o r 300OF A T and 2 years' operations w i l l be available. Throughout t h i s o v e r a l l program a d d i t i o n a l s t u d i e s will be c a r r i e d out on the fundamental aspect of corrosion i n h i b i t i o n and i r r a d i a t i o n e f f e c t s as r e l a t e d t o an LMFR. This work will probably be performed a t mL.

-C. Reactor Physics Xxponential and c r i t i c a l experiments w i l l be performed throughout the development period on the ranges of material composition and 011 the geometries of i n t e r e s t f o r the r e a c t o r experiment, prototype, and full-scale reactor

.

I?arallel with t h e above s t u d i e s would be a programmatic invest i g a t i o n of t h e t r a n s i e n t equations describing temperatures, rea c t i v i t y , and component behavior of the r e a c t o r and heat extraction system. Since t h e behavior o f the r e a c t o r experiment and prototype will no doubt suggest some changes i n t h e c a l c u l a t i o n a l scheme, and since one can expect t h a t t h e analysis w i l l be i n a constant s t a t e of improvement, the a n a l y t i c a l i n v e s t i g a t i o n of t r a n s i e n t behavior can bet expected t o continue a t a reasonable steady pace up through and after the f i n a l construction of a breeder o r f u l l - s c a l e power reactor.

Design Studies - Reactor -

D.

Overall r e a c t o r and p l a n t design s t u d i e s w i l l continue t o be made as t h e LMFR development program proceeds.


-

E. I

Components

-1.

Pumps

Many l i q u i d metals have been successfully pumped i n the tempera t u r e and head range required f o r the LMJ?R, There i s l i t t l e d i f f e r ence between bismuth and lead and other heavy metals which have been pumped successfully. The development program is therefore aimed at modification of e x i s t i n g pumps f o r LMFR service and at extrapolation of pump s i z e s t o those required f o r a l a r g e commercial power plant. The most important f a c t o r s t o be considered are (1) t h e semicontact maintenance requirements, ( 2 ) t h e physical location of the pump at the highest point i n the c i r c u l a t i n g system, ( 3 ) the avoidance of cavitation, and (4) any modifications required i n order t o be able t o pump a 1% t o 3 w/o slurry. A large number of small pumps (1.5 gpm) have been used a t BNL and B&W f o r pumping uranium-bismuth i n out-of-pile loops. These

pumgs have operated very successfully i n forced c i r c u l a t i o n loops; but, no attempt has been made t o modify t h e i r design f o r use i n an experimental reactor. Two large-scale pumps (360 gpm) will begin demonstration runs i n t h e very near f u t u r e on the 4" U t i l i t y Test Loop at E3". These pumps have not been designed f o r contact maintenance and therefore, i t is expected t h a t two 720 gpm pumps designed f o r contact maintenance w i l l be operated p r i o r t o the LME3-1. On present pump designs, s h a f t seals are located j u s t below the pump motor i n a gas atmosphere and t h e i r purpose is t o limit f i s s i o n gas leakage i n t o t h e m o t o r area. Various types o f seals are being considered, primarily labyrinth, f l u i d , c e n t r i f u g a l and wet mechani c a l . Similar seals have been used successfully f o r sodium and salt pumps. Operation o f the LMFRE-I should confirm o r i n d i c a t e modifications t o seal design, Cavitation is an important design a r e a i n LlviFR systems because of bismuth density and plant arrangements which place pumps at high elevations t o permit semi-contact maintenance. It is regarded, however, simply as a design parameter which must be determined f o r bismuth systems. Present pumps are conservatively designed from t h i s point of view and can be designed even more conservatively, (greater suction head o r lower rpm) i f experimental d a t a require such action. The c h a r a c t e r i s t i c s determined f o r s o l u t i o n f u e l s are expected t o apply t o slurry f l u i d s . Required d a t a w i l l be on hand i n Py-1960 p r i o r t o operation of the LMFREP.


- 125 -2.

Intermediate Heat Exchanaers

The use of bismuth does not appear t o introduce any s i g n i f i c a n t heat exchanger design problems. Experience has been obtained with l i q u i d metal heat exchangers of many d i f f e r e n t designs. Some of them, e s p e c i a l l y those used f o r the b r i c o Fermi plant and the EBR-11, are s i m i l a r t o t h e design configuration being proposed f o r the IXFR. The operating experience of these two p l a n t s will be available t o confirm t h e LMFR design approach. Operation of the LEYIFRZ-I exchanger will a l s o provide useful information and i n d i c a t e t h e d i r e c t i o n of devel.opment on d e t a i l items. The properties of s l u r r i e s are not expected t o have a major e f f e c t on heat exchanger design. Use of a s l u r r y will require s p e c i a l a t t e n t i o n t o eliminate crevices and areas of low flow velocity. An engineering flow t e s t s l u r r y loop w i l l be required at an e a r l y d a t e t o provide design information i n t h i s area f o r both pumps and heat exchoangers

.

2.

Valves

There a r e three applications f o r valves i n t h e LMFR primary system. Each pump has a check valve t o l i m i t back flow after pump shutdown. There are three dump valves, and t h r e e smaller f i l l valves b a s i c a l l y s i m i l a r t o t h e dump valves. The influence of s l u r r y properties on dump and f i l l valve design should be s m a l l . They are designed f o r l i g h t service, i.e. each valve! should be used only about 35 times during the l i f e of i t s i n t e x x a b (5 years).

-F.

tIaintenance

I n addition t o incorporating maintenance requirements i n t h e d e s i p of t h e components, development of maintenance t o o l s such as remote c u t t e r s , welders and mobile manipulators w i l l be required. S u f f i c i e n t work should be performed at an e a r l y date t o prove, i n general, t h e f e a s i b i l i t y of t h e proposed maintenance method.


- 126 3.

The Development of the Breeding P o t e n t i a l i t y

a.

The "Target" Breeder

Previous s t u d i e s have resulted i n an externally cooled tworegion breeder r e a c t o r design.' The core i s approximately f i v e feet i n diameter and contains uranium-bismuth solution as t h e fuel. It is surrounded by a blanket region containing 10 w/o thorium suspended i n l i q u i d bismuth (as compared t o 3 w/o Th i n the one-region design). I n both regions, t h e moderator material is graphite. (Figure X-C shows a cross-section of the r e a c t o r designed f o r t h i s concept, Continuous chemical processing of t h e core f u e l solution maint a i n s a steady state f i s s i o n product poisoning value consistent with neutron economy. For example, a core chemical processing cycle time of 76 days will provide a poison r a t i o of .O3 f o r f i s s i o n produ c t s other than xenon and samarium, The processing cycle f o r the blanket material is determined by t h e power generation and U-233 concentration desired i n the blanket region, For example, i f 10% of the t o t a l power i s generated i n t h e blanket system, t h e U-233 concent r a t i o n i n t h e bismuth is 230 ppm and t h e inventory is 48 kg. To obtain rapid reprocessing of t h e blanket, a "soluble'* s l u r r y of M i 2 i n Bi o f f e r s the advantage of quick s o l u t i o n and reconstitution. The chemical processing of both core and blanket must be done rapidly and often and thus w i l l probably require on s i t e processing, The breeding r a t i o expected i n t h i s design i s approximately 1,026 and t h e t o t a l f u e l inventory i n both core and blanket systems is 417 kg of U-233.

This design provides s u f f i c i e n t neutron economy t o meet t h e q u a l i f i c a t i o n s of a hold-your-own breeder. The c h a r a c t e r i s t i c s of t h i s reactor are shown i n Table X-6 and the f u e l c o s t s are summarized The breeding r a t i o can be increased t o 1.05 a t the i n Table X-7. expense of increased f u e l inventory,


4

- 127 t

TABU X-6 SPECIFICATIONS FOR EQUILIBRIUM OPERATION

-

Core : 742 MW 283,000 KW

Thermal Power E l e c t r i c Power Diameter, Inches Height, Inches Fuel vBi/vc

61 91.5 U-233 1.22

N23fiEi Mass of U-233 i n system, kg Total volume of f u e l , f t 3 Breeding r a t i o , o v e r a l l Chemical processing cycle, days Volume flow rate through chemical plant, ft3/day 14ass flow r a t e through chem. plant, g U-233/day Average thermal f l u x i n a c t i v e core Average thermal f l u x i n core system

369

.

1915 1 026

76

25.3 4,850 2.2 x 1015 5.8 x 1013

83 MW

Thermal Power E l e c t r i c Power Thickness, f t . Vslurry/vc

S l u r r y Content: 10% w t Thorium Bismuth 90% w t M23/Mx (atom r a t i o )

620 x 10-6

32 MW

3.5 0.5

- 1190 x

255

48 Mass of U-233 i n system, kg 27 900 Mass of Thorium i n system, kg Total volume of f u e l , f t 3 745 Chemical processing cycle, days 68 Volume flow rate through chem. plant, ft3/day 11.5 Mass flow rate through chem. plant, kg of %/day 410 1.6 1014 Average thermal f l u x i n blanket Average thermal f l u x i n blanket system 5.2 1013

!


- 128 TABLE X-7 Fuel Cost f o r Two-Region Externally Cooled LMFR Breeder B i inventory

440,000 yr.

$

Fuel inventory

353 ,000

Thorium inventory and burnup Chemical processing

145,000 $3,969,000 116,000

TOTAL Less credit f o r U-233

4

83,853,000

Net TOTAL

1

3 031,000

.

1 75 Mills/Kwh

The additional development problems (above those required f o r t h e one-region minimum-cost reactor) would be :

-

Graphite Development A two-fluid breeder requires a " w a l l f 1 t o prevent mixing of t h e core and blanket. Furthermore, t h i s wall must not capture many neutrons. Graphite is a good candidate but more development of graphite properties and fabrication will be needed. The need not be e n t i r e l y leaktight. Some mixing of core and blanket f l u i d s can be t o l e r a t e d but too much mixing is bad since i t puts an e x t r a load on t h e chemical processing. Graphite t o metal seals will be needed and s l i d i n g seals have been t e s t e d and found s a t i s f a c t o r y . Development is being done at present on t h e graphite. (2)

Design problems associated with extending the blanket around core ends will require prototype flow mockups.

( 3 ) It is necessary t o develop a method f o r rapidly processing the blanket fluid. A "soluble" s l u r r y of ThBi2 i n bismuth appears t o have advantages over t h e "insoluble" slurry of Tho2 i n bismuth with respect t o chemical reprocessing. For this reason BNL has been developing a ThBi2 s l u r r y . A s l u r r y of mi2 dispersed i n bismuth containing 0.5% Te and 0.025% Z r has been circulated a t 1.5 f t . per second i n a pumped loop operating i n t h e temperature range o f 400-.50O0C with a bulk d e l t a T of 21OC and film drops of approximately 2OoC f o r periods of up t o 1500 hours without appreciable deposition of 'J!hBi2 o r Fe i n the cooled section.


- 129 No corrosion o r erosion of the s t e e l container o r pump have been detected. Larger d e l t a T, however, might cause p a r t i c l e growth o r deposition i n t h e cooled leg. Therefore the development work should be continued t o increase the permissible temperature r i s e t o t h a t obtained across the core. Operation with a lower temperature r i s e across the blanket would lead t o higher pumping c o s t s and aggravated thermal stress problems.

i

(4) Chemical Reprocessing

-

Whereas only the v o l a t i l e f i s s i o n products would be removed from the f u e l i n the one-region reactors, a l l the f i s s i o n products would be removed i n t h e case of t h e two-region breeder. The breeder would have complete blanket processing.

-

processes f o r removing f i s s i o n Fuel reprocessing products from t h e solution (core) f u e l have already been thoroughly studied a t E3NL.

I n general, the a d d i t i o n a l R&D required f o r the two-region design is t h a t necessary t o develop t h e chemical processing f o r removing the F'P$ * and Fmvl * * f i s s i o n products. Breedi n g d i c t a t e s the removal of f i s s i o n products t o l e v e l s where t h e i r t o t a l reactor poisoning e f f e c t i s no g r e a t e r than about 3%. The individual R&D items required are described b r i e f l y below:

Removal of Fps Fission Products 7

-

Chemistry "he chemistry of t h e fused-salt extraction process i s p:retty w e l l established, although some mopping up remains t o be done. Reaction mechanisms and rates need f u r t h e r study.

-

It i s l i k e l y t h a t the FPS's would Instrumentation and Control be removed continuously; and f o r t h a t , s p e c i a l instrumentation and cont:rol devices a r e needed.

-

-

*

D

-

IPS

-

-

= f i s s i o n products extracted i n chloride s a l t , including a l k a l i s , a l k a l i n e e a r t h s , rare earths.

**

l?BI = noble f i s s i o n products not extracted i n s a l t , including Ru, Rh, Pd, 1210.


- 130 -

Loops, Components, and Zngineering Loop M r e s u l t s w i l l go a long way toward answering these kind of questions, but i t i s r e a l i z e d t h a t f u r t h e r work will be needed, p a r t i c u l a r l y on radioa c t i v e fuel.

-

Equipment Development I n order t o carry out t h e above work a c e r t a i n amount of development work on s p e c i a l equipment w i l l be required. Removal of FPM Fission Products A process has been designed on the b a s i s of small-scale labora t o r y work done a t E3NL and done f o r BNL by the American Smelting and Refining Company. This process i s based upon a combination of s a l t extraction and zinc sludging, Further work w i l l need t o be done, p a r t i c u l a r l y on radioactive handling procedures and engineering operations, The process i s presumed t o be based on batch operation,

-

Blanket Reprocessing I n the case o f the ThBiz-Bi s l u r r y , processes f o r separation and recovery of the bred products, Pa and U, and f o r r e c o n s t i t u t i o n of the slurry have been demonstrated on a laboratory scale. In t h e a l t e r n a t e case of the Th02-Bi blanket, after separation of t h e Tho2 s o l i d s from t h e B i , t h e Tho2 can be processed f o r U recovery and f i s s i o n product separation by the w e l l established Thorex process. But t h e r e is reason t o believe t h a t a cheaper nonaqueous process i s possible. One which looks i n t e r e s t i n g i s a modified fluoride v o l a t i l i t y process. I n this, the Tho2 is dissolved i n N02-HF and t h e U separated as UF6 by t h e addition of F;! o r BrF3. This assumes t h a t t h e non-aqueous process would be s u f f i c i e n t l y cheap, compared t o the Thorex process, t o warrant its complete development,

b,

I n t e r n a l l y Cooled "Short Doubling; Timerr Breeder LMFR

I f thesexpansion of e l e c t r i c power i s faster than can be supplied by "hold-own" breeder reactors, then a design which emphasizes t h e importance of doubling time which i n t u r n places great importance upon high s p e c i f i c power and high breeding gain, will have t o be provided. Design experience on c i r c u l a t i n g f u e l r e a c t o r s i n d i c a t e s t h a t r e l a t i v e l y l a r g e coolant volumes external t o the r e a c t o r core are required t o provide remote o r semi-remote maintenance procedures and a piping l w o u t t o withstand operational s t r e s s e s , This inherent design feature suggests t h a t t h e best method of reducing f u e l invent o r y and improving doubling time i s the use of an i n t e r n a l l y cooled reactor concept.


- 131 Preliminary Reactor Design The core c o n s i s t s of an array of graphite 6-ements made of square graphite s t r i n g e r s , k t h f u e l and coolant holes a r e d r i l l e d a x i a l l y through the s t r i n g e r s and separately headered a t top and bottom of core, Molybdenum f u e l headers a t both ends of t h e core were chosen i n order t o reduce the problems associated with thermal stresses i n t h e graphite t o metal connections. The f u e l c o n s i s t s of U-233 and thorium s l u r r y p a r t i c l e s suspended i n bismuth with a thorium concentration of 50 grams per k i l o gram of bismuth. Almost a l l of the f i s s i o n heat generated i n t h e f u e l stream i s transferred t o the pure bismuth coolant. Fuel i s slowly c i r c u l a t e d outside the reactor vessel i n order t o permit degassing and f u e l addition, The u n i t c e l l i s determined by the c h a r a c t e r i s t i c s of graphite. The m a x i m u m thickness of the graphite between f u e l and coolant channels i s determined by t h e allowable thermal stresses i n graphite. The o v e r a l l diameter of the core (10 ft.1 i s determined by t h e heat t r a n s f e r surface required t o t r a n s f e r 825 MWH.

The core is surrounded on t h e s i d e s by a 2.5 f t . t h i c k blanket containing a 10 w/o thorium i n bismuth s l u r r y , pure molten bismuth as t h e coolant, and graphite as the moderator. The f e r t i l e s l u r r y i s continuously removed from the blanket f o r chemical processing and separation of U-233 f o r feed i n t o t h e core. A graphite r e f l e c t o r , a t least two f e e t thick, a t t o p and bottom of the core i s provided f o r neutron economy.

Preliminary c a l c u l a t i o n s i n d i c a t e t h a t a doubling time of the order o f t e n years is feasible i n this system from a physics standpoint

.

The development of an i n t e r n a l l y cooled LlJIFR will involve many additional research and development programs. These are centered around the design of t h e r e a c t o r i t s e l f , since t h e e x t e r n a l heat removal system can be the same as t h a t presently contemplated f o r e x t e r n a l l y cooled LMFR's. The problems i n t h e r e a c t o r will be the development of graphite s t r i n g e r s of extremely high impermeability and an exceptionally l e a k t i g h t graphite-metal j o i n t which can withstand temperatures i n the range of 1700-1800째F.


- 132 I n addition, highly e f f i c i e n t , low cost, on s i t e chemical processing must be developed. With regard t o t h e time s c a l e on t h i s p a r t of the o v e r a l l

LMFR development program, the whole question of the importance of breeding is sucn t h a t i t should s u f f i c e t o say t h a t no development work i n these s p e c i f i c areas should be s t a r t e d u n t i l t h e r e are more p o s i t i v e i n d i c a t i o n s t h a t a breeder of t h i s type w i l l be desired.


- 133 XI. AQUEOUS HOMOGENEDUS REACTOR PROGRAM Table of Contents Page No. A,

B. C.

D.

Pr.ogram Objectives

. . . . Reactor Types . . . . . . . 1. Two-Region Solution-Core Reactor Plant . 2. Two-Region Slurry-Core Reactor 3. One-Region Slurry Reactor . .. .. S t a t u s of Technology . . . . . Advantages and Disadvantages

1. 2*,

3*, 4,

5.

6,

1

9

4)

..

10

11. 12. E.

. . . . . . . .

3,, 4,, 5. 6,

0

. . .. .

.

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

134

135 136 139

140 140

Two-Region Reactor Core Solution Materials i n Uranyl S u l f a t e Solutions S l u r r y Fuels Reactor Equipment Primary System a. Reactor Vessels b. Heat Exchangers C. Fuel Circulating Pumps d. Pressurizers e. Piping and Valves Low-Pressure Syst e r n Reactor Plant Instrumentation Turbogenerator Plant Reactor Building Reactor Operation and Control Maintenance of Reactor and Other Radioactive Equipment Waste Disposal System

Fuel Processing

l,, 2,

. .. .. ..

. . ...

134

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

Solution Fuel Processing S l u r r y Blanket Processing Slurry Fuel Processing Reconstitution of Fuel Solution Reconstitution of Blanket S l u r r y Preparation of h e 1 Slurry

...

.

.. . .

0

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

F.

Economic Appraisal

G.

Research and Development Program

162 162

163 164 164

165 165 166 169


- 134 -

XI. AQUEOUS HOIQXBEOUS FEACTOR PRGRAM A.

Program Objectives

Development of aqueous homogeneous power breeder r e a c t o r systems which w i l l produce economical nuclear power i n c e n t r a l s t a t i o n i n s t a l l a t i o n s i s the objective of the homogeneous r e a c t o r progrm. Najor emphasis i s placed on development of power breeder r e a c t o r s fueled on the thorium-C23J cycle. The program i s founded on estimates t h a t low f u e l c o s t r e s u l t i n g from simplification of the f u e l cycle by use of f l u i d f u e l s and from complete u t i l i z a t i o n of f i s s i o n a b l e and f e r t i l e material i s an important f a c t o r i n the ultimate economics of nuclear power. I n c i d e n t a l t o the development of power breeder reactors, s t u d i e s have been made of other aqueous homogeneous power r e a c t o r s which u t i l i z e s i m i l a r fuels and equiprnent and have a l l the advantages of the power breeders except t h a t they have lower conversion r a t i o s , and therefore poorer f u e l u t i l i z a t i o n . Included i n this category are U-235 burner r e a c t c r s , plutonium production reactors, r e a c t o r s fueled on the uranium-plutonium cycle and thorium-fueled r e a c t o r s which have conversion r a t i o s less than 1 and require a continuous feed of U-235 t o make up the d e f i c i t . Depending on the p a r t i c u l a r r e a c t o r i t may be r e l a t i v e l y simple t o build and operate; it may o f f e r promise of producing high-priced weapons materials a t low cost; or the f u e l c o s t may be low i n a period of p l e n t i f u l and cheap f i s s i o n a b l e material. A continuing i n t e r e s t i n a l t e r n a t e s i s warranted where they are l o g i c a l s t e p s i n the development of l a r g e power breeder r e a c t o r s and where, as the development progresses, they o f f e r promise of producing power more economically t h a n i t can be produced i n other types of power reactors. However, breeding w i t h complete u t i l i z a t i o n of thorium i n l a r g e power s t a t i o n s remains the primary goal of the work,

B.

Advantages and Disadvantages

The major advantages of aqueous homogeneous r e a c t o r s as compared with other f l u i d f u e l r e a c t o r s result from the favorable characteri s t i c s of heavy water as a moderator and f u e l c a r r i e r , Use of heavy water as a moderator and uranyl s u l f a t e , uranium oxide, and thorium oxide as f u e l compounds makes possible design of c i r c u l a t i n g f u e l r e a c t o r s which have the lowest l o s s of neutrons by leakage and paras i t i c absorption and, therefore the highest conversion r a t i o and


- 135 breeding gain. These same designs have s p e c i f i c powers equal to o r g r e a t e r than those of other f l u i d f u e l r e a c t o r s with resulting shorter doubling time i n breeder reactors. The fuel "concentration i n aqueous system can be a l t e m d by simple physical methods. This characteri s t i c f a c i l i t a t o s both c a n t r o l of the r e a c t o r by c o n t r o l of f u e l concentration, separation of f u e l and f i s s i o n products from c a r r i e r f o r processing and washing of the r e a c t o r t o remove f u e l before performi n g mnintenance operations. The moderator c i r c u l a t e s with the f u e l , i s continuously p u r i f i e d from f i s s i o n products and although subject t o racliation damage, t h e damage products, D 2 and 029 can be reconibined by c a t a l y s t s i n the f u e l or i n external systems. No s p e c i a l provisions are required t o prevent the f u e l s from freezing so many of the equipment and fuel handling problems a r e simplified. There is no requirement f o r maintaining an i n e r t atmosphere i n contact with the fuel. Disadvantages of these r e a c t o r s a r e a l s o r e l a t e d t o the characteri s t i c s of water and the f u e l compounds. Fuels which are now i n use have nlajor faults. The uranyl sulphate solution f u e l i s corrosive and ita s t a b i l i t y decreases w i t h increasing temperature. The t h o r i a f u e l l.s erosive, and s e t t l e s unless i t i s a g i t a t e d continuously; no methods a r e presently envisioned which show promise of separating bred uranium from the thorium without dissolving the t h o r i a p a r t i c l e s . Corrosion of materials i n the r e a c t o r core i s increased by radiation. Special precautionary measures must be t a h n i n the design and operat i o n of the r e a c t o r s t o prevent the moderator decomposition products, D2 and 02, *om introducing a serious explosion hazard. Aqueous systenis operate a t r e l a t i v e l y l o w temperature so the thermodynamic e f f i c i e n c y of the power cycle i s low. The vapor pressure of the water i s high a t the operating temperature of the r e a c t o r so the radioa c t i v i t y is more d i f f i c u l t t o contain. Care must be taken to prevent the expensive heavy water frombeing contaminated by light w a t e r .

C.

Reactor Types

llhree types of r e a c t o r s a r e considered i n this evaluation. Two a r e ci.rculating-fuel, two-region, power-breeder r e a c t o r s being developed art the Oak Ridge National Laboratory. The t u r d i s a one-region r e a c t o r similar t o that proposed by thR Pennsylvania Advanced Reactor Project. One of the .t;Ho-region breeders has a core of dilute s o l u t i o n ( 5 t o 10 g / l i t e r ) of uranyl sulphate i n heavy water surrounded by a t h o r i a s l u r r y blanket of 750 to 1000 g Th/liter i n heavy water. The other has a thorium oxide uranium oxide s l u r r y of 200 t o 300 g Th and 15 to 25 U / l i t e r i n heavy water. U-233 i s produced i n the blankets of the bo-region r e a c t o r s and p e d t t e d t o reach a steady state concentration of 1 t o 5 g U/kg Th before processing, the concentrat i o n depending on the c o s t of processing, the value of a-233 and the importance of a minimum do&ling time. The one-region r e a c t o r is e s s e n t i a l l y the bo-region slurry-core r e a c t o r without blanket.

-


- 136 -

Fuel f o r the reactor would be a thorium oxide uranium oxide slurry containing about 300 g Th/liter and 25 g U/l.iter i n heavy water, A uranium o d d e plutonium oxide slurry i s an important alternative

-

fuel f o r the one-region reactor. The solution-core reactor is the preferred two-region breeder because i t offers the b e s t p o s s i b i l i t y of obtaining the combination of high specific power and high conversion r a t i o that are required t o achieve a doubling time of 10 to 15 y e a r s . Problems of f u e l s t a b i l i t y and corrosion and cooling of the reactor core tank m y so Um3.t the power that can be obtained from a single core a s t o make the solution-fuel reactors unattractive f o r large power reactors, The alternate t o the solutim-core breeder is the s l m q x o r e , tworegion breeder, H e r e the greater d i f f i c u l t y of separating f i s s i o n products from the f u e l and the low specific power required t o limit t b losses by absorption of neutrons i n Pa-233 result i n doubling times g m a t e r than 25 years. The optinnun design based on econollELcs has a conversion r a t i o only slightly greater than 1. Although the conversion ratios may be lower and the doubling times very long, the slwry-core, two-region breeder may have fewer l i m i t a t i o n s on s i z e and power,

It i s possible t o obtain a conversion r a t i o greater than 1 i n a one-region =actor by making the reactor very large, However, the one-region reactor i s most interesting economically and technologi c a l l y as a step i n the development of the two-region, slurry-core breeder and as a l o w fuel cost power producer i n which the d e f i c i t i n conversion i s made up by low c o s t U-23s i n t h i s era of p l e n t i f u l fissionable m a t e r i a l . The 333-WE (net) s t a t i o n t h a t was chosen a s a b a s i s f a r comparison of the f l u i d fuel reactors is described here as a plant i n which the heat i s produced i n solution-core, two-region breeder reactors. Changes which would result from the substitution of slurrycore, two-region and one- region reactors are described. The status of the technology f o r each s y s t e m i s presented and the important problems which r e a n to be solved are outlined. Finally, the costs of power are discussed and programs are outlined f o r the development of each s y s t e m .

1. TwoRegion Solution-Core Reactor Plant A 333-3m plant based on two-region solution-core power breeders would contain three reactors, each operated a t a power of 380

NW

thermal. Three reactors are used because the 2 g / U t e r concentration of fissionable mterial required i n the core t o breed effectively results i n a small c r i t i c a l v o l w and the power density i n the core


- 137 is l i m i t e d by heat removal and corrosion considerations. The three react,ors are operated separately b u t share buildings, control rooms, w a s t e and v e n t i l a t i o n , maintenance and many other f a c i l i t i e s . Steam from a l l three reactors i s piped t o one 333-W turbogenerator. Each r e a c t o r vessel c o n s i s t s of a 4-ft-diameter by 12-ft-long c ~ l i r i d r i , c a lcore operated a t 320 W T (75 KbJ/liter avg, 50 KW/l adjacent t o core tank wall) surrounded by a 2-ft- t h i c k blanket which i s operated a t 60 MW (2.7 Khl/Eter). Under steady s t a t e conditions a 020 solut'on 0.025 m n UO2SO4 ( 5 / l i t e r at the core f l u i d 275OCr 1.9 g U'73, 1.4 g $3Cy Om2 g U q and 1.4 g U 0.025 m i n CrBO4, 0.005 m i n N ~ S O Land ~ 0.025 m i n D2SO4. Copper s u l f a t e is present to reco&ine the r a d i o l y t i c gag, ~ c k e is l a soluble corrosion product and the excess acid i s added t o s t a b i l i z e ttne fuel. The bLaaaE3ftt contains a 750 t o 1000 g Th/liter slurry of thorium oxide In D@ (8.5 to 11.4 volume percent s o l i d s ) . The steady s t a t e concentration of 0-233 and Pa-233 i n the t h o r i a is 2.5 t o 4.5 g/kg Thy depending on the blanket processing cycle. Holybdenum oxide o r palladium catal.ysts, 0.02 m o r 0.002 m respectively, a m incorporated i n the slm~ t o cataIy7,e t h e recoibination of r a d i o l y t i c gas.

-

f

54,

Fuel solution flows through the r e a c t o r core a t a r a t e of 30,000 gpm, entering a t 25OoC ( 4 8 2 9 ) and leaving a t 29OoC (%4?E') Fluid from the core divides into two p a r a l l e l c i r c u i t s where i t passes through the steam generators and is returned to the core by 15,000gpm canned-motor c i r c u l o t i n g pumps. In each c i r c u i t 170 r%T of heat is extracted from the f'uel t o produce saturated steam a t 400 p s i a and 1JLSOF i n the steam generators. D20 is boiled a t 325% (617%) t o 335% ( 6 3 p ~ )i n a pressurizer and surge chamber attached t o the eirciilating system t o control the pressure i n the range 1750 t o 2000 psia.

.

i

Under norma1 operating corditions 100 gprn of f u e l a t Z90째C i s passed through a bydroclone system t o remove suspended f i s s i o n and corrosion pr0duc.t; p r e c i p i t a t e s and discharged through dcounter current heat exchanger i n t o a law-pressure system operates a t 15 to 100 psia. There heavy water is evaporatedffiom the fuel a t a r a t e of 10 gpm carrying w i t h i t the r a d i o l y t i c gas, excess oxygen, fissf-on product gases and iodine t h a t were dissolved i n the fuel. Iodine i s s t r i p p e d from the vapor by p a r t i a l condensation of the D20 and concentzated i n a small reboiler; Xel35 i s removed through the grecursor X135. Deuterium is recombined with p a r t of the oxygen in a c a t a l y t i c recombiner and the heavy water i s separated from oxygen and f i s s i o n product gases i n a condenser. Fuel solution is returned direc%ly from the law-pressure system t o *e high-pressure circiilating system. Condensate is returned partly t o the pressuri z e r and p a r t l y Lo the f u e l c i r c u l a t i n g pumps a s a purge f o r the bearings and r o t o r cavity. Oxygen and f i s s i o n product gases are

thdc


9

- 138 -

i

.i passed through decay tanks a f t e r which most of the gas i s recycled t o the high-pressure system, but a p a r t I s discharged from the p l a n t through cold traps, t o remove small amounts of D20, and. charcoal adsorber beds t o delay the gases u n t i l 10-yr Kr85 i s the only remaining radioactive gas. The gases a r e then discharged through a stack o r the noble gases can be separated from the oxygen f o r f u r t h e r r e t e n t i o n o r sale. Common gas handling and disposal f a c i l i t i e s are used by core and blanket systems of a l l three reactors, Although the letdown and law-pxssure systern has Seen discussed above i n connection with the removal of f i s s i o n products and the production of condensate, i t has several other i n t e r m i t t e n t functions. Fuel i s added t o the r e a c t o r through the low-pressure equipment. The f'uel concentration, and by this means the operating te,nperatuxe, of the reactor i s regulated by c o n t r o l of the r e t u r n of f u e l and condensa+& t o the high-pressure system. When t h e r e a c t o r i s shut down, the f u e l i s contained i n the storage tanks of the low-pressure system. The thoria s l i m y which f i l l s the r e a c t o r blanket is c i r c u l a t e d through the r e a c t o r vessel a t a r a t e of 15,003 gpm. It e n t e r s "he v e s s e l a t 24OoC (4649) and leaves a t 25OoC (482%) The flow r a t e through the blanket i s deter.-nined primarily by t h e requirement t o . keep the slurry well mixed. The temperatures are kept 10:s t o help cool the core tank and t o reduce the neutron l o s s e s , Because only 15 percent of the power i s generated i n the blanket the heat can be used e f f i c i e n t l y a t the lower tenperature f o r heating the feed water before i t e n t e r s the steam generators.

.

Slurry from the blanket i s c i r c u l a t e d through the heat exchanger (feed-water heater) and returned t o the r e a c t o r vessel by a 1~,000gpm canned-notor c i r c u l a t i n g PUT. There is no continuous c i r c u l a t i o n of slurry between the high-pressure system and a low-pressure system, The blanke5 i s pressurized by heating slurry i n the p r e s s u r i z e r which i s attached t o the slurry c i r c u l a t i n g system, Condensate i s produced i n the p r e s s u r i z e r t o purge the bearings and r o t o r c a v i t y of the pump. The blanket pressurizer i s intercomected with tk. core p r e s s u r i z e r +&ough pressure regulating valves t o limit the difference i n pressure across the core tank. Although i t i s not i n continuous use, a low-pessure system i s provided f o r the blanket. It contains storage tanks f o r slurry and condensate, evaporators, recombiners, condensers , and feed pumps. Ths equipment i s used primarihy f o r storing s l u r r y and a d j w t i n g concentrations, f o r feeding material t o the r e a c t o r and f o r withdrawing material f o r processing. Gases released from the system a r e processed i n the general p l a n t f a c i l i t y .


- 139 To make the two-region solution-fuel r e a c t o r an e f f i c i e n t breeder, material i s withdrawn from several places f o r processing f o r removal of f i s s i o n products and f i s s i o n a b l e material. Slurry i s removed from the blanket low-pressure system f o r removal of f i s s i o n a b l e material. S o l i d s and some uranyl sulfate solution are removed from the hydroclone system f o r separation of corrosion a n d f i s s i o n products. Condensate i s removed from the iodine unit i n the law-pressure system f o r separation of iodine and some f u e l i s removed from the low-pressure system f o r removal of soluble corroThe processirg s t e p s a r e sion products, manganese and nickel. discussed i n a l a t e r section. 2.

Trso-Region Slurrv-Core Reactor

The upper limit on s i z e o f core which i s c r i t i c a l with a t l e a s t 2 g/:liter i f f i s s i o n a b l e material i s removed by use of a thorilm uranium oxide slurry f u e l . I n p r a c t i c e , i t appears necessary oxide t o hisve a slurry concentration of 100 t o 390 g Th/liter so t h a t f l u c t u a t i o n s i n s l u r r y concentration which can be expected t o occur w i l l have l i t t l e e f f e c t on r e a c t i v i t y . The corresponding c r i t i c a l concentrations of f i s s i o n a b l e material i n the r e a c t o r a r e 5 t o 15 &iter. S p e c i f i c power (KW/.g f i s s i o n a b l e material) proposed f o r sl.clmjr fueled cores i s lower than f o r solutions fueled cores t o r e s k r i c t the l o s s e s of neutrons by absorptions i n Pu233.

-

The 333-WE power p l a n t based on the two-region slurry-core r e a c t o r would contain one r e a c t o r operated a t a power of 114C ]UT and one 333-14d turbogenerator. The r e a c t o r c o n s i s t s of a core 7 f t diameter ky 2 1 f t long operated a t 1060 El (40 Rd/liter) surrounded by a. 2-ft-thick blanket which i s operated a t a power of 80 NW. Under steady s t a t e conditions ;? thoria, heavy water slurry contai g Th and 4.3 g U per liter--&Z,U233,, 4.8 g U234, 0.6 g U2 Y Pand*0째 2.0 i; C23k--is c t r c u l a t e d through the core a t a r a t e - o f 90,000 gpm. It e n t e r s a t 256OC (493%') and leaves a t 300째C (5720F). Radiolytic gas is recombined i n the slurry by a c a t a l y s t such a s palladium.

S l w r y from the r e a c t o r core is c i r c u l a t e d a t a r a t e of iS,OOO gpm through s i x p a r a l l e l steam generator and pumping c i r c x i t s . Steam i s produced i n the steam generators a t 400 p s i a znd USOF a s with the solution-core rea.ctc?r. The r e a c t o r i s pressurized by heating slurry i n a p r e s s u r i z e r attached t o the circula.ting system. Steam i s condensed i n the press7wixer t o provide purge f o r the c i r c u l a t i n g l Xel3.5. pumps. No provision i s made f o r ~ r n o v &of

t slumy and h

-pressure system a r e s i m i l a r t o those f o r the two-region solution-core reactor. Because this blanket i s larger i t appears necessary t o c i r c u l a t e the 1000 g T h / l i t e r s l w r y


- 140 a t a r a t e of 30,000 gpm t o obtain adequate mixing. Two p a r a l l e l pumping c i r c u i t s a r e provided b u t only one contains a h e a t exchanger; the 80 ranJ of blanket h e a t i s used f o r feed-water heating. Core and blanket p r e s s u r i z e r s are interconnected t o prevent the pressure difference across the core tank: from becoming excessive. Although they are not i n continuous use, low-pressure s y s t e m must be provided f o r core and blanket. They c o n s i s t of storage tanks, recombiners and condensers, evaporators and feed pumps. Gas handling equipment is shared by core and blanket systems. Slurry i s charged i n t o the low-pressure systems f o r feed t o the r e c t o r and i t i s removed through the law-pressure systems f o r processing.

3.

Cne-Region Slurry Reactor

The 333 IWE p l a n t based on the one-region slurry r e a c t o r i s nearly the same as that f o r the two-region slurry r e a c t o r except that the r e a c t o r vessel i s a 12-ft-dimeter sphere w i t h a Conical i n l e t and there a r e no blanket c i r c u i t s . The r e a c t o r operates a t 256 t o 3OO0C and the high-pressure c i r c u l a t i n g system and t h e lowpressure systems are those described f o r the core system of the two-region slurry-core r e a c t o r

.

0.

S t a t u s of Technology

Development work both i n and out-of-pile has brought the technology of f u e l s , materials and equipment f o r aqueous homogeneous r e a c t o r s t o the point where r e l a t i v e l y l i t t l e remains t o be done t o determine whether the s o l u t i o n core system is technologically f e a s i b l e f o r large reactors. The technology of slurry f u e l s and equipment i s not as well developed; there is o n l y a small amount of experience in-pile, The number of c r i t i c a l questions f o r both s o l u t i o n an6 slurry s y s t e m s which remain t o be answered on the r e a c t o r experiment s c a l e i s small and the problems a r e well defined. Most of the equipment and mintenance mthods f o r t h a t s c a l e of operation have been developed.

I

For the solution fuel it i s necessary t o explain the behavior of the f u e l i n the HRT arad t o determine from chemic&, corrosion and r a d i a t i o n data, hydrodynamics experiments, and design s t u d i e s whether a p r a c t i c a l two-region r e a c t o r vessel can be b u i l t , Less work has been done w i t h slurry f u e l s s o 2 l a r g e r nuqber of c r i t i c a l questions remain. Hydrodynamics experiments a r e required t o determine the flow requirements of the r e a c t o r vessel. A s a t i s f a c t o r y recombir,ation c a t a l y s t must be demonstrated. It must be s h m n t h a t the slurry materials r e t a i n t h e i r properties during very long c i r c u l a t i o n periods


- 141 and t h a t long i r r a d i a t i o n does not adverse17 a f f e c t the properties. S a t i s f a c t o r y solutions t o the erosion and s e t t l i n g problems of slurries m u s t be demonstrated , Information concerning f u e l s and. materials obtained on the r e a c t o r experiment s c a l e can be applied d i r e c t l y t o prototype and f u l l - s c a l e p l a n t s , A very l a r g e e f f o r t w i l l be required t o develop r e a c t o r and maintenance equipment f o r prototype and f u l l - s c a l e plants.

:I,

Two-Region Reactor Core Solution

The i d e a l f u e l f o r an aqueous homogeneous r e e c t o r is a s o l u t i o n which has a low p a r a s i t i c absorption c r o s s section, i s s t a b l e against p r e c i p i t a t i o n of f u e l constituents by thermal and r a d i a t i o n e f f e c t s , i s only s l i g h t l y corrosive, and from which f i s s i o n products and corrosion products can be separated e a s i l y . Dilute uranyl s u l f a t e solutAons most nearly satisf’y these c r i t e r i a but they f a l l s h o r t in a nurnl3er of respects. Special care must be taken i n the r e a c t o r design t o compensate f o r the deficiencies

.

lJranyl sulfate s o l u t i o n f o r the core of t h e two-region r e a c t o r contains 0.025 m UO2SO4, 0,025 m C S O 4 , 0.025 m D2SOL i n D20. Copper i s present t o rzcombine radiolyxic gas, e x c e s s a c i d improves the s t a b i : l i t y and the solution must contain dissolved oxygen t o prevent the reduction of uranium. The solution i s thermally s t a b l e t o 340°C. Above 3hO0C i t separates i n t o two l i q u i d phases. One phase i s 5 m or g r e a t e r i n concentration of uranium and contains approximately-a proportionate amount of copper and n i c k e l b u t s l i g h t l y less than the stoiclriometric quantity of s u l f a t e . The second phase i s depleted i n metal ions b u t r e t a i n s the acid. Both the concentration and the f r a c t i o n of t o t a l metal i o n s i n the heavy phase increase a s the temperature of the two-phase mixture i s raised. The heavy phase i s thermdly s t a b l e t o temperatures above 500°C. Uranyl sulfate solut i o n s are s t a b l e a g a i n s t r a d i a t i o n e f f e c t s a t l e a s t t o the two-liquidphase temperature, Very l i t t l e i s known about the s t a b i l i t y of the heavy l i q u i d phase in r a d i a t i o n a t high temperatures and high power density. Iluring operation of the r e a c t o r introduction of f i s s i o n and corrosion products alters the composition of t h e f u e l and precipitates form a t the r e a c t o r operating temperatures. P r e c i p i t a t i o n of f i s s i o n and corrosion products makes possible the use of r e l a t i v e l y simplo hydraulic methods f o r separating them from the f u e l so that this type of i n s t a b i l i t y i s of real concern only i f it involves concentration of uranium or s i g n i f i c a n t loss of copper catalyst. Iron, chromium, zirconium and similar elements hydrolyze and prec i p i t a t e as oxides which may contain as much a s 1 t o 4 percent by


- 142 weight of adsorbed uraniun. Barium, strontium, and most of the r a r e e a r t h s u l f a t e s a r e only s l i g h t l y soluble a t the r e a c t o r temperature. Only a few of the products, e. g., cesium, rubidium, n i c k e l and manganese, are very soluble and normally t h e i r concentrations a r e lirrited by processing s o t h a t the s o l u t i o n i s s t a b l e against p r e c i p i t a t i o n of uraniun and copper salts a t 300째C and lower temperatures. If the nickel concentration i n the f u e l i s permitted t o exceed about 0.005 m while the excess acid i s maintained constant by acid additions, a bxsic copper s u l f a t e p r e c i p i t a t e s until the n i c k e l concentration i s 0.017 m and the copper concentration i s 0.018 m; then nickel s u l f a t e p r e z i p i t a t e s . If the f u e l i s permitted t o Eecome depleted in acid t o 0.015 m by copper and nickel, the system passes i n t o a region of tke phase-diagram where a b a s i c uranyl s u l f a t e i s the unstable phase. A l l of the products have retrograde c o e f f i c i e n t s of s o l u b i l i t y s o t h a t i n s t a b i l i t i e s appear a t lower concentrations of n i c k e l a s the temperature i s r a i s e d . Recently i t has been demonstrated t h a t uranium and other soluble constituents of the fuel can be concentrated on heated surfaces by boiling i f the velocity of f l u i d across the surface i s low. The e f f e c t of the f u e l i n s t a b i l i t i e s i s t o place several major requirements on design and operation of the reactor, A l l p a r t s of the core tank i n contact with the f u e l must be kept below 30OoC. Fuel i n contact with the wall must flow a t high velocity. Suspended s o l i d s mst be removed r a p i d l y and continuously from the r e a c t o r core and c i r c u l a t i n g system, The composition of the f u e l must be controlled properly. F a i l u r e t o comply with these requirements can r e s u l t i n damage t o the reactor.

2.

Materials i n Urarrrl S u l f a t e Solutions

Materials of construction f o r the solution core system a r e zirconium a l l o y s f o r t h e r e a c t o r core tank, a u s t e n i t i c stainless steel-primarily type 347--for equipment e x t e r n a l t o the r e a c t o r core and titanium a l l o y s f o r s p e c i a l applications i n this e x t e r n a l equipment. Corrosion of zirconium a l l o y s in-pile has been shown t o increase with increasing temperature and power density--or f i s s i o n rate--in f u e l s o l u t i o n i n contact with the metal. The corrosion rate a t constant temperature and power density decreases with increasing uranium concentration, a c i d i t y and f l u i d velocity, f a c t o r s which tend t o reduce the r e l a t i v e e f f e c t of increased f i s s i o n i n g a t the surface r e s u l t i n g from adsorbed o r deposited uranium. The following a r e estimates of corrosion r a t e s f o r Zircaloy i n the reference f u e l , They are based on the b e s t c o r r e l a t i o n o f data f o r in-pile loop and autoclave t e s t s over a wide range of conditions.


- 143 Temperature OC

Solution Velocity ftlsec

260 280 300 330

-

11 10

250

5

-

Corrosion Rate, mils/yr a t 25 KW & J* 50 rn&*

11 11

15

15

13

5 15

27 21

32

5 15

37

50

27

40

5

53 34

85 58

28

;;i]alculated d r r p u m power density i n solution adjacent t o the core tank wall i n reference reactor. S t a b i l i t y could be improved and corrosion of Zircaloy reduced by a d d i t i o n a l acid i n the f u e l , but the a c i d i t y i s limited t o permit the use of stainless s t e e l as a material of constpuction. Below 150째C s t a i n l e s s steel corrodes a t a r a t e l e s s than 1 mil/year. As the t e - q e r a t u r e i s increased t o 22S0C the corrosion r a t e depends strongly on the concentrations of uranyl s u l f a t e and acid; f o r the refgrence solution i t i s about 10 mils/year. A t temperatures above 250 C, 0.1 t o 1 m i l of corrosion occurs while forming a protective f i l m on the surface and the metal corrodes a t a r a t e of a few hundredths of a m i l per ye<Ur. T b i s s i t u a t i o n holds a t flow v e l o c i t i e s up t o a c r i t i c a l v e l o c i t y above which rapid. corrosion occurs f o r an unlimited time. The c r i t i c a l v e l o c i t y decreases w i t h increasing uranyl sulfate conce8t r a t i o i i and a c i d i t y and increases with increasing te-rature t o 300 C and above.

with the reference solution the minimum c r i t i c a l velocity is 15 t o 20 :Pt/sec i n standard, high turbulence t e s t s a t 25OoC. It is about 55 f t / s e c i n smooth pipe. A velocity of 20 f t / s e c is specified f o r r e a c t o r systems and care i s taken i n design t o eliminate regions of abnormal turbulence. Pretreatment of the piping by exposure t o oxygenated water i s used t o provide an a d d i t i o n a l s a f e t y factor. Experience i n laboratorg and i n e n g i n e e r i q scale tests has shown that a film which 1s formed by pretreatment p r o t e c t s the metal f o r thousands of horns under conditions w h e r e the c r i t i c a l velocity normally would be exceeded. Operation of the reactor under abnormal conditions f o r hundreds of hours should ham L i t t l e e f f e c t on the n o m 1 protective f iln.


Owgen i n the f u e l prevents the reduction of uranyl ion and the subsequent hydrolysis and p r e c i p i t a t i o n of the uranium which releases sulfuric acid and causes the solution t o become very corrosive t o stainless s t e e l . The chloride content of the f u e l must be kept below 5 ppm as a preventive against stress corrosion cracldng. The oxygen concentration is normally kept i n the range 100 500 ppm.

-

Sgeoial parts, such as pump impellers, heat exchangers which operate a t 200 t o 24OoC o r i n which uranyl sulfate i s boiled a t reactor temperatures, special instruments and valve parts, which are outside the r e a c t o r core and a r e subjected continuously t o unusually corrosive conditions are mde of titanium alloys. These materials e x h i b i t very low corrosion r a t e s i n uranyl s u l f a t e solut i o n s of a l l concentrations, show no v e l o c i t y e f f e c t s , and do not crack i n the presence of chlorides. Experience in loop t e s t s both in and out of p i l e and in H9E-1 and HRE-2 i n d i c a t e s t h a t a u s t e n i t i c stainless s t e e l s (AIS1 347 i n p a r t i c u l a r ) are acceptable materials of construction f o r the reference f u e l i f proper precautions are med i n design and operation

of the equipment. Data from in-pile loop and r e a c t o r e x g e r b n t s lead t o the conclusion t h a t corrosion behavior of metzls i n equipment external t o the r e a c t o r core i s e s s e n t i a l l y the same as observed out of p i l e

.

3.

Slurry Fue Is

Suspensions of thorium ofide, containing uranium oxide i n soLid solution, are the thorium-containing f u e l s f o r aqueous homogeneous reactors, No thorium solutions or s o l s have been found which s a t i s f y the requirements of a core o r blanket fuel. Problems encountered i n the use of t h o r i a s l u r r i e s are r e l a t e d t o the flaw properties and maintenance of homogeneity of the suspensions, t o recombination of r a d i o l y t i c gas and t o erosion by the t h o r i a p a r t i c l e s . S l u r r y f u e l s f o r the one-region reactor and f o r the core of t h e two-region r e a c t o r (200 t o 300 g Th/liter) differ from the blanket slurry (lo00 g Th/Uter) t o the extent that the concentration of the s o l i d s a f f e c t s the properties S l u r r i e s which presently appear t o be useful f o r a r e a c t o r are flocculated and behave as Bingham p l a s t i c o r pseudo p l a s t i c materials. The y i e l d stress and the c o e f f i c i e n t of r i g i d i t y both increase with increasing volume f r a c t i o n solid and decreasin& p a r t i c l e size. A g i e l d stress less than 0.1 1b/ft and a c o e f f i c i e n t of r i g i d i t y less than about 4 centipoises have been established as goals i n the slurry development program s o that the flow behavior w i l l approach t h a t of a Newtonian f l u i d and w i l l not seriously a f f e c t the equipment design.

5

t


-45A wide variety of preparations s a t i s f y these requirements i n the 200 t o 300 g/U.ter s l u r r i e s . They are met i n the 1000 g / l i t e r s l u r r i e s by use of p a r t i c l e s of average s i z e greater than about p. Spheres have the optirmun p a r t i c l e shape although properly sized cubes and p l a t e l e t s meet most of the requirements also. Dispersion of the p a r t i c l e s by additives such as sodium s i l i c a t e makes even the m,ost concentrated s l u r r i e s beham a s Newtonian f l u i d s but such s l u r r i e s have not mtained t h e i r properties indefinitely when circul a t e d a t high temperature o r i n radiation o r the slurries have been d i f f i c u l t t o resuspend when s e t t l e d f o r long times.

Hindered g e t t l i n g r a t e s for the promising slurries are 0.2 t o 2 d s e c a t 30 C and increase to 1 t o 3 cm/sec a t high temperature. Slurrles have been prepared which enter the compaction region of s e t t l i n g a t concentrations from 800 t o 3000 g / l i t e r and s e t t l e t o concentrations of 1200 t o 4000 g / l i t e r . Velocities i n the range of 2 t o 4 ft/sec are required t o obtain f u l l y developed turbulent flow i n pipes and r u l l y turbulent flow is required t o maintain the s l u r r i e s i n wtiform suspension. I n order f o r the slurries t o r e t a i n t h e i r properties over long periods of time i t is necessary that the p a r t i c l e s retain t h e i r i n t e g r i t y and t h a t the products which accumulate i n f u e l have an unimportant e f f e c t on the environment. Slurry p a r t i c l e s produced by two methods have been found t o r e t a i n their i n t e g r i t y adequately f o r test periods which vary from a few hundred t o a few thousand hours. One method requires that the thoria be f i r e d t o 160OoC ( o r t o temperaturea a s low as 10sO째C f o r oxides which contain 8% U) t o fuse the crfs1;allite which form the particles. T h i s material has a surface area of 1 m /g or less. The second requires controlled digestion of thorj.um oxalate c r y s t a l s which are subsequently decomposed t o thorium oxidq and f i r e d a t 80OoC. This inaterial has a surface area of about 15 m'-/g.

h

The e f f e c t of adsorption of corrosion and f i s s i o n products on the th0ri.a p a r t i c l e s has not been resolved. With hydrogen atmosphere the f i e l d stress of slwq w a s observed t o increase as corrosion products a c c d a t e d . Changing the atmosphere t o oxygen resulted i n a return to low yield stress, apparently as a result of change i n oxidation s t a b of i r o n and chromium adsorbed on the thoria. L i t t l e or no change i n properties has been observed i n most tests under oxygen atmosphere. Concentrations of f i s s i o n products i n most f u e l s should be considerably l e s s than the concentrations of corrosion products. Blanket s l u r r i e s have been i r r a d i a t e d i n many tests f o r 150 t o 300 h r and i n one instance f o r several months a t power densities of 0.5 t o 5 KW/liter without obvious change i n properties. Work was begun only recently on the i r r a d i a t i o n of core s l u r r i e s so no data are available.


-46 Early i n the slurry development program d i f f i c u l t y was experienced with ftcaking" the formation of hard deposits of throria i n c i r c u l a t i n g systems, T h i s d i f f i c u l t y has been overcome by c o n t r o l of preparation and p a r t i c l e s i z e i n the development of p a r t i c l e s which maintain their i n t e g r i t y .

--

One specification which has been placed on the slurry f u e l s i s t h a t they, l i k e solution fuels, m u s t contain a c a t a l y s t which w i l l recombine r a d i o l y t i c gas i n l i q u i d phase. Although t h i s has complicated the development, two promising materials have been found. One i s a molybdenum c a t a l y s t which i s added t o the slurry as molybdenum trioxide. This material appears t o be s a t i s f a c t o r y f o r the blanket sluny but has not worked well with s l u r r i e s which contain more than 1% of uranium. The second c a t a l y s t i s palladium deposited on thoria, "his material works well both with the core and blanket slurries. Both c a t a l y s t s a r e activated by r a d i o l y t i c gas. The molybdenum oxide i s activated with hydrogen and the palladium w i t h hydrogen o r oxygen before it is used t o eliminate any induction period. Although both c a t a l y s t s have been used t o recombine gas i n - p i l e i n autoclave tests, considerable additional work is required before e i t h e r one can be considered t o be demonstrated f o r r e a c t o r applications. Corrosion i n slurry systems i s e s s e n t i a l l y water corrosion with the added effect of erosion by the t h o r i a p a r t i c l e s . The combined corrosion-erosion r a t e depends on p a r t i c l e s i z e and shape, and the environment. It increases i n proportion t o the concentration of s o l i d s and approximately the square of the velocity. If the p a r t i c l e s a r e less than 1 p, the shape i s r e l a t i v e l y u n i q o r t m t j above about 5 p, erosion can be severe unless the p a r t i c l e s are spherical. Severe erosive a t t a c k r e s u l t s from direct impingement of p a r t i c l e s on objects i n the line of flow and from the c i r c u l a t i o n of material i n eddies i n regions of flow separation. Corrosion-erosion rates with the reference slurry containing dissolved oxygen are less than 1 m.il/yr f o r austenit3.c s t a i n l e s s steel piping systems where the velocity is 20 ft/sec o r l e s s and care is taken i n the design t o eliminate projections and a r e a s of flat separation, Stress-corrosion cracidng has been experienced i n stagnant regions w i t h oqgenated slurries which contained chloride, so s t r i c t control of chloride content i s very important. Although i t m y be possible t o a l l e v i a t e stress-corrosion cracking by maintaining a l a r g e excess of dissolved hydrogen i n the slurry the corrosion rates f o r stainless s t e e l are a t l e a s t four times those observed f o r oxygenated slurries. Zircaloy and titanium have demonstratzd e x c e l l e n t resistance t o erosion by s l u r r i e s , p a r t i c u l a r l y oxygenated slurries, They are used f o r pump impellers and i n other locations where the velocity i s high, but where there i s no severe t h r o t t l i n g service.


- u7 Preliminary information oncerning the e f f e c t s f r a d i a t i o n on corrosion i n slurry i s being obtained from in-pile autoclave experimnts. Indications a r e that the corrosion r a t e of Zircaloy i s increased by r a d i a t i o n but that the e f f e c t i s much l e s s than encountered with solution fuels.

-

4.

Reactor Equipment

With the exception of flow i n the reactor core, the problems of r e a c t o r experiment scale equipment f o r the uranyl s u l f a t e s o l u t i o n system have been solved and denonstrated i n â&#x201A;ŹBE--2. The flow problem i s under i n v e s t i g a t i o n a n d there Gppear t o be a t l e a s t two satisf a c t o r y solutions f o r HRE-2. It remains t o be determined whether these! solutions to the problem w i l l be s a t i s f a c t o r y f o r l a r g e reactors. Less work has been done with slurries, b u t s p e c i f i c a t i o n of the flaw design f o r the r e a c t o r vessel and development of r e l i a b l e valves and feed pumps appear t o be the major remaining major equipment problems f o r EU3-2 s c a l e slurry reactors. Work on l a r g e r equipment has been limited t o preliminary t e s t i n g of a 4000-gpm test loop by the PAR P r o j e c t and design s t u d i e s by both the PAR P r o j e c t and the HRP a t ORNL., However, much of the equipment for aqueous homogeneous r e a c t o r s i s adapted froni equipment f o r pressurized water r e a c t o r s t o meet the s p e c i a l requirements of the uranyl s u l f a t e and slurry f u e l s , The technology now being developed f o r l a r g e pressurized water p l a n t s w i l l provide a firm b a s i s f o r extension t o large homogeneous r e a c t o r systems

.

5.

primary System

a.

Reactor Vessels

The r e a c t o r primary systems are the high-pressure c i r c u l a t i n g systems consisting of the reactor vessels, c i r c u l a t i n g punps, h e a t exchangers, p r e s s u r i z e r s and piping* The envisioned equipment can be constructed and assembled f o r a l l of the three r e a c t o r s being consi.dered here. Problems a r i s e i n ensuring t h a t the equipment will have long l i f e , that i t can be maintained and t h a t the r e a c t o r vessel can be operated s a t i s f a c t o r i l y a t the design power. A vessel f o r the two-region solution-core reactor has not been designed. The p r i n c i p l e s presented here have been suggested from HRE-2 data, in-pile loop tests and recognition of hydrodynamics problems, The vessel c o n s i s t s of a 4-ft-dia by 12-ft-long cybind r i c a l zirconiumalloy core tank which i s contained i n an 8-ft-dia by 16-ft-long c y l i n d r i c a l pressure v e s s e l fabricated of cmbon s t e e l clad with stainless steel. The pressure vessel i s 6 o r 7 in. thick; the core tank i s 1/2 i n , t h i c k and i s surrounded by, and i n t e g r a l w i t h , a 3/8-in.-thick zircmium-alloy shroud which i s used t o d i r e c t


- 148

-

~

the f l m of blanket f l u i d . Fluid e n t e r s and leaves the core through concentric pipes a t one end, Cold l i q u i d e n t e r s through the annulus w i t h an a x i a l velocity of 15 f t / s e c and, along the w a l l , a tangential velocity of 15 f t / s e c b q a r t e d by vanes a t the i n l e t , It then flows the length of the tank, back through the center and out through the c e n t r s l o u t l e t pipe. On the blanket side, the cold s l u r r y e n t e r s the bottom of the pressure vessel, f l m s the length of the core through the annulus between the core tank and the shroud a t a velocity of 15 f t / s e c and then r e t u r n s t o an o u t l e t a t the bottom of the vessel through the large annulus between the shroud and the pressure vessel wall,

This design is proposed t o provide high f l u i d velocity and low temperature required t o obtain good cooling and moderate corrosion r a t e s on the core tank. It i s estimated that the maximum wall temperature can be k p t beluw 260% i n the absence of t r a n s i e n t s and t h a t the core tank l i f e w i l l be 10 to 20 years. The problem of t r a n s i e n t s has not been analyzed, Lower zirconium temperatures can be achieved a t the expense of lowering the r e a c t o r temperatures o r incorporating a separate 1000=gpm cooling c i r c u i t i n a multiwall tank, A vessel for the two-region slupry core r e a c t o r i s similar b u t larger: 7-ft-dia by 21-ft-long core and l l - f t - d i a by 25-ft-long pressure vessel, The 1/2-in,-thick core tank i s constructed i n t e g r a l with a 1/2-in,-thick shroud and the pressure vessel wall i s 8 t o 9 in. thick. The flow p a t t e r n cannot be specified b u t there is some evidence that corrosion r a t e s i n slurry systems are considerably lower than i n solution s y s t e m s o the flow requirements may be l e s s difficult, The one-region r e a c t o r vessel i s t h a t proposed by the PAR P r o j e c t but l a r g e r , It i s a sphere with conical bottom i n l e t , 4 f t OD, 12-ft-dia core with a 6- t o i'-in,-thick shell. The remaining space is occupied by thermal shields. Fluid e n t e r s the vessel as a j e t a t the apex of the inverted cone, traverses the length of the vessel, r e t u r n s along the wall and leaves through an annulus around the i n l e t nozzle, Mixing i n the v e s s e l i s promoted by induced i n t e r n a l recirculation, None of the r e a c t o r vessels contain c o n t r o l elements, Dependence i s placed on the negative temperature c o e f f i c i e n t of r e a c t i v i t y t o meet the regulation and s a f e t y requirements of the reactor. Vessels of #e s i z e s and shapes discussed above can be fabricated using present technology, Their f e a s i b i l i t y f o r r e a c t o r application depends on development of s a t i s f a c t o r y hydrodynamics and d e t a i l e d mechanical designs, The d e t a i l s of bringing f l u i d s i n t o


-1LY

-

the vessel, the s t r u c t u r a l d e t a i l s of the core tank, and he core tank t o vessel connection are among the most important problems.

b,

Heat Exchangers

Heat i s removed from core and blanket f l u i d s in shell-and-tube heat exchangers. High-pressure fluids are circulated through the tubes, feed water is heated i n the s h e l l of the blanket exchangers and lowpressure steam i s generated i n the s h e l l of the core c i r c u i t exchangers. Designs incorporating s t r a i g h t o r tubes, arranged i n horizontal or verticalbundles, with e i t h e r fixed or floating heads have been investigated. A l l types propose the use of a recirculating saturated water leg with e i t h e r an i n t e r n a l separating drum o r a separate drum connected by r i s e r s and downcomers t o the evaporating vessel. The exchangers are simila,r t o those being developed f o r large pressurized-water reactor plants,

1 i 1

Principal problems associated with the primary heat exchangers a r i s e from the s i z e and the leak tightness requirements. The designs propose the use of carbon s t e e l shells, heads and tubesheets with stainless steel cladding on surfaces i n contact with the primary f l u i d s and s t a i n l e s s s t e e l tubes. The tubesheets a r e 4 f t . dia and 8 t o 10 i n , thick and each exchanger contains 3000 to 4000 tubes welded i n t o the tubesheets, The large tubesheets are subject t o high t h e r m 1 stress. Special care must be taken i n the design and fabrication t o eliminate cracks, crevices and stzgnation points, t o have s u f f i c i e n t l y smooth flaw t o prevent f r e t t i n g corrosion and t o provide f o r complete drainage of tubes and headers. These requirements may be d i f f i c u l t t o m e t i n the complex configurations of heat exchange equipment. Absolute leak tightness i s required t o prevent the steam t o the turbine from being contaminated and because the heat exchangers are, next t o the reactor vessels, the most diffic u l t item t o maintain, they must remain leak t i g h t f o r several years. Large diameter high-pressure heat exchangers have been b u i l t t o stringent i n i t i a l leakage specifications; i t remains to be demons t r a t e d t h a t the units w i l l remain leaktight. Hold-up drums and monitoring of the steam are required t o prevent radioactivity from reaching the turbine i n the event of a leak, Designs w i t h double tube sheets and dual t d e s and intermediate heat exchange systems have been considered and discarded f o r economic reasons. C.

Fuel Circulating Punrps

Because of the stringen ge requirements, canned-inotor pumps are used exclusively f o r the f u e l circulating pumps, Condensate i s pumped i n t o the r o t o r cavity t o lubricate the bearings and prevent

F


- 150 the entrance of uranyl s u l f a t e solution and slurry p a r t i c l e s . s t a t o r s are heavily shielded t o minimize r a d i a t i o a damage.

The

Basic designs are available f o r canned-xotor pumps i n s i z e s t o 20,000 gpm. Modification of these designs i s required t o incorpor a t e the top maintenance desired f o r homogeneous r e a c t o r s and t o provide other s p e c i a l features. T i t a n i u m o r zirconiwninqpellers, d i f f u s e r s , t h e r m a l b a r r i e r s , labyrinths and s c r o l l l i n e r s may be required f o r both solution and s l u r r y pumps, Experience i n d i c a t e s t h a t long service l i f e can be expected fâ&#x20AC;&#x2122;rom s o l u t i o n f u e l pumps. D i f f i c u l t i e s have been experienced with erosion of impellers, s e a l rings and s c r o l l l i n e r s i n slurry pumps. Recent experience with Zircaloy p a r t s i n small pumps has encouraged the b e l i e f that long l i f e can be obtained from slurry pumps but they w i l l r e q u i r e more frequent maintenance than the solution pumps. d.

Pressurizers

Pressurizers proposed f o r the reference p l a n t s are e s s e n t i a l l y enlarged units that have been used on HF3-2 and on the slurry development loops, The solution system p r e s s u r i z e r involves a surge drum which i s connected by means of large-diameter piping t o the core c i r c u l a t i n g c i r c u i t s and an attached e l e c t r i c b o i l e r . Condensate produced i n the law-pressure systemis pumped t o the b o i l e r and evaporated a t rates a s high as 10 lb/min t o provide t h e desired steam pressure

.

S l u r r y p r e s s u r i z e r s are tall columns through which a p a r t of the slurry i s c i r c u l a t e d continuously. The column a c t s as a s e t t l i n g khamber leaving c l e a r water a t the top. The water i s heated t o provide the desired vapor pressure and a p a r t of the vapor i s condensed t o provide a purge of about 10 lb/min t o the r o t o r cavity of the slumy c i r c u l a t i n g pump. Similar, but much smaller, pressurizers have been operated s a t i s f a c t o r i l y f o r both s o l u t i o n and s l u r r y fuels. Problems may be encountered i n providing s a t i s f a c t o r y l e v e l control, heating and connections t o the main c i r c u l a t i n g systems i n l a r g e units. e.

Piping and Valves

Piping components f o r 2000 p s i service a r e available i n AISI/300 s e r i e s s t a i n l e s s s t e e l s i n a l l the s i z e s required f o r these reactors. A l l welded piping systems are preferred b u t the equipment can be assembled with flanged connections and leak detection equipment a s i n RRE-2 i f this s i m p l i f i e s the maintenance of equipment. Ring j o i n t flanges have proved s a t i s f a c t o r y i n s i z e s to 4 in. a t 2000 p s i i n


- 151 IB3-2 and only the main circula,ng l i n e s a r e l a r g e r i n & highpressure system. The f e a s i b i l i t y of using a few large flanged j o i n t s =quires a demonstration that 18-in., 1500-2si flanges can be ke:pt t i g h t through many thermal cycles a& when subjected t o high bending moments.

4 i

'Valves i n most of the sixes required f o r the solution-core system have been demonstrated i n HRE-2. "he l a r g e s t required f o r solutiion or slurry are 4-in. valves f o r the dump l i n e s and they must 'be developed. Work on valves f o r s l u r r y systems i s i n progress; erosiveness of the slurry makes the problem more difficult. Zircaloy and tungsten carbide show promise of s a t i s f a c t o r y l i f e as t r i m materials f o r some applications but no s a t i s f a c t o r y material has y e t been found for severe t h r o t t l i n g service.

6.

I

I

1

i

Low-Pressure Systems

:Lm-pressure equipment i s provided f o r s t o r i n g fuels when the r e a c t o r is not operating, f o r adjusting the f u e l concentration during s t a r t u p and during normal operation, f o r charging n e w fuel t o the reactor and discharging f u e l and f i s s i o n product gases t o the processiq f a c i l i t i e s , and f o r providing clean water f o r waship4 the highpressure system p r i o r t o maintenance operations. Equipment required t o c a r r y out these operations c o n s i s t s of storage tanks, evaporators having a t o t a l capacity of about 300 lb/min f o r the 1140-1NT s t a t i o n , r e c o d i n e r s , condensers and feed pumps. The low-pressure systems are designed f o r 300 psia service t o contain the r e a c t o r f l u i d s i f they a r e discharged without cooling i n the high-pressure system. Normally they are operated a t 15 t o 100 psia. The equipment is operated i n t e r m i t t e n t l y f o r slurry systems b u t

i s usled continuously with solution f'uels t o remove iodine and thereby the Xel35. This requires an additional p a r t i a l condenser and r e b o i l e r with a capacity of about 10 lb/min and the feed pump must have a capacity of 1000 lb/min instead of 100 t o 200 lb/min. %re i s no change i n s i z e of compressor used t o r e c i r c u l a t e oxygen t o the reactor b u t i t must operate continuously. The o p e r a b i l i t y and r e l i a b i l i t y of most of the low-pressure equipment f o r s o l u t i o n f u e l s has been established on small s c a l e in HRE-2. Exceptions are the exact counterparts of the 1000 lb/min feed pump, the gas compressor and the iodine removal unit. Larger multistage canned-notor p u q s w i l l have t o be developed f o r the reactor. A single-stage turbine pump having a capacity of 100 lb/min against 300 p s i head has performed s a t i s f a c t o r i l y i n 5000 hours of testing. A k e - s t a g e , 2-cfm, diaphram type gas conpressor, s u f f i c i e n t l y large and designed f o r this application, i s being i n s t a l l e d


- 152 i n a t e s t f a c i l i t y a t the present time, Silver-plated s t a i n l e s s steel mesh i s used f o r removing iodine in HRE-2, Adsorption i n liquid phase i s considered to be a more satisfactory method f o r the l a r g e r reactors , Components for low-pressure slurry systems have been designed and t e s t e d b u t have not been operated i n integrated systems even on small scale. It has been demonstrated that the s l u r r i e s can be kept suspended o r can be resuspended after s e t t l i n g i n several types of tanks by use of steam sparging, pumping and mechanical s t i r r i n g . Evaporators, s e t t l i n g tanks, hydroclones and centrifuges have been used f o r concentrating slurries. The major foreseeable problem is the feed pump, Rcrbine pumps cannot be used i n s l u r r y service. T r i m materials have not been found t h a t w i l l last more than loo0 h r i n check valves of positive displacement pumps. Batch charging methods i n which slurry i s pmssurized i n a container t o the circulating system pressure before valves i n the connecting lines are opened ham been used successfully i n loop experiments. T h i s m e t h o d may be acceptable f o r l a r g e r systems since the charging and discharging a r e normally intermittent

.

An important problem common t o the low-pressure systems is t h a t of inventory control. F a c i l i t i e s must be provided f o r sampling contents of the tanks and f o r providing continuous indication of the weight o r density and volume of materials in a l l tanks. Inventory determination by weight and sampling a s used i n HRE-2 i s the most satisfactory method that has been developed but many improvements w i l l be required t o obtain accurate control i n a large plant. The off-as s y s t e m i s an enlarged version of one which has performed s a t i s f a c t o r i l y on HRE-2, Gases from the law-pressure systems pass through cooled tanks where the short-lived a c t i v i t i e s decay. P a r t of the gas i s recycled t o the reactor from t h i s point; the remainder i s routed through cold traps or alumina dryers t o remove traces of D$ and are disc rged i n t o carbon beds t o delay the gas until a l l but the 10-yr K r 5 has decayed, Then i t i s discharged through a stack or processed f o r removal and bottling of noble gases i f t h i s i s mquired. The carbon beds are water cooled t o 1 s O q . Although i t i s possible f o r the beds t o i g n i t e since they contain oxygen, calculations, laboratory tests and experience w i t h HRE-2 show that a f i r e can be stopped and the a c t i v i t y retained by closing valves a t the i n l e t and e x i t of the affected unit, The bed can be used again i f i t i s isolated u n t i l the gaseous a c t i v i t i e s decay and then i s purged or evacuated t o remove adsorbed carbon dioxide. MolecuLar sieve materials can be used as adsorbents i n place of carbon but they are l e s s e f f i c i e n t .

P


7. Reactor Plant Instrumentation Instrumentation and control systems f o r aqueous homogeneous reactors a r e similar t o those used i n modern high-pressure steam power and chemical plants. Some nuclear instruments are required but the largesk p a r t i s d i m c t process instrumentation, Problems attendant t o the special requirements of fluid fuel reactors has required development of special components. Special liquid-level transmitters, l e v e l probes, d i f f e r e n t i a l pressure c e l l s , weigh syst e m s and f l o w transmitters were developed f o r use i n HRE-2. I n most instances they have proved satisfactory and provide a b a s i s f o r improved devices f o r large solution f u e l reactors. For slurry systems, substantial improvemats must be made t o instruments now i n use f o r measuring flow, density l e v e l and pressure before they can be used i n a power reactor , A combined electric-pneumatic signal transmission system i s used i n "3-2 and would be recommended f o r large reactors. I n the control room, e l e c t r i c signals f'rom prirnary variable sensing elements are converted by transducers t o pneumatic signals and these are used t o actuate miniature pneumatic display instruments and pneumatic valves i n the reactor. The a b i l i t y to use all metal, radiation r e s i s t a n t cons.t,ruction f o r transmitters and valve actuators makes the pneumatic system particularly a t t r a c t i v e f o r application i n high radiation fields. Radiation damage t o the primary e l e c t r i c a l elements i s avoided by the use of inorganic e l e c t r i c a l insulators such as ceramic, mica, magnesium oxide and magnesium s i l i c a t e ,

8.

htrbogenerator Plant

The plant which u t i l i z e s steam f'rom the reactor differs from a fossfl-fuel plant i n two respects. The steam is law-pressure saturated steam and protection m u s t be provided t o prevent the turbine equipment from becoming highly contaminated i n the event of a burst tube i n a steam generator. Shielded drums are provided t o delay the steam for 5 sec between the steam generators and the turbine. The steam i s monitored and the dnuns a r e isolated by valves i n the steam l i n e i f a c t i v i t y i s detected. The turbine is the same as those being designed f o r presswized water and boiling water reactors. They a r e presently l i s t e d i n s i s e s t o 300 NW.

9 . Reactor Building The power plant contains a variety of strmctures, the l a r g e s t of which are the reactor and turbogenerator buildings. Others are provided f o r f u e l processing, waste handling, mintenance, water t r e a t ment, etc. The reactor building i s considered here because of i t s


- 154 specZa1 requirements. The turbogenerator building i s conventional and should require no discussion. Where other s t r u c t u r e s are Unique they w i l l be described i n conjunction with t h e i r functions.

Two types of buildings have been considered f o r housing aqueous homogeneous reactors. They d i f f e r p r i n a r i l y i n the philosop& of containment and i t is d i f f i c u l t t o decide which w i l l be required f o r large p l a n t s or whether the two w i l l be combined. The first was used i n HRE-2 and involves i n s t a l l i n g the r e a c t o r and a d l i z r i e s i n compact c e l l s which c o n s i s t of an inner metal l i n e r , the complete b i o l o g i c a l s h i e l d and an outer metal s h e l l assembled as an i n t e g r a l structure. The top of the c e l l i s l a r g e l y removable i n sections which a r e keyed i n t o girders that support the shield blocks i n two layers, A metal diaphragm i s i n s t a l l e d between layers and welded t o the g i r d e r s so t h a t t h e c e l l i s sealed completely during operation. Service and instrument lines which penetrate the c e l l walls terminate i n service g a l l e r i e s adjacent t o the cells. CeUs and operating galleries f o r one or more r e a c t o r s are contained i n a canyon which communicates with the maintenance f a c i l i t i e s . A crane with shielded cab is provided for use i n maintenance operations and f o r transporting equipment from the c e l l s t o the maintenance storage pool. The building t o house three solution-core r e a c t o r s i s 300 f t long by 100 f t wide, divided i n t o three r e a c t o r equipment c e l l s and adjoining service g a l l e r i e s . The c e l l s and g a l l e r i e s are 60 f t deep, and the crane bay r i s e s t o a height o f 50 f t above the cells, The building f o r the two-region slurry core or the one-region s l u r r y reactors would be about 125 f t wide by 225 f t long. The second type of c o n t a i m n t i s t h a t proposed by the PAR Project. It involves constructing c e l l s similar t o those described above for the r e a c t o r and a u x i l i a r i e s except that the top i s not sealed and the shielding i n most areas i s l e s s than that required t o reach b i o l o g i c a l tolerance when the r e a c t o r i s operating. These c e l l s and the service g a l l e r i e s are housed i n a l a r g e spherical o r c y l i n d r i c a l s t e e l vessel which provides the f i n a l containment i n the event of the d m u m credible accident. A crane with shielded cab i s ins.t;E;lled i n the containment vessel f o r maintenance operations, and provision i s made f o r removing equipment t o the shops. A 2-ft-thick concrete wall surrounding the containment vessel completes the b i o l o g i c a l shielding during operation and reduces the r a d i a t i o n levels i n the surrounding areas i n the event that f i s s i o n products are dispersed i n the sphere. Three 15O-ft dia spheres or three v e r t i c a l cylinders approdmately 120 f t dia by 180 f t would be required f o r the two-region solution-core reactors. The two-region-slurry core or one-region slurry r e a c t o r could be contained i n a sphere about 200 f t dia,


Both types of containment would r e t a i n ompletely the radioa c t i v i t y from small leaks and s p i l l s o r the most l i k e l y major s p i l l , rupture of a s m a l l pipeline that had become weakened by erosion o r corrosion. The more hazardous, though far less l i k e l y , occurrence would be the "maximum credible accident," rupture of the large primary piping o r the r e a c t o r vessel i t s e l f . I n the event o f a major rupture, the pressure i n the completely sealed c e l l s , which a r e n o r d l y kept a t 7 t o 10 p s i a , would r i s e t o a maximum of about 50 p s i a and then gradually decline to l e s s than 15 p s i a i n 8 hr. A f a i r assumption as t o d i s p e r s a l of a c t i v i t y would include a l l of the ga.seous component and 30 percent of the a c t i v i t y associated with the s o l i d phase. I d e a l l y these would be retained within the c e l l ; however, with the large nutnber of penetrations, a complete s e a l i s p r a c t i c a l l y impossible A reasonable leakage r a t e is estimated t o be about 1% per day when the pressure i s 50 p i a , s o during the first 8-hr period approximately 0.1% of t h e dispersed a c t i v i t y could be expected t o l e a k from the c e l l t o the service g a l l e r i e s . Ventilation from the g a l l e r i e s and the crane bay i s passed through filters and washed so that the a c t i v i t y discharged from the s t a c k i s limited t o krypton and xenon isotopes. By proper c o n t r o l of v e n t i l a t i o n i n the g a l l e r i e s the contaminated areas can be kept t o a minimum.

.

I n the event of such a major rupture i n the large vapor cont a i n e r , the s e a l s of the primary concrete c e l l would r e l i e v e t o the vapor container. The pressure i n the vapor container would r i s e t o a maximum of 1 2 psig and gradually decline t o about 6 psig a f t e r two hours; airborne a c t i v i t y would be uniformly dispersed throughout the c e l l s and the space within the vapor container. Three elements of r a d i a t i o n hazard following such an unlikely accident a r e of concern: d i r e c t r a d i a t i o n through the secondary conc:rete s h i e l d , "sky-shine" o r s c a t t e r e d r a d i a t i o n from the open top of the concrete shield, and airborne contamination f'rom small l e a k o r penetrations through the vapor container. I n t h e PAR s t u d i e s - i t was calculated t h a t the d i r e c t r a d i a t i o n l e v e l a t a p o i n t on the ground 50 f t from the concrete cylinder w i l l i n i t i a l l y be 100 r/hr, decreasing a t such a rate t h a t the time required t o a c c d a t e a dose of 25 r w i l l be 45 minutes. The dose r a t e a t the end of a month, assuming no decontamination i n the meantime, would be 1 r/hr. A t a point 1000 f t from the base of the r e a c t o r , the

The numbers apply t o a LSO-Nt?T, one-region slurry r e a c t o r and are based on all the f i s s i o n products, becoming airborne. They would be _. increased i n proportion t o the t o w 1 r e a c t o r power and decreased i n proportion t o the f r a c t i o n assumed t o be dispersed t o apply t o other r e a c t o r systems

st

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d i r e c t r a d i a t i o n dose rates w i l l be 5 r/hr and 50 mr/hr, i n i t i a l l y and a t the end of one month, respectively. A t a point 2500 f t from the reactor, assumed to be the s i t e boundary, the corresponding two r a t e s w i l l be 100 mr/hr and 1 mr/hr, respectively.

The r e s u l t s of a rough calculation i n d i c a t e t h a t the e f f e c t of sky-shine a t the 50-ft point w i l l be about equal t o t h a t of d i r e c t radiation; thus, based on a limiting dose of 25 r, p l a n t personnel would have about 20 minutes t o evacuate t o the c o n t r o l room, which has additional l o c a l shielding. The e f f e c t of sky-shine a t points more remote i s s i g n i f i c a n t l y l e s s important.

It i s assumed t h a t a t 1 2 psig, the vapor container w i l l leak a t a d a i l y r a t e of 0.1% of i t s f r e e volume. The r e s u l t i n g dose r a t e from the radioactive plume, a t a point 2500 f t downwind of the r e a c t o r under the most unfavorable meteorological condition of tlfumigation,fl w i l l be 4 r/hr after 15 minutes and 2 r/hr after 15 hours. A more favorzble wind condition of tlconinglt w o u l d halve those rates.

As t o breathing tolerances, the most hazardous element released would be radioiodine. Under the extreme assumption t h a t a l l of the iodine i n the r e a c t o r i s uniformly dispersed, the maximum downw'nd concentration 2500 f t from the r e a c t o r i s approximately 3 x 10 pc/ml, or about double the m a x i m u m permissible concentration f o r an 8-hr exposure. The calculation does n o t take c r e d i t , however, f o r the tendency of iodine to nucleate on airborne p a r t i c l e s and thus t o be washed down t o the bottom of the vapor container.

2

The most desirable type of construction i s one which confines the contamination t o a minimum a r e a and from this standpoint the sealed c e l l s are preferred. F e a s i b i l i t y of this s y s t e m depends on the a b i l i t y of the builders t o ensure freedom from leaks through the penetrations. I n the event of a catastrophic rupture of the primaqy system i n a p l a n t b u i l t with unsealed cells i n a l a r g e vapor cont a i n e r , contamination i s wide-spread i n t h e container b u t plant; personnel w i l l have adequate t i m e t o move to the shielded control room and there w i l l a l s o be time f o r an orderly evacuation of people living i n the immediate neighborhood of the p l a n t site. 10.

Reactor Operation and Control

Aqueous homogeneous reactors rely f o r control completely on v a r i a b l e f u e l concentration and temperature c o e f f i c i e n t of react i v i t y rather than poison-containing control rods. Shim c o n t r o l t o compensate f o r f i s s i o n product poisons, t o change the operating temperature, t o put the reactor i n t o operation and t o shut down i s accomplished by adjusting the core fuel concentration.


- 157 -

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I n s t a r t i n g the reactors the core fue i s concentrated i n the lowpressure s y s t e m and heavy water i s pumped i n t o the reactor core and core circulating systems. A t the same time the s l u r r y i s pumped i n t o the blanket of a two-region reactor. Circulation of the f l u i d s i s started when the f i l l i n g i s completed, pressurizing i s s t a r t e d and steam frox an auxiliary supply i s admitted t o the heat exchangers. The reactor systems are heated a t a r a t e of 4OoC (lCh%) per hour until the temperature reaches about 20O0C (3929). During this time fuel i s pumped i n t o the core system but a t a r a t e that w i l l keep the reactor subcritical. Fuel is added a t t h a t temperature until the reactor becomes c r i t i c a l and then i s continued so t h a t the temperature rises t o 250°C i n about two additional hours. Release of s t e m from the steam generators t o the turbine i s begun a t 250°C and the steaq release r a t e and addition of f u e l t o the F a c t o r are controlled t o keep the reactor i n l e t temperature a t 250 C as the reactor o u t l e t temperature and power are raised t o the normal operating level. I n the case of the solution-core reactor the addition of f u e l i s stopped and only small concentration changes are made t o compensate f o r the growth of small amounts of poison or t o change the temperature level. Feed t o the slurry-core reactors i s continued as the Xe13.5 builds i n and further u n t i l the temperature begins t o decrease s l i g h t l y as a r e s u l t of increasing resonance absorption. A t this point fluctuations i n concentrations w i l l have l i t t l e e f f e c t on the r e a c t i v i t y and the s t a r t u p i s complete.

The temperature r i s e of the f l u i d passing through the reactor is 4OoC s o r a p i d changes i n power output can be made without serious thermal shock t o the reactor equipment. Changes i n output are effected e i t h e r by regulating the flow of steam a t the turbine o r by regulating the flow of feed water t o steam generators t o maintain a constant steam pressure. A normal reactor shutdown i s accomplished by reducing the f u e l concentration i n the core by discharging f u e l t o the low-pressure system and r e t d n g condensate t o the high-, pressure system. During this several-hour cooling period the cohcent r a t i o n i s reduced a t about the maximum capability of the low-phssure s y s t e m and the flow of steam from the steam generators i s r e g y h t e d t o control the temperatures t o prevent the thermal s t r e s s e s from exceeding safe limits. When the equipnent has cooled t o a$ut 100°C, the f l u i d s are discharged t o the storage i f maintenance i s t o be done or they are retained i n the reactor i f the shutdown i s f o r other reasons. If a serious leak develops ir the equipment, an emergency dump can be made from high temperature and pressure i n 30 min.

/’

Dependence i s placed on the negative temperature coefficient of r e a c t i v i t y to meet the regulation and safety requirements of th reactor. me temperature coefficient v a r i s from -3 x 10-3 a Qk per OC f o r the two-region reactors t o -1 x 10’ k/k per OC f o r the large

3


- 158 one-region reactors, A s there i s no requirement t h a t the shorttime variations i n power be kept below several percent, regulation i s achieved with ease. Control of f u e l feed r a t e , steam release rate from the steam generators and pump startup r a t e i s provided t o l i m i t the foreseeable r e a c t i v i Q r i s e r a t e t o values t h a t can be accommodated e a s i l y by the s y s t e m . Mechanical operation o f the plant i s a combination of standard power plant operation and radiochemical plant operation. Operating d i f f i c u l t i e s which d i f f e r from those i n the normal power p l a n t are concerned with control of inventory and transfer of radioactive fluids. Because the f u e l i s f l u i d , care must be taken and controls provided t o prevent the inadvertent escape of radioactive material. The major hazard i n the operations i s the taking and transporting of f u e l samples, a p a r t of the inventory procedure. With continuing development, equipment w i l l be devised s o t h a t only minor hazards e x i s t there,

ll. Maintenance of Reactor and Other Radioactive Eaubment A major problem of f l u i d f u e l systems and one that may be a deterf a c t o r i n t h e i r ultimate economic f e a s i b i l i t y i s that of maintenance. Because the systems are highly radioactive the maintenance m u s t be done remotely o r sed-remotely. Special tools are required, the operations are time consuming and care must be taken t o control contamination t o prevent i t s spread throughout the plant, Naintenance of radioactive equipment on a large scale has been demonstrated i n the processing p l a n t s a t Hanford and Savannah R i v e r . Haintenance of aqueous homogeneous reactors has been demonstrated It remains t o be determined whether on p i l o t plant scale with HRE-2. maintenance of a large parer reactor plant can be done economically.

The maintenance methods chosen have a profound influence on I n the first place, they d i c t a t e the arrangement and spacing of almost a l l components and runs of piping, and a s a consequence the physical size of the plant as a whole. Second, i t i s found t h a t a l l components, including those that are quite conventional, m t b e specifically designed w i t h a remote maintenance procedure foremost i n mind.

major elements of plant design.

I n any plant, components m y be e i t h e r repaired i n place during shutdown or, alternatively, replaced by a spare and repaired after the plant is restarted. The choice depends i n each case on the r e l a t i v e costs of downtime against the cost of the spare component. I n 8 f l u i d reactor plant, t h i s is no l e s s true than i n a conventional plant, except t h a t the d i f f i c u l t i e s of remote in-place r e p a i r and the

,


- 159 high cost of downtime keep t o a very small number t h e components which w i l l be repaired i n place. I n HRE-2 some maintenance operations are done w i t h special t o o l s through special shields with the c e l l dry. However, operations which1 involve removal. of large pieces or opening of large areas of the cell are done. by use of long handled t o o l s with the c e l l flooded t o provide shielding. Although i t i s possible t o design a large plant f o r the same type of co&ined dry and underwater maintenance, the PAR Project chose remote dry maintenance f o r a l l operations on a large reactlor. This general concept would be used i n the plants proposed here except t h a t the c e l l s would be designed so that they could be flooded t o m a b semi-direct work possible i f necessary,

I

I n the PAR design, the basic maintenance t o o l i s a shielded-cab bridge crane which bas access from above t o the majority of the components. The functions of t h i s crane are t o l i f t the concrete slabs which form the roofs of the several large compartments; t o place positioners, cutters, welders, wrenches and other remotelyopera-ted tools; and t o remove the affected component t o the maintenance pool adjoining the reactor building. The crane i s a s s i s t e d i n these operations by bridge- and wall-mounted manipulators which a r e placed a t s t r a t e g i c locations within the compartments. Wherever advisable, small components are grouped f o r removal as a sub-assembly and i.n each case the l i f t i n g point is d i r e c t l y over the center of gravi.ty of the component o r sub-system. An important fraction of the PAR e f f o r t was devoted t o the development o r adaptation of the special machines needed f o r the execution of the concept described above. A t the time of suspension o f the Project, these were at different stages of design or development, but i n a l l cases encouraging progress had been made. An experimental model of a 10-in. automatic welding machine applicable t o remote operation was b u i l t , and a number of full-strength welds were made in s t a i n l e s s s t e e l pipe. A co&ination remote positioner-cutterwelder f o r one-inch pipe and i t s control console was b u i l t and t e s t s about t o begin. Prelidesigns were completed f o r a 16-;in, floor-mounted positioner and f o r a pipe-mounted positioner of intermediate size. Progress had been made i n the adaptation o f ultrasonic techniques f o r weld inspection, but a considerable amount

be used in e l i n e s i n the low-pressure s y s t e m , in s m a l l l i n e s i n the high-pressure s y s t e m , and i n the large& l i n e s of the primary system i f remote welding proves to be unsatisfactory o r uneconomical, Ring j o i n t flanges could be used i n all s i z e s , but equipment would have t o be developed f o r remote positioning and boltji-bg of the flanges and f o r machining damaged grooves. If such

e


- 160 operations appear t o be feasible, the leak detection system, which has been demonstrated on HRE-2, provides an effective means f o r testing the flanged connectiions. There are two major exceptions t o the applicability of the techniques a d machines described above. Repair of a reactor vessel would require that i t be removed t o the maintenance pool, Depending on the damage, the vessel might be discarded or i t might be repaired by semi-remote methods using t o o l s prepared f o r the particular job. Special provisions must a l s o be made f o r repairing the steam generators. I n the PAR work methods were under development f o r doing the steam generztor maintenance i n place from rooms adjacent t o each of the steam generators. Design zmd some development were done on equipment f o r surveying the tube sheets, detecting leaky tubes, cutting orf the tube ends, plugging the tubes and welding the plugs, Successful tests were made of the plugging and welding elements of the machine. A t the present time, no estimate can be made as t o the frequency with which component repairs w i l l have t o be made. Estimates have been made a s t o t h e length of plant down-tine which might be incurred by some of the major maintenance operations. By way of example PAR Project estimated that i t would take about 5 days from full-power t o full-power, t o remove and replace a primary pump. Included i n this t i m e are 21~hours f o r a normal plant shut-down and for ventilation p r i o r t o opening the primary compartment, and 32 h r f o r plant startup a f t e r the compartment i s closed, To open a steam generator, search out and repair one or more tube leaks, and reclose the generator would require about 6 days, exclusive of the time required f o r plant shutdown and startup. Three days would be used f o r the Peakdetection survey i t s e l f and 2 days f o r the rewelding of the two 16-in. access nozzles. A maintenance f a c i l i t y i s provided adjacent t o the reactor b u i l d ing f o r repairing equipment and f o r preparing f o r disposal equipment o r p a r t s t h a t must be discarded. The f a c i l i t y i s provided with decontamination equipment for removing surface a c t i v i t y and a storage pool ab8 f t by 40 f t by 30 f t deep capable of storing heat exchangers o r reactor vessels.

A portion of one end of the pool projects i n t o a hot-shop, wherein items which have been allowed t o decay i n the pool but which are s t i l l too radioactive f o r d i r e c t maintenance can be reclaimed. The hot-shop i s approximately 20 f t wide, 60 f t long and 20 f t high, surrounded by 5-ft-thick concrete walls f o r biological shielding. Since a l l operations are performed remotely, the shop i s equipped with maintenance devices which are operable from outside of the shield,

c

c


- 161 Such devices are a 1 -ton overhead crane with an auxiliary hook f o r l i g h t equipment, a heavy-duty overhead manipulator, and wallmounted d p u l a t o r s f o r l i g h t and intermediate service. I n order t o observe and d i r e c t the operations, the shielding walls are penetrated a t appropriate places with windows, and a t two of these observation points srnall master-slave manipulators a r e mounted t o f a c i l i i , a t e maintenance of a more d e l i c a t e nature than could be handled by the heavy-dubj manipulators. A turntable i s provided t o faci.li.tate work on large heavy it,ems.

I n addition t o the hot-shop the maintenance f a c i l i t y contains the mockup shop w i t h t h e j i g s and f i x t u r e s t h a t are required f o r fabricating new piping and equipinent for the reactor. Also there are the instrument shops, nacbine shop and general maintenance and s t o r e s f a c i l i t i e s f o r the e n t i r e plant. 12.

Waste DisDosal Systems

IVastes from the reactor plant consist of the ventilating air, s o l i d s which accunulate fron maintenance and laboratory operations and f r o n i o n exchange beds and l i q u i d wastes from s p i l l s , sampling and a n a l y t i c a l operations, regeneration of ion exchange beds, equipment decontamination, and the f u e l t r a n s f e r areas where c a r r i e r s are charged for shipnent to c e n t r a l processing f a c i l i t i e s . Addi+Aonal tankage required f o r handling wastes from ill? on site processing p l a n t would be included i n the waste area but the amount i s not specified here and the c o s t i s included as a c o s t i n the operation of the processing plant. Ventilating a i r withdrawn from the p l a t f a c i l i t i e s a t a r a t e of abont 50,000 c f r n i s washed i n a spray chamber, f i l t e r e d , and dischmged up a 250-ft stack, S o l i d wastes are stored i n a b u r i a ground; fine residues o r large numbers of small pieces are combined i n druns before burial..

.

Eore elaborate f a c i l i t i e s a r e required f o r the liquid wastes S p i l l s or laboratory wastes that involve a considerable q u m t i t y of fissionable material are discharged t o a 2500-gal c r i t i c a l l y safe storage tank. Xaste i s discharged from thki tar& t o an evaporator where it i s concentrated, and then transferred t o the on s i t e processing p l a t or t o the fuel. c a r r i e r c'narging area f o r shipment. Xastes t h a t do not involve l a r g e q u a n t i t i e s of fissionable material a r e discharged t o one of four 60,000-gai storage tanks. From the storage tanks they a r e s e n t to evaporators f o r concentration. Gondens a t e from the evaporators i s collected i n one of two 10,000-gal condens a t e tanks, sampled, discharged t o the ion exchange units as inake-up f o r


- 162 the maintenance pool o r t o a low l e v e l storage pond, o r recycled i f the a c t i v i t y level is t o o high. Bottoms from the evaporators are discharged t o a 200,000-gal permanent storage tank or, i f

s u f f i c i e n t uranium is involved, routed t o the c a r r i e r loading area. E.

Fuel Processing

The processing of aqueous homogeneous reactor f u e l s i s c a r r i e d out t o remove f i s s i o n products and corrosion products which absorb neutrons and cause solution f u e l i n s t a b i l i t i e s . For the blanket of a two-region reactor, processing has the primary purpose of separ a t i n g fissionable and f e r t i l e material, and removal of f i s s i o n and corrosion products i s of secondary importance. The Thorex process i s r e a d i l y adaptable f o r the recovery of decontaminated U-233 and thorium from both solution f u e l and slurries f o r aqueous homogeneous reactors.

1.

Solution Fuel Processing

Uranyl sulphate solution from the reactor core is processed through a s y s t e m of hydroclones a t a rate of 100 gpm, one reactor volume p e r hour, t o concentrate insoluble f i s s i o n and corrosion products i n t o a small volume of f u e l solution i n the underflow tank. Slurry discharged from t h i s tank a t convenient i n t e r v a l s represents the major withdrawal of material from the f u e l s y s t e m . The use of hydroclones f o r this purpose was demonstrated on HRE-2, where a s i n g l e 0.S-in. hydroclone removed the insoluble materials. The demonstration was not c o q l e t e l y s a t i s f a c t o r y because only 10 t o 20% of the s o l i d s produced i n the reactor reached the hydroclone. The f u e l solution after passing through the hydroclones, i s l e t down t o the low-pressure system, flashing about 1%of the D20. T h i s D20 vapor e f f e c t i v e l y s t r i p s iodine, xenon, krypton, and dissolved 02 from the f u e l solution as demonstrated with "3-2. Separation of these gases from the D20 can readily be effected in a liquid-gas absorption tower. Iodine i s relatively non-volatile i n pure water at the lower temperature and is removed from the bottom of the tower dissolved i n a small stream of D20. The bulk of the D@, s u f f i c i e n t l y pure f o r reuse without further treatment, may be recovered by condensing the overhead from the tower, and Lhe xenon, krypton, and 02 are discharged as non-condensables.

With hydroclone processing and continuous removal of iodine and rare gases, soluble nickel from corrosion of s t a i n l e s s s t e e l becomes the major chemical and nuclear contaminant i n the f u e l solution. The concentration is kept a t 0.02 m, an acceptable l e v e l from the standpoint of neutron poisoning, by %e withdrawal of f u e l


- 3-63 solutfon from the hydroclone underflow tank a t the rate of 60 liter:s/day. However, since dissolved nickel contributes so heavily t o so:Lution i n s t a b i l i t y , i t i s advisable t o remom nickel a t a rate s u f f i c i e n t t o hold the concentration t o 0.005 m or less. It i s proposed t o treat about 200 liters/day of f u e l s o l u t i o n from the reactor low-pressure system i n an e l e c t r o l y t i c c e l l f o r removal of coppe:r and nickel, Such a c e l l works b e s t with a mercury cathode. A f t e r adding f r e s h copper catalyst, the treated f u e l solution can be returned d i r e c t l y t o the reactor. The mercury can be purified f o r reuse by passing s u l f u r i c acid solution through the electrol y t i c c e l l and reversing the electrode potential. The r e s u l t i n g nickel, copper, sulphate solution becomes a radioactive waste. This method has been developed and successfully tested only i n the laboratory, The e l e c t o l y t i c removal of nickel i s highly desirable f o r a reactor operated as a breeder where the amount of fissionable mater i a l i n the chemical plant becomes an important consideration i n determining doubling t i n e , because l i t t l e decay time i s required before electrolysis. However, i n case doubling time i s not an overriding consideration, sizing the hydroclone underflow withdrawal rate t o control nickel appears t o be the simplest, most economical procedure.

The hydroclone underflow tank contents and the iodine-D20 solution from the iodine removal system are combined with slurry removed from the reactor blanket low-pressure system f o r subsequent D20 and U-233 recovery as described i n the next section. 2.

Slurry Blanket Processing

Slurry f o r processing i s removed from the blanket system a t convenient intervals. The t o t a l amount of s l u r r y processed for a 333-HdE s t a t i o n v a r i e s from about 1350 liters/day (1350 kg Th/day) f o r a 3 reactor power breeder s t a t i o n operated t o obtain a 6- t o 10-yr doubling time, t o only about 120 liters/day (120 kg Th/day) f o r tihe blanket of a s l u r r y fuel reactor operated t o minimize fuel cycle cost. The slurry, cornbined with the underflow from the hydroclone, is then sent t o D2Q recovery. The D20 is recovered f o r i m c t i a t e reuse by evaporating the slurry t o dryness and heating the s o l i d s t o about 300oC. T h i s w i l l drive off 99.9% of the D20, based on laboratory tests, Larger scale t e s t s indicated that foaming during evaporation and entrainment of Tho2 p a r t i c l e s i n the off-gas are the most serious problems associated w i t h D 2 0 recovery.

After D20 recovery, the material i s stored f o r about 150 days f o r decay of short-lived f i s s i o n products, Pa-233 and Th-234, before


- 164 solvent extraction processing. This period i s s u f f i c i e n t t o reduce the U-233 loss as Pa-233 and the Th-234 contained i n the thorium t o i n s i g n i f i c a n t levels. If processing i s done on the reactor s i t e , the solids may be dissolved i n HNO following the drying step and stored as a solution for decay. I the material i s t o be shipped t o a c e n t r a l chemical processing plant, both the drying f o r 020 recovery and decay storage may be done i n a suittibly designed shipping container.

3

By using two f u l l cycles of Thorex solvent extraction, the thorium and U-233 are recovered e s s e n t i a l l y f r e e of f i s s i o n product a c t i v i t y , but both contain isotopic contaminants t h a t Eake them d i f f i c r i l t handling problems. The uranium contains U-232 which adds to the alpha hazard and whose daughters are energetic gamma emitters. The thorium contains appreciable Th-228 zf ter several cycles through the reactor. T h i s a l s o contributes added alpha a c t i v i t y and energ e t i c gamma emitting daughters.

3.

Slurry Fuel Processing

I n the case of a slurry f u e l e s s e n t i a l l y a l l the f i s s i o n products and corrosiofi products are adsorbed on the surface of the particles. There is a p o s s i b i l i t y t h a t 12, Xe and K r c a n b e stripped from the slurry i n the reactor, but the technology for accomplishing this does not now exist. Therefore, a t present, slurry f u e l processing must be considered as removing the slurry from the reactor, recovering 320 by evaporation, decay storage, and decontamination by Thorex. For a 333-1WE i n s t a l l a t i o n the amount of slurry f u e l processed per day i s about 200 liters o r 50 kg of thorium. Laboratory t e s t s have indicated that i t may be possible t o remove the bulk of the f i s s i o n product and corrosion products from the slurry with a HNO3-H2O2 leach, without dissolving the slurry. However, more work i s necessary t o determine whether or not this represents an improvement over sending a l l the thorium through Thorex and reconstituting the slurry from decontaminated thorium and uranium.

4.

Reconstitution o f Fuel Solution

This has been s a t i s f a c t o r i l y accomplished i n two ways, both equally a t t r a c t i v e . F i r s t , uranium from the Thorex product uranyl n i t r a t e solution can be sorbed on a cation exchange resin, eluted w i t h H2SOb, the uO2SO4 evaporated t o dryness and redissolved i n D20. Second, the uranium can be precipitated from the uranyl nitrate sol u t i o n as U04, f i r e d t o anhydrous U@3, and redissolved i n QSO4 solution.


Both methods can be c a r r i e d out rapidly enough t o avoid serious radiation problems from the daughters of U-233

5.

Reconstitution of Blanlcet Slurry

The preparation of purp Tho2 f o r the blanket of a two-region reactor involves precipitating thorium oxalate from t h e Thorex thorium nitrate product solutlon and decomposing this materizl t o theanhydrous oxide by f i r i n g . Present indications are t h a t a f i n a l f i r i n g temperature of 800-900째C i s s u f f i c i e n t i f care i s taken to thoroughly digest the thorium oxalate after precipitation. The p a r t i c l e s i z e of the final oxide i s controlled by the cond i t i o n s under which the oxalate i s precipitated. The concentration of the reagents, the r a t e of reagent addition, and temperature a l l a f f e c t p a r t i c l e size. Present practice consists of adding 1H oxalic acid t o 1 M Th(NO3)4 a t 25% and a t a rate determined & produce a f i n a l aGerage p a r t i c l e s i z e of 2.2 *, 0.2 microns. The oxzkite p r e c i p i t a t e i s then digested f o r 48 hr a t 85%, f i l t e r e d and f i r e d t o 800OC. The present scale of this operation is about 200 lib of ThO2/batch. I f subsequent work indicates that a f i r i n g temperature well above 1000째C i s required, the procedure becomes more complicated. The oxide, after being f i r e d i n the e l e c t r i c furnace, m u s t be transferred t o a gas f i r e d furnace, f i r e d t o the desired temperature, discharged, s l u r r i e d i n water containing oxalic acid t o disperse the oxide?, c l a s s i f i e d t o remove oversize s i n t e r s , centrifuged, and r e f i r e d i n an a l e c t r i c furnace t o remove the oxalic acid.

6 . Preparation of Fuel Slurry The simplest, most s a t i s f a c t o r y method of adding uranium t o controlled p a r t i c l e size ThO2 i s the following. The t h o r i m o x a l a t e is prepared as before t o control the f i n d p a r t i c l e size; however, the thorium oxalate i s f i r e d only t o 65ooc t o produce a high surface area oxide. Ur.zn5.a i s deposited on the preformed Tho2 p a r t i c l e s by slurrying the thoria i n a. solution of uranyl ammonium carbonate which decomposes t o UO3 on heating t o 100째C. The UO p r e c i p i t a t e s on the Tho2 p a r t i c l e s , and the s o l i d s are recovered y centrifugation. The uranium i s then diffused i n t o the Tho2 by f i r i n g a t 10sO째C t o give a homogeneous mixed oxide. This oxide has been produced i n 200-lb batches.

a

The decay products of U-232 and Th-228 w i l l make f u e l slurry r e c m s t i t u t i o n a highly radioactive operation. Shielding w i l l be

e 1

t


- 166 required and a l l materials must be carefully contained t o prevent spread of the alpha active materials. F.

Economic Appraisal

(Project Directorls Appraisal)

Power c o s t s are composed of fixed charges on invested c a p i t a l , operating and maintenance labor and supervision, maintenance m a t e rials, and f u e l costs. The present s t a t u s of technology i s such t h a t only an "order of magnitude" estimate can be made f o r most of these costs, p a r t i c u l a r l y when they are f o r plants whose proposed construction i s many years i n the future. Even this wpe of ar, estimate requires preparation of a conceptual design of the plants, equipment, processes, and operations i n considerable d e t a i l . The reactors discussed here have not been studied i n the necessary d e t a i l . Costs presented f o r the two-region s o l u t i o n core breeder were obtained by examining a preliminary study of a reactor plant made several years ago and escalating the c o s t estimates of t h a t study taking i n t o account changes i n cost index and experience. i n reactor equipment fabrication. Power p l a n t c o s t s were based on a comparison with today's c o s t f o r conventional plants. Costs f o r two-region slurry core breeder and one-region slurry converter were obtained by comparison of the p l a n t requirements with those of the reference plant. Although the c o s t s are considerably higher than has been projected by the Homogeneous Reactor P r o j e c t and others f o r ''nth generation plants" they agree with c o s t s estimated by the PAR Project f o r a first large commercial plant.

i

I

-

Investments i n the three types of plants excluding f u e l inventory, fuel processing and waste f a c i l i t i e s and. fixed charges based on a 14% annual charge, 0.8 plant f a c t o r and 333 MkE were estimated t o be the following: P l a n t Cost Fixed Cost Reactor Type

$M

I

Two-region solution core breeder Two-region slurry core breeder One-region slurry converter

111,000 107,000 104 000

,

$ I k w E ' m 334 322 312

6.7 6.4 6.2

Details of the c o s t estimate f o r the solution core breeder a r e shown i n Table VII-1 and XI-2 Cost of operating the power plants, again excluding the f u e l processing and waste f a c i l i t i e s was estimated from personnel requirements t o be 0.23 %lls/Kwh f o r each of the f l u i d f u e l plants, and i s assumed here t o be independent of type of aqueous reactor. A n n u a l maintenance c o s t s were assessed a t 3% of the plant c o s t f o r lack of a b e t t e r number. Total operating and maintenance costs f o r solution

,

c L

!


- 167 -

t

core, s l u r r y core a d one-region reactors are Mills,/Kwh re spec t i v e l y

.

.7, 1.6 and 1.5 i

On s i t e processing of f u e l was specified as being requir;!d f o r a l l the plants. The minimum c o s t f o r a processing plant, including a n a l y t i c a l and waste f a c i l i t i e s which w i l l be shared by the reactor plant, was judged to be $11,790,000. T h i s p l a n t would be capable of processing and reconstituting 100 kg 1 day o r less of thorium and associated uranium from aqy of the aqueous homogeneous reactors. A costbreak-down i s shown on page 37 and i n Table XI-2. Annual operating and maintenance costs were assumed t o be 15% of the p l a n t cost, For the two-region solution core breeder, higher processing rates are required t o obtain s h o r t doubling time. An estimate of the relationship of c o s t t o s i z e f o r on s i t e processing p l a n t s is shown below: Plznt capacity, @/day Investment, $M Fixed charges, $M/yr Operating costs, $M/yr Total, $.X/yr U n i t cost, $/kg Th

4100 11,790

1,650

1,770 3y L20

11,700 kg/d

300 14,790

1,000 22 000

4,000 50,000

2 j 070

3,080 3j 300

7 y 000 7 ,500

2,220 L y 290 49

,

6,380 22

14,500 12.5

Fuel cycle costs are presented i n Table X I - 2 as calculated from the f u e l cycle d a t a i n Table XI-1. They were calculated f o r several blanket processing rates f o r the two-region reactors and f o r 180-day and 60-day residence i n processing t o show the e f f e c t of doubling time ton c o s t .

"he effect of the short cycles in both reactor and

processing is t o reduce the inventories a t the expense of higher processicg costs. It was assumed t h a t the c o s t i n $/kg of processing on the 6 0 4 7 period would be 1.5 times that estimated f o r the 180-day period because of the higher r a d i o a c t i v i t y l e v e l s and the necessity f o r extracting or recycling Pu233. The lowest f u e l c o s t s calculated, including fixed charges on the rocessing, waste and laboratory f a c i l i t i e s a r e 2.5 t o 2.7 Itills Kwh. The 2.7 Mills/Kwh i s fox the two-region solution core reactor and corresponds t o a doubling time of 18 years. Reducing the doubling time t o 14, 10 and 6.5 years results i n f u e l c o s t s of 3.1, 4.1 and 5.6 M i l l s / K w h respectively. The minimum doubling time f o r the two-region slurry core breeder i s shown t o be 26 years. Reduction of this time requires development of a process f o r reducing the xenon poison level.

P

The t o t a l energy c o s t s are summarized below:

i


- 168 Reactor

Two -Regi on Solution Core

Fixed charges on power plant, Mills/Kwh Operation & Maht of power plant Mills/Kwh Fixed charges on processing plant,

Mills/Kwh Operation & Maint of processing plant, Mills/Kwh Inventory charges, Ivlills/Kwh Burnup and losses, Mills/Kwh Total energy cost, Mills/Kwh

Tw o-Regi on One -Region Slurry Core Slurry

6.7

6.4

6.2

1.7

1.6

1.5

0.89

0.72

0.71

0.95 1.61 -0 10 11.2

0.78

1.07

0.76 0.82

-0.05

0.34 10.3

.

10.5

Increase i n power p l a n t s i z e i s the f a c t o r t h a t should have the largest e f f e c t on reducing c o s t s of power from aqueous homogeneous reactor s t a t i o n s . This i s because many of the f a c i l i t i e s provided f o r the 333 MWE s t a t i o n would increase very slowly i n size with increase i n nwiber of reactors and s t a t i o n output a t one site. It i s premature t o attempt t o evaluate the e f f e c t of s t a t i o n s i z e on cost of the reactor plant; much development i s required t o e s t a b l i s h the f e a s i b i l i t y of the systems and whether the c o s t s estimated here are too high o r too low. Eowever, the f u e l processing i s s i m i l a r enough t o existing processes t h a t the e f f e c t of p l a n t s i z e on f u e l cost can be considered. The t o t a l f u e l cycle c o s t s including inventory, burnup and losses, and fixed, operating and maintenance c o s t s have been estimated f o r s t a t i o n s which produce 1000 and 2000 ,W a t one s i t e taking i n t o account the e f f e c t of processing plant s i z e on the u n i t c o s t of processing discussed above. Costs a s compared w i t h those f o r .the 333 WE s t a t i o n are the following: S t a t i o n size, MldE Total f u e l cycle cost, Mills/Kwh Two-region solution core breeder 10 y r doubling time 14 y r doubling t i m e 18 y r doubling time Two-region s l w j core breeder 45 y r doubling time 26 y r doubling time One-region slurry converter

333

1000

2000

4.1 3.1

2.5 2 .o

1.9

2.7

1.7

1.5

2.5

1.7

3.5 2.6

1.4

2.1

1.7

2.1

1.9 1.L

The growth trend i n the power industry i s toward very large generating stations. i n 1950 8% of the e l e c t r i c a l generating capaci t y i n the United S t a t e s was in p l a n t s having capacities of 500 t o 1000 M.4 and there were no p l a n t s of capacity g r e a t e r than 1000 I4k-J l i s t e d i n the Federal Power Commission report. I n 1960 the percentages


- 169 and 1000 M plants respectively. w i l l be 23 and 11 f o r 500-1000 Aqueous homogeneous reactors are b e s t s u i t e d f o r the large generating s t a t i o n s where advantage can be taken of large on s i t e processing p l a n t t o achieve low f u e l cost.

The requirement f o r processing f u e l s on site places a c o s t penalt,y on the one-region r e a c t o r and the two-region slurry core breeder reactors. It has been estimated f o r the 333 MiCE s t a t i o n that the investment i n arialyticzl, waste and handling f a c i l i t i e s on s i t e could be reduced from the $11,790,000 t o $7,000,000 o r less. The c o s t of shipping, processing, and reconstitutiEg the f u e r i n a c e n t r a l p l a n t having a capacity of 6000 kg/day (investment cost $100,000,000) is estimated t o be $25/kg. This qeduces the annual charges f o r processing of the 50 kg/day f o r the one-region slurry reactor from $3,~20,000 t o $2,~00,000, or l e s s , resulting i n a savings of a t l e a s t 0.4 I$ills/Kwh and 2 reduction i n t o t a l f u e l c o s t (including fixed charges, operation, and maintenance of on s i t e f a c i l i t i e s ) from 2.6 t o 2.2 Elills/Kwh. Fuel c o s t s f o r the two-region slurry core breeder would be-reduced from 2.5 t o 2.2 Ylills/Kwh by shipment off site f o r processing. These savings decrease with increasing s i z e of power p l a n t and a r e i n s i g n i f i c a n t when the power plant s i z e s reach 1500 t o 2000 IWT.

-

!

G.

-

Research and Development Program (Project Director 9s Appraisdl )

m e r e are three inportant stages i n developing a reactor system i n t o a. commercial power plant. They a r e (I) determination of the compatibility of f u e l and s t r u c t u r a l materids and development and demonstration of processes and engineering on a reactor experiment or p i l o t plant scale; (2) extension of the engineering t o large-scale systems and demonstration on a prototype; ( 3 ) improvement of equipment and processes and construction of a commercial plant. S t e p 1 is the important step i n determining the t e c b o l o g i c a l f e a s i b i l i t y ; s t e p 2 i s the important s t e p i n determining the economic f e a s i b i l i t y ; s t e p 3 is the beginning of exploitation of the system. A schedule f o r this type of program f o r the Am and estimates of the c o s t s are presented i n Table XI-3. Development of the solution f u e l system has passed through much of stage 1. Completion of this stage requires t h a t the core tank and s o l u t i o n s t a b i l i t y questions be resolved and t h a t HRE-2 be reb u i l t with a slurry blanket to demonstrate the f e a s i b i l i t y of both core and blanket systems. FY 1961 is the most importznt year of the program. Solutions t o several c r i t i c a l problems m u s t be obtained i n t h a t year i n order for the! program t o continue.


- 170 1.

Analysis of the behavior of the f u e l i n HRE-2 must be completed, making use of data from HRE-2 and f r o m i n p i l e loops and autoclaves.

2.

The correlation used i n predicting corrosion r a t e s f o r zirconium alloys in-pile must be confirmed by experiments a t fluxes approaching those specified f o r large power reactors

.

3. A p r a c t i c a l design m u s t be developed f o r a large reactor vessel. Hydrodynamics experiments f o r both core and blanket and thorough analysis of the nuclear and mechani c a l design problems are required.

4.

Encoursging progress m u s t be made i n developing slurry pwnps and valves, and methods for charging and discharging s l u r r y from the blanket of a r e a c t o r experiment.

5

Slurry with engineering properties acceptable for a reactor experiment must be demonstrated i n t e s t out-ofpile

.

6.

It must be demonstrated t h a t extended i r r a d i a t i o n of the slurry i n autoclaves has no serious adverse e f f e c t s on the properties of the blanket slurry, the a c t i v i t y of the reconibination c a t a l y s t , o r the corrosion of Zircaloy.

.

7. An acceptable conceptual design must be made f o r the HRE-2 modifications

Assuming t h a t s a t i s f a c t o r y answers are obtained f o r the c r u c i a l questions i n FY 1961, modification of HRE-2 can begin i n FY 1962. During the modification period p a r t of t h e equipment development e f f o r t w i l l be aimed a t insuring s a t i s f a c t o r y operation of equipment and processes f o r the r e a c t o r experiment. Work on f u e l s and mater i a l s w i l l emphasize the investigation of e f f e c t s of radiation on slurries and container materials i n in-pile loops and autoclaves. In-pile work on the solution f u e l s w i l l be continued but a t a reduced l e v e l t o improve data on the e f f e c t s of radiation on the solution f u e l s and on corrosion by those fuels. The schedule shown requires t h a t a conceptual design be made f o r the prototype reactor i n FY 1962 and t h a t work begin on the development of maintenance tools and methods, c i r c u l a t i n g pumps, feed pumps, valves, core tank t o pressure vessel j o i n t s and other items t h a t involve long times f o r development, procurement or testing. Work must begin on development of s p e c i a l methods and equipment f o r processing the f u e l s from the prototype.


- 171 Operation of HRE-2 would begin again i n FY 1964 and p a r t of the development e f f o r t would be associated wit!! the reactor operatiort. The development e f f o r t cannot be projected beyond this time w i t h certainty, It i s assumed that design of the prototype

wod-d begin e a r l y in FY 1965 and construction would begin late i n the year, Effort on fuels and materials would be reduced gradually unless unexpected troubles developed. Development e f f o r t spent on t e s t i n g and improvement of equipment, instruments, and maintenance devices would continue a t a steady l e v e l during the reactor cor,e;truction. Operation of IIlzE-2 is shown as being terminated a f t e r theee years although i t i s possible t h a t this operation would be continued t o provide additional data f o r operation of the prototy-pe. The f u e l processing section of the prototype p l a n t would begin operation i n FY 1968 t o prepare f u e l f o r the reactor and the e n t i r e p l a n t would be completed i n FY 1969, A t this time most of the developnent e f f o r t would be associated d i r e c t l y with the plant operation, Design and construction of the f i r s t commercial p l a n t cou:ld begin any time after FY 19â&#x20AC;&#x2122;70 with FY 1974 a s the e a r l i e s t completion date a Programs and schedules f o r development of one -region s l u r r y converter and two-region slurry core breeder reactors would be sixilar t o that presented above. Problems would be confined t o those of the slurry f u e l s o the costs should be l e s s . The development of a s a t i s f a c t o r y recombination c a t a l y s t assuws much greater importance i n these programs because the power density i n the core slurry is much g r e a t e r than i n the blanket slurry, The reactor experiment would begin as a c r i t i c a l experiment t o confirm that inhomogeneities i n the slurry create no serious nuclear s t a b i l i t y problems. Successful operation as a c r i t i c a l experiment would be followed by i n s t a l l a t i o n of heat removal equipment t o permit operation a t power d e n s i t i e s approaching those of the large pover reactors. The one-region converter reactor would be emphasized i n the early development, A two-region slurry core breeder would be a n a t u r a l outgrowth of the one-region converter.

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XI1

.

The Atomic Energy Commission, Dr. Frank K. Httman, Director, Division of Reactor Development, and t h e Chairman and Vice Chairman of the Task Force wish t o express t h e i r appreciation t o t h e members of t h e Task Force and t h e i r parent organizations. The Task Force met f o r a one-week, preliminary meeting i n December of 1958, and convened f o r a continuous two-month session during January and February of 1959. The p a r t i c i p a n t s worked many long and d i l i g e n t hours i n order t o perform t h e i r t a s k i n t h e l i m i t e d time available.

--

A number of organizations Ebasco Services Incorporated; Edison E l e c t r i c I n s t i t u t e and Commonwealth m i s o n Company; Tennessee Valley Authority; and United Ebgineers & Constructors Incorporated provided high c a l i b r e personnel t o s e r v e on t h e Task Force with no reimbursement from t h e Atomic Ehergy Commission. Their i n t e r e s t i n a s s i s t i n g t h e Commission evaluate these programs i s deeply appreciated.

Acknowledgement i s a l s o due t o Robert T. Schomer, b n a g e r o f t h e & \Ililcox Company; David H. Fax, Technical Assistant t o t h e Manager of t h e PAR Project, Westinghouse E l e c t r i c Corporation; William N. Flobley, Hanford Atomic Products Operation, General E l e c t r i c Company; and Robert H. Shannon, Director of Muclear Power Development Division, United Engineers & Constructors Incorporated who served as consultants during all or p a r t of t h e study.

LMFRE P r o j e c t , The Babcock

Many o t h e r people from The Babcock s( Wilcox Company, Brookhaven National Laboratory, O a k Ridge National Laboratory, and Westinghouse E l e c t r i c Corporation appeared before t h e Task Force on numerous occasions. The Commission and t h e Task Force wish t o express t h e i r appreciation t o these persons.

i


- 177 i

I

Excerpt o f s e c t i o n of ltAd Hoc Advisory Committee's Report on Reactor P o l i c i e s and Programs" on "Appraisal o f Present S t a t u s and Prog~~arn'~ of "Fluid &el Reactors

.

ttFlu:id Fuel Reactors "There are two p r i n c i p a l reasons f o r i n t e r e s t i n f l u i d f u e l r e a c t o r s . One of t h e s e i s t o supplement the preponderance of work on heterogeneous o r s o l i d f u e l r e a c t o r s t o give added insurance f o r achieving low c o s t nuclear power through high temperature or s i m p l i c i t y of t h e f u e l cycle, o r both. The second reason i s t h a t f l u i d f u e l systems appear t o o f f e r t h e only p o s s i b i l i t y f o r thermal breedi n g on t h e thorium U-233 f u e l cycle with an acceptable doubling time say t e n years o r l e s s . Uncertainties i n t h e value of e t a ( t h e number of neutrons emitted per neutron absorbed) make i t impossible t o give p o s i t i v e assurance t h a t breeding on t h i s cycle can, i n f a c t , be achieved. The Committee s t r o n g l y recommends t h a t urgent e f f o r t be devoted t o measurements of eta f o r U-233 by various l a b o r a t o r i e s and by various methods.

--

" U n t i l b e t t e r knowledge of eta i s obtained, work on thermal breeders should be held t o a minimum. However, development of a thermal breeder r e a c t o r , i f shown t o be f e a s i b l e , should be included as p a r t o f t h e overall program a t approximately t h e present l e v e l of e f f o r t on t h e aqueous homogeneous project. Based on present informat i o n , breeding with t h e thorium U-233 cycle i s uncertain f o r aqueous homogeneous r e a c t o r s , even more uncertain f o r l i q u i d metal f u e l r e a c t o r s , and n o t possible f o r molten s a l t r e a c t o r s .

3

"Of t h e three f l u i d f u e l r e a c t o r types, t h e aqueous homogeneous i s of i n t e r e s t as a breeder and because of its s i m p l i f i e d f u e l cycle, t h e molten salt is of i n t e r e s t because of high temperature operation and s i m p l i f i e d f u e l cycle, and t h e l i q u i d bismuth because i t might incorporate a l l t h r e e advantages. A l l three types, p a r t i c u l a r l y t h e l i q u i d bismuth, present many d i f f i c u l t development problem. For example, i t i s probable t h a t f o r s u c c e s s f u l breeding i n t h e aqueous homogeneous o r i n t h e l i q u i d bismuth r e a c t o r s , a s l u r r y w i l l have t o lie used at least i n a blanket, and t h e development of adequate s l u r r y technology is a very d i f f i c u l t job.


- 178 ""he three fluid fuel systems should be evaluated first as to promise for producing low cost nuclear power and not more than one project supported on the basis of that objective. idhen additional information on the value of eta for U-233 is available, the situation with respect to thermal breeding should be re-examined, with a view toward increased support. IJe see no present basis for the construction of fluid fuel reactors of any of the three types."


i

APPEDDIX B

List of P r i n c i p a l Items Considered To Be Included i n Power Ilant Investment For Power S t a t i o n With Fluid Fuel Reactors Shown i n Table V I I - 2

310

LAND

31:~

STRUCTUIES & IMpROVE.NE8TS

11 .12 e

13

S i t e F a c i l i t i e s (General Use F a c i l i t i e s ) Access Railroad & Yard Trackage Access Road, Plant Roads, Parking h e a s & Walks Fences & Gates Yard Lighting System I n t r a s i t e Communication System Service Water System Wells & Pumps Elevated Storage Tank Service Water Piping Fire Protection System Domestic Mater Treatment & Distribution S a n i t a r y Sewer System

.2 21 22

. 0

23 24

25 .26 ,261

262 263 ,264 265 27

.

03 .31. 311 312 e313 314

,315

.316 e 317

,318 e

a

1

319

LAND RIGHTS

S i t e Improvements S i t e Preparation S i t e Drainage Final Grading & Landscaping

01

I

&

.

S t a t i o n Buildings Nuclear-Steam Generator Building Excavation & Backfill Concrete work ( i n c l , Shielding 8t Floors) S t r u c t u r a l & Misc. S t e e l Encls, (Walls & C e i l . o t h e r than concrete) Floor Covering Doors ?i Windows Roofing Painting Building Services

k


- 180 e3191 -3192 3193 3194 03195 03196 3197 03198

.. .

Plumbing & Drain. F a c i l i t i e s Lighting F a c i l i t i e s Heating & Air Conditioning Ventilation F a c i l i t i e s aevators Building Crane Hoists F i r e Protection Equipment

..32321 322

323 .324 325 .326 327 -328 0329 3291 3292 e3293 e 3294 0329s 3296

Service & Plaintenance Building Excavation & B a c k f i l l Concrete Work ( i n c l Floors) S t r u c t u r a l & Elisc. S t e e l &CIS. (\falls& C e i l . o t h e r than c o m r e t e ) Floor Covering Doors Windows Roo f i n g Painting Building Services Plumbing & Drain. F a c i l e Lighting Facil. Heating, Ventil., & Air Cond. Building Crane Kitchen Equipment Fire Protection

e33 331 332 0333 334 335 336 -337 e 3371 3372 03373 3374 e3375 3376

Turbine & A u x i l i a r i e s Building Xxcavation & B a c k f i l l Found. & Concrete (incl. Floors) S t r u c t u r a l & Misc. S t e e l Encls. (l.lalls & C e i l . o t h e r than concrete) Doors & Windows Painting Building Services Plumbing & Drainage F a c i l i t i e s Lighting F a c i l i t i e s Heating, Vent., & Air Cond. Personnel L i f t Hoists Fire Protection Equip.

e

..

.

.

.

Chemical

35 e351 e352 e353

Other S t a t i o n Buildings Warehouse kater Treatment Building Chlorine Cylinder Storage Shed

34

Waste Buildings (Listed Separately)


- 181 e354 0355 356

G a s Cylinder Storage Shed Combustibles Storage Bldg. Gate House

31.2 e 1

e11 e

12

-13 14 015

16

02 e21 .21a e e

21al 21a2 21a3

21a4 21a5

21a6 21a7 21a8 -21a9

21a10 21b

21bl 21b2 21b3 2lb4 .21b5 21b6 21b7 21b8 e 21bg 21b10 *22 22a

.22al i

i

.22a2 22a3 22a4 22a5

Nuclear Reactor Equipment Core Vessel Core Graphite Thermal Shield Reactor Vessel (Pressure S h e l l Control Rods & Nisc. Hardware Supports & Shield Heat Transfer System Primary (Fuel-Coolant 1 System Fill & Storage Equipment Melt Tank Filter F l o w Control Valve Freeze Valve Storage (drain-dump) Tanks Storage Tank I n t e r n a l s Fill Pumps Pressure Syphon Fill System ( i n c l . Valves) Dump Valves Fuel Addition-lhrichment Equipment

k t

Primary Loop Equipment Primary Loop ,Pumps Blanket Loop P u m p s Primary Loop Heat Exchangers Blanket Loop Heat Exchangers Primary Loop P r e s s u r i z e r s Blanket h o p P r e s s u r i z e r s Primary Loop Check Valves Blanket Loop Check Valves Fuel Sampler Remainder of System Intermediate (Coolant) System Fill & Storage Equipment Filter Fill Valve F i l l Pump Storage Tanks Dwnp Valve

L

t

k


- 182 -

. .. .

22b .22bl 22b2 22b3 22b4 22b5 22b6 .22b7

03 31 311

..312

313 .0315 314

-316

.317 318 319 32 .321a .321b 322a 322b *323a 323b 324a 324b 32%. 325b 326a 326b 3273 327%

.. .. .

.

.

.328

*33 331 332

333 334 0335 336 337 0338

.

Intermediate h o p Equipment Intermediate Loop Pumps Expansion Tanks Sugerheat ers Reheaters Cold Trap P r e c i p i t a t o r s Cold Trap Heat Exchangers Plugging I n d i c a t o r s Reactor Plant Auxiliaries Radioactivity Containment-Confinement Reactor Vessel Containment Off-Gas System Containment f i e 1 Fluid (drain-dump) T a n k s Containment Fuel Fluid fill Pumps Containment Primary Loop Pumps Containment Primary Loop P r e s s u r i z e r s Containment Fuel Fluid Valves containment FJel Fluid Piping Containment Fuel Addition-Enrichment Equip. Containment Let-Down & Recombiner Equipment Core Fuel G a s Separators Blanket Fuel Gas Separators Core Fuel Let-Down Economizers Blanket Fuel Let-Down Economizers Core Fuel Led-Down Coolers Blanket f i e 1 Let-Down Coolers Core Fuel Evaporators Blanket Fuel Evaporators Core Fuel Fhtrainment Separators Blanket Fuel Entrainment Separators Core Fuel Recombiners Blanket Fuel Ilecombiners Core Fuel Condensers Blanket Fuel Condensers condensate Storage Tanks Reactor Off-Gas Equipment Hold-Up T a n k s Cold Traps FP Absorbers Vacuum Pumps Compressors Hg Seal Pots F i l t e r s - P u r i f i c a t i o n Equipment Heat Exchangers


- 183 -

.34

..35351 . .3!S3

3-52

I n e r t Gas Equipment Purge Equipment Purge Coolers Purge Pumps Purge Pump Coolers

.361 36 ..36la 3~iai ...361a3

HeatinE, Cooling, & Ventilating Equipment Primary System Heating & Cooling Equipment Reactor Heating & Cooling - Equipment Helium Blowers E l e c t r i c Heaters 36h2 Heat &changers (Water o r Dowtherm Cool. Fill & Storage Heating & Cooling Equip. .361b Helium Blowers .3Bibi E l e c t r i c Heaters 36lb2 Heat Exchangers (Water or Dowtherm Cool.) ,361b3 Primary Loop Pumps & Pipe Heat. & Cool. Equip. .36lc H e l i u m Blowers .31~1~1 E l e c t r i c Heaters .361~2 Heat Exchangers (Water o r Dowtherm Cool.) .361c3 primary Heat Exchangers Heat. & Cool. Equip. .361d Helium Blowers .361di E l e c t r i c Heaters 361d2 Heat Exchangers (Water OF Dowtherm Cool. .361d3

.

.

.

362

Intermediate System S t a r t u p Heating Equipment Cell Cooling Equipment (closed system) "Fin-Fan" Cooling U n i t s Water Circulating P u m p s Heat Exchanger (water coolers) Storage (dump) Tank

d

Shield Cooling Equipment (closed system) Thermo Panel Coils Water Circulating Pumps Heat &changer Storage (dump) Tank

Raw Water Cooling System

..g661 366 ..3662 3663

Pumps Ventilation System

Fans

Filters Ductwork

k


F

- 184 -4

..4142 .421 422

.516 517

*515

.52521

0522

523

53 531

-532 e533

.6 6a 6b

..

6c 6d 6e .6f 6g

..6h6 i ..6j61 .6m 6k

.60

6n

6P

.696r

Steam Generator Equipment Steam Generator (Boiler) Loeffler System Components Steam Pumps Zvaporator D r u m s Steam Generator Auxiliaries Boiler Feedwater Equipment Feedwater Heaters E o i l e r Feed Pumps Deaerator & -4ux. ( i n c l , Pumps) Condensate Transfer Pumps Heater Drain Pumps Condensate Storage Tanks Reducing & Desuperheating S t a t i o n Feedwater Testing. Equipment Demineralizing Equipment Chemical hmps & Tanks Laboratory Equipment Service Boiler & Fuel O i l Equipment Service ( o i l burning) Boiler Service O i l Unloading & Transfer Pumps Service O i l Storage Tank Reactor & Steam Generator Piping Systems ( i n c l . Valves) Primary Fuel-Coolant râ&#x20AC;&#x2122;ill & Storage Piping Primary Heat Transfer Loop Piping Intermediate Coolant F i l l & Storage Piping Intermediate Coolant Loop Piping Loeffler System Piping Off-Gas System Piping I n e r t G a s System Piping LSeactor Heating & Cooling Piping fill & Storage Heating & Cooling Piping Primary Pump & Pipe Heating & Cooling Piping Primary Heat Exchanger Heating & Cooling Piping C e l l Cooling System Piping Shield Cooling System Piping Raw Water Cooling System Piping Steam System Piping Feedwater System Piping Feedwater Treating System Piping Service O i l System Piping


i - 185 -

. .. .

.J

i

‘?a .‘7b ‘7c ‘7d 7e .’7f 7g 7h

.-77ij .‘7k

08

..81 ..838485 ..8686a .. ..86e . ..86j 86i 82

86b 86c 86d

86f 86g .86h

..86k .861 86m 86n .860

.91 09

92

Reactor & Steam Generator P l a n t I n s u l a t i o n lieactor Vessel Contain. & Heat. & Cool. Equip. Off-Gas System Contain. & Heat. & Cool. Equip. Fuel-Fluid F i l l & Stg. Con. & Heat. & Cool. Equip. Prim. Loop Pump & Pipe Contain. & Heat. & Cool. Equip. Prim. Heat Exch. & Heat. & Cool. Equip. Interm. System Equip. & Piping Lk Heat-Up Equip. C e l l Cooling Equipment & Piping Shield Cooling Equipment & Piping Steam Generator Equipment Steam System Piping Feedwater System Equip. & Piping Controls & Instrumentation Equipment Panels & Consoles Central Control System I?rimary Loop System Intermediate Loop System Steam Loop ( i n c l . L o e f f l e r ) System Auxiliary Systems Nuclear Instrument & Health Monitors I n e r t & Off-Gas System Fuel Fluid 2311 & Stg. System k’eactor Heating & Cooling System m e 1 Fluid Fill & Stg. Heat. s( Cool. System Primary Loop ?ump & Pipe Heat. & Cool. System Primary Heat Exch. Heat. & Cool. System Intermed. Loop S t a r t u p Heat System Cell Cooling System Shield Cooling System R a w Water System Ventilation System Feedwat e r System Feedwater Treating System Elec. & Penumatic Equipment Hot C e l l & Remote Maintenance Equipment Hot C e l l Equipment Remote Maintenance Equipment TUKBINE-GB4ERATOR PLANT EQUIF”T,

314 01

!L‘urbine Foundation

.2

Turbine-Generator & Exciter

ETC

.


I i

- 186 03 31

Turbine-Plant Auxiliaries Lubricating O i l Equipment Hydrogen Cooling Equipment Gland S e a l Equipment Closed System Cooling Water Equipment Turbine Room Instruments & Control Panel Crane & Aails Turbo-Generator S h e l t e r

32

33 e 34 35 .36 37

0

04 41

Turbine Plant Piping General Piping Insulation

42

05 51

Condenser & Auxiliaries Condenser Circulating Water a m p s Condensate Pumps Vacuum Pumps

52

53 054

.6 61 611 612 62 6 2 ~ 622

.623 6231 6232 6233 6234

63 631 632 0

64 65 651 652

66 67

31.5

ACCESSOIlY EUCTRICAL EQUIPMXNT

01 .11 12 0

0

Circulating Water System River Intake Hack S t r u c t u r e P i l i n g & Fender S t r u c t u r e s S t r u c t u r a l S t e e l Hack Screen & Pump Chamber Excavation & B a c k f i l l Concrete S t r u c t u r e Equipment Traveling Screens Stoplogs Pumps & Piping Chlorination Equipment Intake & Discharge Pipe Reinforced Concrete Pipe S t e e l Pipe, Valves & Connections Discharge S e a l Well S t r u c t u r e Discharge Canal Excavation Blanket & Riprap Zscharge Structure River Bank Revetment

13

Foundations & S t r u c t u r e s Generator Leads Foundations & Supports Equipment Foundations Manholes & Handholes


I

i i

23 231 232

Power & Conversion Equipment S t a t i o n Trans formers S t a t i o n Service Transformers Lighting Transformers Einergency Power F a c i l i t i e s Motor Generating Equipment D-C Power Equipment Storage B a t t e r i e s Battery Charging Equipment

e3 31 e311 312 e 32 321 e 322 333 .334 335

Conduits, Conductors, & I n s u l a t o r s Conduits, Ducts, & Trays Underground Conduits & Ducts EXposed Conduits & Trays Conductors & I n s u l a t o r s Generator Leads S t a r t e r Transf, Feeders Power Wiring Control Wiring Miscellaneous

e4 41 .411 412

.433 434 e

Switching;, Control, & Protective Equipment Control Boards & Panels blain Control Board f o r BTG Load Frequency Control Equipment A-C & D-C Distribution Panels Metal-Clad Switchgear 4160 V. S t a t i o n Auxiliary 480 V. S t a t i o n Service Other Control Equipment Motor Control Load Centers Misc. Remote Control Equipment Fault Protectors Indicators

05

S t a t i o n Grounding System

e2

21 e 211 e 212 e 22 .221

e

e

.

413

. .

e.

42 421 422 43 431 432

MISCEUANZIUS POWER PLANT EQUIPMENT

316

.1 .2

-3 04 e5 -6 .7

Compressed Air Equipment General Shop & Maint. Equipment Office f i r n i t u r e & Fixtures S t o r e s Fixtures and Equipment First Aid Equipment F i r e Extinguishing Equipment Miscellaneous


I

- 188 SUGGEW LIST jiLT3ENTS O F DISTRIBUTIVE COST

DIRECT CONSTRUCTION COST (Direct Material & Labor)

INDIRECT CONSTRUCTION COSTS Temporary Construction Buildings, U t i l i t i e s , & F a c i l i t i e s Construction Equipment & Small Tools Office Equipment, Supplies, & Expense Warehousing Ekpense Local Sales & U s e Taxes Payroll Insurance & Taxes (Workman's Cornp., Sot-Sec, Tax, etc.) I n c i d e n t a l Labor Ekpense (recommitment, bargaining, weld, qualif., etc.) Technical Services (inspection, laboratory t e s t s , e t c , Field Supervisory & C l e r i c a l Payroll (all i n d i r e c t on s i t e l a b o r ) Construction Permits & Licenses Performance Bond

OvERIIEaD CONSTRUCTION COSTS Preliminary Studies & Investigations Engineering & Design Legal Fees & Taxes During Construction Insurances During Construction ( o t h e r than p a y r o l l insurance) I n j u r i e s & Damages Mot Covered by Insurance (floods, land damages, etc.) Contractors' Fees (home o f f i c e expense and p r o f i t ) Start-up & f i e l i d n a r y Operation &pense ( p r i o r t o commercial operation 1 Administrative & General Escpense Interest During Construction ALLOWANCE FOR OMISSIOKS & CONTINGENCIES (on everything above)

TOTAL CONSTRUCTION COST (sum of above)

TID-8507  

http://www.energyfromthorium.com/pdf/TID-8507.pdf

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