ORNL-3791 by Jon Morrow

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Contract No.

W-7405 eng-26

CHEMICAL TECHNOLOGY D I V E ION

PRELIMINARY DESIGN STUDY OF ACONTINUOUS FLUORINATIONVACUUM-DISTIIUTION SYSTEM FOR HEGENERATING FUEL AND F E R T I U STREAMS I N A MOLTEN SALT BREEDER REACTOR I

\ I

Cm D. S c o t t

W. L. Carter

UARY 1966

OAK RID03 NATIONAL LABORATORY Ridge, Tennessee operated by ' CARBIDE COReORATION f o r the I C ENERGY COMMISSION

L ; . ,

iii

CONTENTS

------------------------------------------------------------Abstract 1 .. ----------------------------------------------~---------Introduction 3 ------------------------------The Molten SaltBreeder Reactor System 5 _ -----------------------------------------------------Des& Criteria _-_---_-------_-_----------------------------Basic Considerations 77 Process for the Fertile Stream ----------------_------------------9 Process for the Fuel Stream ----------------------,---------L,,,,,, Waste Storage --_--_--------------_______c____________------------1: . Operating Policy ---,----,--,--,,-,,,,,,,,,,,,,,,,,,,,,,,---------10 Process Data -------_--_--_---------------------------------------11 11 Description of Process ------*--,~,----,----------------,---,-,~--------_-------_---------------------Summary of the Process Flowsheet 16 ----_------__----------------------------------------Fluorination -16 Purification of Uranium Hexafluoride*by Sorption and Desorption --- 19 , ----_-------_--_------------------------------Vacuum Distillation 25 Reduction of Uranium Hexafluoride and Reconstitution of the Fuel -- 28 29 Off$as Processing ------_----_-_------------------------------------------c-_-c_-_--_--------------------------------Waste Storage 29 Flow Control of.the Salt Streams --_-----------_------------------;z * Removal of Decay Heat ---__-__-------_----------------------------Sampling of the Salt and Off-Gas Streams ----------_------_-------36 Shielding, Maintenance, and Repair of Equipment CL----------------36 --i-,----,---,,,,,-,-,,,-,,,,,,,,,,,,,,,I Materials of Construction General Operating Policy --------------------________L___________-i; Process Design ,-----------,----,--,--------,,,-,,,,,,,--------------Fuel Stream -------_-----_-_--------------------------------------$ Fertile Stream -----------------_---------------------------------Plant Design and Layout _----_---------------------------------------52 Cost Estimate ---------_--_-----_------------------------------------Process Equipment -------------,-,,--,--,---------,,,,--------;i Structure xand Improvements ---_---_-------------------------------Interim Waste Storage _-_______----_------------------------------56 Other Plant Costs -------_-__----_--------------------------------57 Total Fixed Capital Cost ------_----------------------------------Direct Operating Cost -----------------------------------,--~-:i Processing Cost ,-,,--,,--,,-;,,,-,,,,,,,,,,,,,,,,,,,,,,----------~ 60 62 Conclusions and Recommendations -------------------------------------66 Acknowledgement ,,,,,,,,-,,,-,,,------------------~------------------References --C--------------------------------r---------------~--------67 forFuel Salt Fluorinator and Appendix A. Design Calculations Cooling Tank ---------------------------------------,---~ 73 I _

c ?

bid

iv Appendix B.

Fission Product Heat Generation Rates in the MovableBed Sorbers and N&l? Waste Tanks ------------------------~-78 Movable-Bed Sorber -----------------------------------Sodium Fluoride Waste Containers. ---- ----------------Short-Term Cooling Station for Waste Sodium Fluoride-Interim Storage of Waste ------------------------,-------

z -*+s J/-T

g 80 81

Appendix C. Appendix D.

Estimation of Distillation Rate in Vacuum Still -,------83 Fission Product Accumulation and Heat Generation Rate in Lithium Fluoride Pool in Vacuum Still ----------------89 Analytical Expression for Heat Generation Rate ---A--- 90 Evaluation of Vacuum-Still Design ----------'---------95 Appendix E. 'Design Calculations for WasteStorage System -----L--i--99 100 FuelStream Waste System _--_-__--------------------109 FertileStresm Waste System ------------------------Appendix F. Physical-Property Data and Drawings -------------------1 110

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PKELIMINWX IESIGN STUDY OF A CONTINUOUS FLUOFUNATIONVACUUM-DISTIIJAmON SYSTFN FOR IIEGENERAmNG FUEL AND Ii*ERTIm STRJUMS IN A MOLTEN SALT BF3EDER HEACTOR C. D. Scott

W. L. Carter

ABSTRACT The purpose of t h i s study was t o make a preliminary design and an engineering evaluation of a conceptual plant f o r t r e a t i n g t h e f u e l and f e r t i l e streams of a molten-salt breeder reactor. The primary requirements of the process are t o recover the unburned f u e l (233UJ?4) and fuel-stream c a r r i e r salts (LiF-BeF2 from the f u e l stream, and the L i F J l " 4 plus the bred 33U from the f e r t i l e stream. Both streams m u s t be s u f f i c i e n t l y decontaminated f o r attractive breeding performance of the reactor. The plant was designed t o operate continuously as an i n t e g r a l part of the reactor system, f i t t i n g i n t o two r e l a t i v e l y small cells adjacent t o the reactor c e l l . In t h i s study, the plant capacity is based upon t r e a t i n g 15 ft3/day of f u e l salt and lo5 ft3/day of f e r t i l e s a l t removed as side streams. !These capacities are adequate f o r a 1000-M~( e l e c t r i c a l ) power reactor.

A s t o the f u e l stream, b a s i c a l i y , i t is purified by fluorination and vacuum d i s t i l l a t i o n . The f i r s t s t e p removes uranium as v o l a t i l e UF6; the second recovers the LiF - B e 2 by simultaneously v o l a t i l i z i n g these two compo nents from t h e less v o l a t i l e f i s s i o n products. Fortunately, t h e f i s s i o n products so separated are primarily the rare earths, which are the most serious neutron poisons. The UF6 from the fluorinator is accompanied by some v o l a t i l e f i s s i o n product fluorides, ;primarily those of Mo, Te, Ru, . Z r , an& Nb, which are removed by sorption on granular NaF and M@2. Finally, t h e UF6 i s reduced t o UF4 by hydrogen ,and recombined w i t h the decontaminated LiF-BeF2 c a r r i e r i n a single operation. Fission products are removed from t h e plant by discard of N a F and MgF2 sorbents and the s t i l l residue, which is a highly concentrated solution of the rare earths i n LiF. Wastes a r e permanently stored underground.

e f e r t i l e stream, the process consists only of fluorination llowed by decontamination of t h e UF6 on NaF and MgF2 sorbents. It i s only necessary t o 'remove the bred *33U s u f f i c i e n t l y fast t o keep a low concentration i n the blanket, thereby ensuring a low f i s s i o n rate and negligible With respect t o

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poisoning by f i s s i o n products. Discard of the barren f e r t i l e stream a t a slow rate suffices t o keep the f i s s i o n product concentration a t a tolerable level. The chief -conclusions of t h i s study are: (1)that the f l u o r i n a t i o n ~ i s t i l l a t i o nprocess f o r the f u e l stream and the fluorination process f o r the f e r t i l e stream comprise a compact and r e l a t i v e l y slmple system that can be engineered w i t h a normal amount of developmental work, and (2) that integration of the processing plant i n t o t h e reactor f a c i l i t y i s both feasible and economical and the l o g i c a l way t o take advantage of the processing p o s s i b i l i t i e s of a fluid-fueled reactor. The nominal cost of t h i s plant is presented i n t h e following summary of major items:

Process equipment and building space Fuel s a l t inventory . F e r t i l e salt inventory NaK coolant inventory Direct operating cost

$5,302,000 89,500 69,200 40,000 788,000/year

These costs contribute about 0.2 milllkwhr t o t h e f u e l cycle cost when the reactor operates a t an 8 6 plant f a c t o r and c a p i t a l charges are amortized a t l@/year. T h i s cost i s s u f f i c i e n t l y low t o add t o t h e incentive f o r developing t h e molten -salt breeder reactor.

Some of the steps of t h e evaluated process are based on wellestablished technology, whereas others are based on extrapolations of laboratory and small-scale engineering data. Fluoride v o l a t i l i t y and associated UF6 decontamination by sorption are well-known operations, having been demonstrated in a p i l o t plant. However, vacuum d i s t i l l a t i o n and l i q u i d phase reduction of UF6 t o UF have been demonstrated only a t the bench. Certainly, more evelopment of these steps i s required f o r a complete process demonstration. A singularly serious problem is the corrosive nature of the fluorinelnolten s a l t m i x t u r e i n the fluorinator. However, t h i s study shows that t h i s and other inherent processing problems can be solved by proper design and operation of equipment.

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INTRODUCTION i

A reactor concept that has a high p o t e n t i a l f o r economic production of

'nuclear power and simultaneous breeding of f i s s i l e material on the thoriumuranium fuel cycle is t h e molten-salt breeder reactor (MSBR). u t i l i z e s two f l u i d streams.

This reactor

For t h i s study, the stream compositions are:

(1)a f u e l stream consisting of an LiF-BeF2 (69-31 mole $) c a r r i e r that

contains the f i s s i l e component, and ( 2 ) a f e r t i l t stream, which surrounds

71-29 mole $ LiF-ThF4 mixture. I n each stream the lithium is about 99.995 a t . $ 7L i . The above compositions are not unique; other MSBR designs might use d i f f e r e n t compositions. 39 me conthe f u e l stream, consisting of a

figuration of t h e system allows a r e l a t i v e l y high neutron leakage rate f r o m the f u e l stream i n t o the surrounding f e r t i l e stream, where capture

by thorium breeds additional 233U fuel.

The reactor i s operated a t an

average temperature of about 650 C; f i s s i o n energy i s recovered i n external heat exchangers through which the f u e l and f e r t i l e streams are circulated. M

Sustained breeding performance of the MSBR depends on the removal of f i s s i o n and corrosion products from the two f l u i d streams so t h a t p a r a s i t i c neutron capture i s . k e p t t o a tolerable rate.

Portions of the two circu-

l a t i n g streams are continuously removed, processed f o r removal of f i s s i o n products, f o r t i f i e d w i t h makeup f i s s i l e and f e r t i l e materials, and returned t o t h e reactor i n a cyclic operation. A primary consideration of any process f o r recycling reactor f u e l i s that minimal losses of all valuable fuel components be a t t a i n e d without

intolerable c a p i t a l investment t h i s requirement applies t o bo

operating expense.

I n t h e MSBR system,

1 and c a r r i e r components.

As mentioned

rier component is the highly enriched 7Li isotope. The other major constituents, beryllium and thorium, are less expensive; y e t l a r g e losses of these cannot be t o l e r a t e d either because of above, t h e f u e l is 233U,

and o

the adverse e f f e c t on fuel-cycle cost and f u e l conservation.

The process

evaluated here accomplishes the objectives of conservation w h i l e providing f i s s i o n -product 'removal s u f The work reported here

i e n t f o r a successful breeding system. unique i n t h a t it examines a processing

plant integrated d i r e c t l y i n t o the reactor system, which, i n effect, accomplishes on-stream processing. This method obviates t h e cumbersome

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and expensive transfer of highly radioactive material by c a r r i e r shipment; furthermore, common use can be made of services and equipment necessary t o the reactor, thus avoiding the duplication that r e s u l t s i n a separate processing building.

Also, since t h e spent f u e l flows d i r e c t l y i n t o the

processing plant, there is minimal out -of -pile inventory of valuable f u e l components. Another i n t e r e s t i n g feature of t h i s study i s the use of t h e r e l a t i v e l y recent concept of vacuum d i s t i l l a t i o n as a means of purifying t h e c a r r i e r

salt.

A modest vacuum of only 1 t o 2 mm Hg i s required, but the tempera-

ture (about 10CQ째C) i s higher than any normally encountered i n handling molten fluoride salts. The operation was explored first by Kelley3 i n laboratory experiments that indicated LiF-33eF2 decontamination from f i s s i o n products by factors of 102 to 103 The attractiveness of the

process l i e s i n the f a c t that it involves only a physical operation t h a t is e a s i l y controlled and t h a t can be made continuous.

Fission products

can be concentrated i n t h e s t i l l residue (primarily 7L i F ) by a f a c t o r of about 250 by using the decay heat of the f i s s i o n products t o v o l a t i l i z e LIF-BeF2.

This high concentration f a c t o r ensures a low discard rate f o r

the valuable 7L i .

Cyclic operation of the s t i l l was assumed, allowing

the f i s s i o n product concentration t o increase with t i m e .

The corresponding

increase i n the rate of decay heat generation limited the cycle time t o about 68 days.

Although limited experience w i t h the d i s t i l l a t i o n s t e p

indicates that high-nickel alloys are s a t i s f a c t-o r y s t r u c t u r a l materials f o r use a t t h i s r e l a t i v e l y high operating temperature, a more extensive

investigation is needed t o define t h e design limitations. A novel idea has a l s o been studied i n the evaluation of liquid-phase

reduction of UF6 t o UF4 by hydrogen.

I n i t i a l bench-scale experiments 26

have given promising data. The reaction i s carried out by absorbing m6

i n a molten m i x t u r e of LiF-BeF2-UF4 a t about 6oo0c followed by contacting with 5 t o reduce UF6 i n s i t u . This technique avoids the troublesome problem of remotely handling s o l i d s (small UF4 particles) that would be

m e t i f t h e customary gas-phase reduction of m6 w e r e used. Aside from indicating t h e f e a s i b i l i t y and economy of t h e process,

this study a l s o uncovered important design and engineering problems associated w i t h t h e scaleup of laboratory and batchwise operations t o

larger, continuous ones. In this regard, recommendations are presented at the end of the text, along with importantconclusions. The most noteworthy recommendation is that the key operations, vacuumdistillation and continuous fluorination, be given high priority in development. The material that follows is arranged in this sequence: First, a brief description of the MSSRsystem is given to put the study in perspective; second, design criteria and ground rules are stated for-each phase of the study; 'third, iz process flowsheet and a description of eachâ&#x20AC;&#x2DC; unit operation is presented; fpurth and fifth, a descr2ption and pertinent design data for each major componentand the processing cells are listed; sixth, equipment and building-cost,data are presented; seventh, an overall evaluation of the process is given in a set of enumerated conclusions and recommendationsarranged according to plant characteristics and process operations. Six Appendices, giving detailed data and calculations, ar& attached.

TEIEMOLTENSALT BREEDER REACTOR SYSTEM

LF) , i i t

The processing system of this study is designed to meet the requirements of the molten-salt breeder reactor shown in Fig. 1. This is a conceptual design1 of a power reactor capable of producing 1000 Mw (electrical) with a thermal efficiency of 45s. Basically, it consists of a graphite matrix enclosed in a'cylindrical Hastelloy N vessel for ~containment. Graphite occupies about 79 vol $ of the core, fuel salt about 15.~01 6, and fertile salt about 6. The flow passages are such that the fuel and fertile streams.'do not mix. The core ii surrounded radially and axially by a 3.5-ft blanket of LiF-ThF4 mixture, and the ', blanket is in turn surrounded by a 6-in.-thick,graphite reflector. The core is about 8 ft in diameter and 17 ft high; overall, the reactor plus breeder blanket is about 1.6 ft in diameter and 25 ft high. I ,Fission energy is recovered in a battery of external heat exchangers through wh%chthe fuel and fertile streams are continuously circulated. The coolant may,be either a'molten carbonate or fluoride salt mixture which transports the heat to boilers ,for producing steam. Small sidestreams of fuel and fertile fluids are continuously withdrawn from the ,

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circulation loops and routed to a chemical processing cell adjacent to the reactor cell. After being processed for fission product removal and reconstituted with makeup materials, the two fldds-are returned to the reactor via the fuel-makeup system. The pro&sing cycle is selected to give the optimum combination of fuel-cycle cost and breeding gain. The data presented in !C!able 1 are typical for the MSBR and were the 2 bases for this study. Since the reactor concept is undergoing engineering and physics~evaluation, these data represent no fixed design and are subject to change as the studies progress. DESIGN CRITERIA . The following discussion delineates ground rules and arguments for the brticular choice of process and design used in this evaluation. Choices were made on the basis of existing knowledge and data. The study presented â&#x20AC;&#x2DC;judicious

here is expected to verify alternatives. @sic

One basic

consideration

basic assumptions

Considerations

concerns the fuel

or indicate

yield

(the fraction

fissile inventory bred per year) which, for a breeder reactor, is inversely proportionalto the total inventory of the reactor and chemical plant This characteristic is essenti&l to the design of an MSBR pro'systems. cessing plant and suggests close-coupling of the reactor and processing plant to give minimal o&-of-reactor inventory. A fluid-fueled system is readily amenable to this type of operation, and for this evaluation the processing plant is integrated with the reactor plant. This design permits fast, continuous processing, restricted only by the rather stringent, design requirements for fission-product decay-heat removal and corrosion resistance. The integrated plant occupies\cells adjacent to the reactor cell, and all services available to the reactor are available to the chemical plant. These include mechanical equipment, compressed gases, heating and ventilating equipment,

electricity,

etc..

The cost savings

for an integrated

-8

Table 1. Typical Characteristics of a Molten S a l t Breeder Reactor

General 1000

Reactor power, Mw ( e l e c t r i c a l ) Thermodynamic efficiency, $ Reactor geometry Core diameter, f t Core height, f t Blanket thickness, f t Moderator Volume fraction of moderator i n core Volume fraction of f u e l in core Volume f r a c t i o n of f e r t i l e stream i n core Reactor containment vessel Fraction of fissions i n f u e l stream Plant f a c t o r Breeding r a t i o

Cylinder

8 17 3.5

Graphite 0.79 0.15 0.06 INOR -8 0.972 0.8 -1.08

Fuel Stream composition, mole 4 LW (99.995 at. k 7 ~ 5 ) kF2 U F (~f i s s i l e ) Inventory a t equilibrium volume, f t 3 L-, ke;

68.5a 3 1 . ~ 0.31

671 19,530

16, loo 736 11.1

Other U (asTUF4), k g Cycle t i m e , days Power, MW (thermal) Liquidus temperature, OC Density (calculated) p (g/cm3) = 2.191 0.0004 t ("C)

58.1 58

2160 -500

f o r 525 i t i 1200째C

F e r t i l e Stream composition, mole $ (99.995 at. k 7 ~ i )

71 29 0.012b 0.022b

Inventory a t equilibrium volume, f t 3 Lfl, 43 Th (as =4), 43 2 3 3 ~(as UF~), kg 2 3 3 ~ a( a s ~ a ~ 4 kg ), Cycle time, days Power, Mw (thermal) Liquidus temperature, OC Density ( I c d a t e d ) p (g/cmF = 4.993 0.000775 t ("c) for6565 s t s

1783 '38,760 iL,290 60 110 22 62

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moot

Steam Conditions Pressure, psia Temperature, 9 Condenser pressure, in. EIg abs %sic

composition of c a r r i e r salt is 69-31 mole $ ( L F F - B ~ F ~ ) .

%quilibrium composition f o r t h i s cycle.

3515 1000

1.5

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f a c i l i t y are immediately apparent when one considers the large amount of equipment and f a c i l i t y dup A f u r t h e r basic consideration i s that there w i l l be no large

extrapolation of technology i n the process design.

Accordingly, the process is based on t r e a t i n g the molten salt by fluorination and distillation, with the supporting operations of from beds of pelletized N@,

uF6 sorption on and desorption

followed by reduction of t h e UF6 t o IF4.

large amount of data. is available f o r each s t e p except f o r the distilla?

t i o n and reduction operations, both of which have been demonstrated i n the laboratory. 3126 However, t h i s study does assume the necessary engineering extrapolations t o convert from the current batchwise operations t o continuous operations Process f o r the F e r t i l e Stream The two streams of the breeder reactor require d i f f e r e n t processing

rates and must be treated separately t o prevent cross contamination.

!I!he

first s t e p i n process f o r the f e r t i l e stream consists only of continuous fluorination, which removes t h e bred uranium as the v o l a t i l e hexafluoride. ,/

N o other treatment is needed i f t h i s s t e p i s designed to maintain a low

uranium concentration.

To accomplish this, the stream is required t o go

through the processing plant on a r e l a t i v e l y s h o r t cycle, f o r example, once every 20 t o 50 days.

cycle time f o r t h i s study is 22 days.

The

f i s s i o n 'rate i n the blanke

10% and the f i s s i o n products are kept a t a t o l e r a b l e l e v e l by periodic discard of barren LiF-ThF4 salt. A 30-year discard cycle suffices.

I n the second step, the v o l a t i l i z e d UF6 i s sorbed on NaF beds, desorbed, and f i n a l l y wught i n cold traps. f o r the Fuel Stream The f u e l stream of the reactor is processed by fluorination and >

vacuum d i s t i l l a t i o n t o recover both uranium and c a r r i e r salt s u f f i c i e n t l y decontaminated of f i s s i o n products.

A cycle time of 4-0 t o '70 days is

required t o maintain the fission-product concentration a t a low enough l e v e l f o r a t t r a c t i v e breeding performance. t

The calculations of t h i s study

- 10 are f o r a 58ilay cycle.

The uF6 is recovered by NaF sorption and cold

traps, j u s t as f o r the f e r t i l e stream.

Decontaminated f u e l and c a r r i e r

are recombined i n a reduction step t h a t converts W6 t o UF4 and f u r t h e r -

p u r i f i e s by reducing corrosion products (iron, nickel, chromium) t o t h e i r

metallic s t a t e s .

Makeup f u e l and c a r r i e r are added a t t h i s point, and

the stream is returned t o the reactor.

The time that the f u e l stream spends i n the processing plant i s kept as s h o r t as practicable t o minimize out-of-reactor inventory.

Waste Storage The chemical plant provides i t s own storage system f o r proeess

wastes.

Incidental wastes, such as s l i g h t l y contaminated aqueous sol&ions

and f l u s h salts, are assumed t o be handled by the reactor waste system, thus such f a c i l i t i e s are not duplicated f o r t h e processing plant.

Separate

storage is provided f o r fuel-and fertile-stream wastes, which are primarily LiF plus fissionproducts, and LiF-ThF4, respectively.

The f a c i l i t i e s are

designed f o r a 30-year capacity and are located underground a s h o r t distance from the chemical processing area.

The fuel-stream process also produces a less radioactive waste than t h e LiF-fission-product m i x t u r e . This w a s t e i s i n t h e form of p e l l e t s of sodium fluoride and magnesium fluoride p e l l e t s used f o r decontaminating uranium hexafluoride. Interim storage of 5-year duration i s provided f o r these solids. . Fission-product decay heat i s removed either by forced a i r o r natural convection, as required by the heat load. Operating Policy

C e r t a i n ground rules consistent w i t h convenient and safe operation w e r e adopted f o r t h i s study.

Maintenance operations are f a c i l i t a t e d by

assigning unit-process steps t o either a high- o r a low-radiation l e v e l cell.

Operating and maintenance personnel, who are not required on a f u l l -

t i m e b s i s , are t o be shared w i t h reactor operation.

No water o r aqueous

solutions are t o be admitted t o the process c e l l s ; f l u i d s required f o r heat transport w i l l be either air o r sodium-potassium e u t e c t i c (NaK).

A barren

L I .

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fluoride salt, f o r example, an NaF-KF mixture, would be an acceptable substitute. Process D a t a The primary concern 'of a processing cycle f o r short-cooled fluoride mixtures i s i n dissipating f i s s i o n product decay energy s o t h a t process operations can be controlled.

A maximum of about

6.59 of the totftl energy

of t h e system i s associated w i t h be.ta and gamma energy i n the f i s s i o n products; t h i s amounts t o about 140 Mw (thermal) i n t h e MSBR f u e l stream. Most of t h i s energy i s emitted quickly, decreasing about 82Q i n 1 h r and

959 i n 1 day.

The data f o r t h i s study were calculated f o r the reactor

described i n Table 1, using the PHOEBE Code,

4 which computes gross f i s s i o n -

product decay energy as a function of exposure and time after discharge from the reactor.

The data are presented graphically i n Figs. 2a-b and

3 - b f o r f u e l and f e r t i l e streams, respectively.

me graphs give an upper l i m i t f o r heat generation because t h e calculations do not account f o r possible intermittent reactor operation a t t r i b u t a b l e t o t h e 8@

plant factor.

I n addition, t h e graphs include

decay energy associated with gaseous products t h a t are sparged i n the reactor c i r c u l a t i n g loop and with those f i s s i o n products that m i g h t ,deposit on surfaces throughout the system. 0

It is d i f f i c u l t t o separate

t h i s energy from gross energy u n t i l more i s known about the behtivior of

f i s s i o n products i n molten fluorides. A process flowsheet showing material balances f o r the f u e l and f e r t l e

streams i n the processing

s included i n Appendix F.

ION OF PROCESS The processing f a c i l i t y

,have t h e capability of removing the major

portion of f i s s i o n products from t h e molten f u e l s a l t and ,

after necessary reconstitution with 233, purified salt t o t h e f u e l sys and carrier salts. As t o blanket-salt processing, t h e f a c i l i t y must achieve recovery of the major portion of the bred uranium f o r recycle t o t h e f u e l

stream o r f o r s a l e . f

turning the

These goals can be met w i t h present technology o r with

processes that can be reasonably extrapolated from current technology.

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rn mrl

dcI%-

, jab

ORNL DWG 6E2988RI

t \

w I

TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. 2b. Fission-Product Decay Heat i n MSBR Fuel S t r e w . Fuelstream volume, 671 ft3; power, 2160 Mw (thermal); cycle time, 58 days.

ORNL DWG 652989R1

-r

I-

$ I

, TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. 3a. Fission -Product Decay Heat i n ElsBR F e r t i l e Stream. F e r t i l e s t r e m volume, 1783 ft3; power, 62 Mw (thermal); .cycle time, 22 days.

ORNL DWG 6529%R1

vl I

TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. 3b. Fission-Product Decay 'Heat i n MSBR F e r t i l e Stream. F e r t i l e stream volume, 1783,ft3; power, 62 Mw (thermal); cycle time, 22 days.

- 16 &4 Summary of the Process Flowsheet

I n Fig. 4 it can be seen that fuel-stream processing i s achieved by f i r s t removing t h e uranium ( a s 23%l?6)

and v o l a t i l e f i s s i o n products from

a side-stream of the molten s a l t by fluorination.

’

Then the c a r r i e r s a l t

is vacuumdistilled from the residual f i s s i o n products.

N e x t , the

UF6,

further purified by a sorptiondesorption process based on the use of NaF

pellets, i s dissolved i n t h e c a r r i e r salt.

Finally, the uF6 i s reduced

t o UF4 by hydrogen, thereby reconstituting the f u e l salt. Blanket s a l t i s processed concurrently with the f u e l salt by fluorinating a side-stream.

The uF6 gas i s separated from t h e v o l a t i l e

fission-product fluorides by sorption and desorption, using beds of NaF pellets as mentioned above.

and a portion of the 233,6

The blanket s a l t needs no f u r t h e r purification,

i s sent t o t h e f u e l stream by dissolving it i n

the carrier s a l t and reducing it with hydr‘ogen.

The excess over t h a t

needed t o r e f u e l the reactor i s sold. The two chemical reactions (fluorination and reduction) i n the process are simple, fast, and quantitative.

The other interactions are

physical and require only heat and mass transport; however, i n the case of d i s t i l l a t i o n , a rather high temperature (about 1000°C) i s involved. The salt i s extremely stable a t any temperature anticipated i n this

process, and other physical properties, primarily vapor pressure and s o l u b i l i t y f o r f i s s i o n products, are i n accord with process requirements. I

Fluorination *

As noted above, uranium and v o l a t i l e f i s s i o n products are removed from both streams by fluorination.

A batchwise, molten-salt fluoride-

v o l a t i l i t y process f o r recovering uranium has been under development f o r several years. OFUIL.’~’’~

Currently t h e process is i n the p i l o t plant phase a t

One of the steps i n t h i s process i s batch fluorination of the

molten fluorides t o remove uranium hexafluoride. i n these streams i s i n the form of UF4.

As noted before, uranium

T h i s uranium and some of the

f i s s i o n products a r e converted by fluorine t o higher-valent,

volatile

fluorides which leave t h e s a l t and go t o t h e uranium hexafluoride

c \

c"'

. purification system.

-18-

The reaction f o r producing m6 in the present batch

ii”

processing i s the same as that f o r the proposed continuous system and may be represented as follows:

UF4 + F2 4UF6 Certain fission-product fluorides are a l s o v o l a t i l e . ones are:

Ru, Nb, Cs, Mo, and Te.

The principal

Zirconium i s v o l a t i l i z e d t o a lesser

degree. Continuous processing of the MSBR f u e l s a l t can be best achieved i f t h e fluorination process is made continuous, preferably with counter-

current contacting of the molten salt and fluorine.

Such a process has

not been developed, and the present v o l a t i l i t y process a t ORNL i s e n t i r e l y

a batch process.

Therefore, development work w i l l be necessary t o provide

f o r continuous fluorination.

-Rate data f o r molten-salt fluorination are limited and conflicting,

although the reported rates have been s u f f i c i e n t f o r the batch process.

Mailen’s &tal9 on the fluorination of f a l l i n g droplets of molten s a l t support the view that the reaction i t s e l f i s very fast; whereas, the slower rates that r e s u l t from sparging a pool of molten s a l t with fluorine 22 (Cathers e t al., 2o Pitt,21 and Moncrief ) can be explained by assuming a

mass-transfer-controlling rate mechanism due t o i n e f f i c i e n t mixing of the gas and l i q u i d phases.

Contact times (fluorine and molten s a l t ) of 1 t o

2 hr have been shown t o be adequate f o r uranium removal down t o 10 ppm i n

such batch tests.

The countercurrent, continuous operation, envisioned

here, would probably give better contact.

It is d i f f i c u l t t o predict the

fluorine u t i l i z a t i o n i n a continuous fluorinator; however, it should be I

better than that f o r batch fluorination, which suggests that a u t i l i z a t i o n of 33-1/$

w i l l p r o b b l y be achieved.

A r e l a t i v e l y low mass flow rate of gas must be maintained i n t h e

fluorinator t o prevent salt entraznment i n the off-gas.

The highest mass

f l o w rate that has been used successfully Without entrainment is 0.28 slpm . per in.2 of f l u o r i n a t o r cross section. 23

Experience i n the V o l a t i l i t y P i l o t Plant a t ORNL showed that corrosion on bare metal walls ( L a i d k e l ) i n the f l u o r i n a t o r i s r e l a t i v e l y high.

’Kessie e t

developed a technique f o r keeping a frozen

( J *

- 19 protective layer of salt on the m e t a l w a l l , and we adopted t h e i r approach f o r t h i s study.

This frozen wall of salt, which i s - k e p t s o by a high

cooling rate on the m e t a l wall, prevents gross corrosion. heating keeps the rest of the salt molten.

Internal

For the f u e l stream, f i s s i o n -

product heat i s s u f f i c i e n t f o r t h i s purpose, but f o r the blanket stream

it may be necessary t o supplement decay heat by using suspended electrodes 2 f o r resistance heating. Since a maximum of 1.5 kw of heat per f t of wall a r e a can be removed from such a system, t h e f u e l stream will have t o be delayed i n a cooling tank u n t i l the heat generation due t o decay heat is a t EL s u f f i c i e n t l y l o w level. As t o the blanket salt, decay heating M l l be insignificant

The continuous fluorinator i s a t a l l column i n t o which molten salt

is fed a t the top and flows t o t h e bottom; fluorine is introduced a t t h e ' bottom and passes t o the top, accompanied by the v o l a t i l e fluorides (Fig. 5).

An expanded deentrainment section i s added t o the top, and the body of the column i s jacketed w i t h a coolant t o maintain the frozen s a l t wall.

gravity l e g is used i n t h e molten s a l t o u t l e t t o hold a constant s a l t l e v e l i n the column. Purification of Uranium Hexafluoride by Sorption and Desorption The UF6 from both streams is purified i n t h e same way before being

returned t o the f u e l stream.

Since the

contains v o l a t i l e f i s s i o n pro steps.

llnese are batch steps,

that leaves the fluorinator

it is purified by a series of sorption t h e process is made. continuous by

using parallel beds a l t e r n a t e 1 i n a NaF absorption system where %e

The first separation o

gas

stream passes through fixed of N ~ Fp e i l e t s . This s y s t e m consists of b O 0 C and one a t about 100째C. I n the two d i s t i n c t zones, one he higher -temperature zone, corrosion products, entrained salt, and niobium removed from the fluorinator off-gas w h i l e and zirconium are irrevers t s pass through (Fig. 6). In t h e second the uF6 and other f i s s i o n 0

NaF zone (lOO째C),

UF6 and some of the molybdenum are

while the remaining f i s s i o n products pass through.

ul?6 has been separated from a l l 7

held up by sorption

A t t h i s point, the

the f i s s i o n products except molybdenum.

- 20 ORNL DWG 65-3037

râ&#x20AC;&#x2122;

Fig. 5 . Continuous Fluorination. NaK coolant, flowing through the jacket, freezes a layer of salt on the inner surface of the column, thus protecting the Alloy 79 -4 from corrosive a t t a c k by the molten -salt-f luorine mixture.

"41

- 22 Molybdenum fluoride is removed from the

uF6 by i s o l a t i n g the 100°C

NaF zone, desorbing the uF6 and molybdenum fluoride (by m i s i n g the

temperature t o >150°C and passing fluorine through the bed), and passing which is held a t about 150°C.

desorbed gas through a fixed bed of M@*,

The MgF2 sorbs the molybdenum but allows the p u r i f i e d m6 t o pass through i n t o the cold traps. The N a F p e l l e t s used i n the high-temperature zone must be replenished

periodically since they accumulate f i s s i o n products and corrosion products. This discard constitutes one of t h e waste streams. The two NaF sorption zones may be integrated i n t o a single unit, one zone on t h e other, and, as

Nal? i s discharged from the lower zone, it can be replenished by N a F from the low-temperature zone, whi,ch i s i n t u r n fed with fresh N a F p e l l e t s .

Such a system has been used e f f e c t i v e l y i n p i l o t plant operations, and a similar system would be desirable f o r the MSBR processing f a c i l i t y . 17,1€3 I n our concept, the system has a movable bed of N a F pellets, and, after eachcsorption cycle, some of t h e lower NaF i s mechanically ejected t o waste.

Annular design with a i r cooling would probably be necessary t o

allow removal of f i s s i o n product heat (Fig. 7) As the

UF6 leaves t h e last sorption t r a p i t - m u s t be collected and

ultimately used as feed f o r the fuel-reconstitution step. by collecting the

uF6 i n cold

This i s done

Two cold t r a p s a r e connected i n s e r i e s . The first, o r primary trap, i s operated a t about -kO°C, and the second, o r backup trap, is operated a t about &O°C. The principal design consideration i s t h e heat t r a n s f e r rate. Conventional designs are availtraps.

able f o r such u n i t s i n which there are i n t e r n a l cooling f i n s fop collecting

u F 6 y and Calrcd heaters f o r vaporizing the uF6 f o r removal (Fig. 8).

Uranium hexafluoride from the heated cold t r a p is fed d i r e c t l y t o the reduction process.

This c a l l s f o r a t least three primary cold t r a p s f o r

continuous operation:

one f o r collecting

UF6, one f o r feeding uF6, and

one i n t r a n s i t i o n between these two functions. A fixed bed of N a F a t ambient temperature very e f f e c t i v e l y removes

t r a c e amounts of uJ?6 from gas streams.

Such beds a r e used as backup

t r a p s i n t h e fluorine exhaust from the cold t r a p s and i n other process streams that might'contain

m6. Uranium is recovered from such t r a p s by

using t h e NaF as charge material f o r t h e main absorption beds.

w *

- 23 9

L f

ORNL- LR- DWG 5045lR-3

r N a F CHARGING CHUTE

1Vz-in. NPS, SCHED-40(AIR COOLING AND THERMOCOUPLES)

5-in.NPS, SCHED-40 INCONEL

INCONEL-X

PISTON

5-in. NPS, SCHED-80 INCONEL

HYDRAULIC CYLINDER

Fig. 7. Movable-Bed Temperature-Zoned Absorber. When the lower zon.e of t h e bed becomes loaded withâ&#x20AC;&#x2122;fission products, t h e hydraulic cylinder operates t h e piston t o discharge t h a t portion of the bed i n t o the waste c a r r i e r . Fresh N a F i s added a t the top. This apparatus has alreadv been t e s t e d i n t h e ORNL p i l o t plant.

. .

...

.~ .

. . . - . . . . ._.I..

"... .

. " . ...

~ - - .. . _....I__ ~

ORNL-LR-DWG 19091 R-l

OUTLET

-r I

REFRIGERANT TUBES (4)

5-in:SPS

INLE

INLET END HEATER

Fig. 8. Cold Trap f o r uF6 Collection. 'successfully used i n t h e ORNL p i l o t plant.

COPPER PIPE

This design has already been

- 25 <

Since there is excess fluorine i n the f l u o r i n a t o r off-gas (33-1/$ u t i l i z a t i o n ) , ' fluorine is recycled.

This recycle contains some f i s s i o n

products, so it is necessary t o remove a side-stream (le) t o prevent t h e i r buildup.

Fresh fluorine is used f o r t h e desorption step, f o r a l l

processing i n the blanket section, and f o r fluorine makeup. Vacuum D i s t i l l a t i o n The vacuum d i s t i l l a t i o n s t e p applies t o the f u e l stream only and is used t o separak t h e c a r r i e r s a l t from the f i s s i o n products after the uranium i s removed by fluorination (see above).

The LiF and BeF2

v o l a t i l i z e , leaving f i s s i o n products i n the s t i l l bottoms. consists l a r g e l y of r a r e e a r t h fluorides.

This residue

If t h e r e l a t i v e v o l a t i l i t i e s

of the f i s s i o n products, compared w i t h the v o l a t i l i t i e s of LiF and BeF2, a r e low, then a good separation can be achieved i n a single-step d i s t i l l a t i o n without r e c t i f i c a t i o n . A t t h e average operating temperature (about 650'~) of the MSBR itself, the s o l u b i l i t y of rareearth fluorides i n f u e l salt is only a few.mole

the s o l u b i l i t y i n LiF alone i s about 50 mole

percent; however, a t 1000째C,

f o r the more insoluble compounds, f o r example, LaF3, PrF3, and CeF

Other

rare e a r t h (HE) fluorides have even higher s o l u b i l i t i e s a t t h i s temperature. 5 Physics calculations on the 5 8 a y - c y c l e MSBR indieate that a t equilibrium -the molar r a t i o of Lil?:(RE)F

i n the f u e l i s about 1400, a number considerably greater than the 1:l r a t i o permitted by the s o l u b i l i t y l i m i t a t 1000째C.

It i s therefore apparent that, based. on s o l u b i l i t y data alone,

d i s t i l l a t i o n a t about 1000째C can t o l e r a t e an extremely large r a r e e a r t h concentration f a c t o r before d i s t i l l a t i o n u n i t is concerne configuration that w i l l permi f a c t o r and, a t t h e same t

e c i p i t a t i o n occurs.

The design of the

primarily with determining t h e appropriate large! f i s s i o n -product concentration

de adequate heat-removal capability f o r

the short -cooled fuel.

The s t i l l design de charged i n i t i a l l y with 4 f

f o r t h i s study is shown i n Fig. 9.

It i s

t h a s the same isotopic composition

as t h a t i n the reactor fuel; t h i s volume f i l l s the tubes t o a depth of about an inch above their tops. The pressure above the LiF pool is reduced

- 26 -

ORNL DWG 65-180212 34" 1 D -I

AIR COOLANT

Li F-BeF2-q PRODUCT

Ir)

NaK 817 TUBES1/2" x 16 GAUGE

1/2" INOR-8 TUBE SHEETFUFH-LWLTL

UNT

LiF WASTE DRAIN L

30" D I-

Fig. 9. Vacuum S t i l l f o r MSBR Fuel. Barren fuel-carrier flows continuously i n t o t h e s t i l l , w h i c h is held a t about 1000째C and 1 ma Hg. LW-BeF2 d i s t i l l a t e is removed a t t h e same r a t e t h a t s a l t enters, thus keeping t h e volume constant. Most of the f i s s i o n products accumulate i n the s t i l l bottoms. The contents a r e drained t o waste storage when t h e heat generation rate reaches a prescribed l i m i t . This concept o f ' t h e vacuum s t i l l has not been tested.

- 27 t o 1 t o 2 mm of Hg by evacuating the product receiver (see Fig. D-3 i n Appendix D), ‘and the temperature i s adjusted and held a t about 1000°C. Fluorinated f u e l salt i s continuously a W i t t e d t o the LiF pool i n the s t i l l , and d i s t i l l a t i o n is allowed t o proceed a t the same rate as the i n l e t f u e l rate s o that there is no net volume change. The operating principle is t o allow the rare-earth f i s s i o n products,

which have a much lower vapor pressure than e i t h e r LFF o r BeF2, t o concentrate i n the s t i l l .

The l i q u i d pool i n the s t i l l reaches an equilibrium

concentration i n LiF and BeF2, and t h e two components then d i s t i l l a t t h e 3

-c

rate a t which each is entering the s t i l l . The condenser i s a conical region located j u s t above the evaporating surface; it i s kept a t about

850Oc by forced convection of heated air.

D i s t i l l a t e is collected i n a

.circular trough and drained t o a product receiver.

The s t i l l i s operated

i n t h i s way u n t i l the heat-generation rate due t o fission-product decay reaches the heat-removal capabili&y of the NaK cooling system.

This

67 days of continuous operation, a t which time the f i s s i o n product concentration i n the s t i l l is about 14 mole $, a value considerably less than the approximately 50 mole $ s o l u b i l i t y l i m i t point occurs a f t e r about

a t 1000°C. A t t h i s t i m e , the contents of t h e s t i l l are drained t o a permanent w a s t e receiver, and the cycle i s repeated. These calculations are conservative since they a r e based upon gross fission-product heat release and do not subtract the e f f e c t of those f i s s i o n products removed

’

o r deposited i n t h e reactor before chemical processing. The a t t r a c t i v e feature carrying out the d i s t i l l a t i o n i n t h i s way

i s that it minimizes the vo of expensive LIF ‘relegated t o waste. Since d i s t i l l a t i o n i s carried out a f t e r fluorination, less than O.l$ of the uranium removed from t h e reactor should e n t e r the still, and a t 1000°C he vapor pressure of UF

4 is favorable t o the recovery of a s i g n i f i c a n t

portion of t h i s fraction, fluoride losses should be B ~ aFt IOOOOC ~ is a b u t

w i l l not e f f e c t decontam pounds has a greater vap

&i -

t h e o v e r a l l uranium loss.

Beryllium

icant because the vapor pressure of

t of LiF. D i s t i l l a t i o n probably sF and RbF; each of these comeither BeF2 o r LiF.

Because

r a r e e a r t h s are concentrating i n t h e s t i l l as a function of time, t h e i r decontamination f a c t o r i n the product w i l l decrease with time.

It i s not

possible w i t h e x i s t i n g data t o assess the magnitude of t h i s e f f e c t .

- 28 -

Detailed calculations f o r t h e s t i l l design are given i n Appendices C

and D.

Vapor pressure data f o r principal components of the f u e l stream

are included i n Appendix F. D i s t i l l a t e from t h e s t i l l is collected i n an evacuated tank operating

a t s t i l l pressure. When filled, the receiver is isolated from the s t i l l , and the LIF-BeF2 mixture i s transferred by gravity flow o r pressurization t o the reduction and fuelaakeup operations. An i n - c e l l waste receiver i s provided f o r the i n S t i a l cooling of the

s t i l l residue before transfer t o t h e underground waste-storage f a c i l i t y . The tank has a 4-ft 3 volume, allowing a one-cycle delay (about 67 days)

inside the c e l l w h e r e heat i s conveniently removed by the circulating NaK coolant.

During this delay, the heat generation rate decreases from

6, imposing less stringent design requirements

3.2 x lo7 Btu/hr t o 6.8 x 10

i n the permanent waste receiver.

The interim receiver is a s h e l l a n d -

tube type, similar t o t h e s t i l l ; however, no condensing surface o r provisions f o r air cooling are needed. Reduction of U r a n i u m Hexafluoride and Reconstitution of the Fuel The.combined W6StreamsJ that form the f u e l salt and the f e r t i l e

salt, are reduced t o UF4, and only a s u f f i c i e n t amount t o maintain c r i t i c a l i t y i s returned t o the rea-ctor.

The excess 233U from the f e r t i l e

stream is sold.- The usual m e t h o d f o r t h i s reduction has been by reaction w i t h the excess

i n an %-F2

flame:

This reduction is carried out i n a tall column i n t o which

introduced i n t o an %-F2 product.

UF6 and

H2 are

flame, and dry UF4 powder i s collected as the

However, according t o our proposal, a more convenient method

f o r preparing UF4 f o r the MSBR is by reducing the UF6 t o UF4 w i t h H2 after the

UF6 i s dissolved i n

the molten salt.

There are some experimental

data indicating the f e a s i b i l i t y of such a process; however, the kinetics of t h e absorption and of the reduction must be further investigated. 26 It i s possible t h a t t h i s two-step process could be carried out continuously

-. i

- 29 -

i n a single column i n which the molten salt flows upward, the UF6 is introduced and dissolved i n the bottom of the column, and t h e H2 i s introduced a t an intermediate point t o reduce the

$6 (Fig.

10).

Some

of the reconstituted s a l t has t o be recycled t o the column t o provide

enough dissolved uranium f o r proper

uF6 absorption.

O f f a s Processing

Most of the off-gas from t h e processing plant comes from the continuous fluorinators; smaller amounts are formed i n various other processing vessels. The gases a r e processed t o prevent t h e release of any contained f i s s i o n products t o the atmosphere.

Excess fluorine used

i n the fluorinators i s recycled through a surge chamber by a positive displacement pump, and a s m a l l side-stream of the recycling fluorine is sent through a caustic scrubber t o prevent gross buildup of f i s s i o n products.

Each of the processing vessels and holdup tanks have off-gas

lines,which lead t o t h e scrubber f o r removing HF, .fluorine, and v o l a t i l e

f i s s i o n products. The scrubber operates as a continuous, countercurrent, packed bed with recirculating aqueous KOH.

, s e n t t o waste, and

A small side-stream of KOH solution i s

the scrubber o f f - 9 s is contacted w i t h steam t o

A f i l t e r removes the hydro-

hydrolyze f i s s i o n products such as tellurium.

. -.

lyzed products. The nollcondensable f i s s i o n products are sent t o the o f f gas f a c i l i t y f o r gases generated by the reactor. \

Waste Storage i

Four waste streams requiring storage leave the processing f a c i l i t y :

( 1)Na3? and MgF2 sorbent f r

UF6 purification system, (2) aqueous

) molten-salt residue from the d i s t i l l a waste from the KOH scrubbe t i o n unit, and (4) molten salt from f e r t i l e stream discard. The aqueous waste stream is small, and it is assumed t h a t adequate capacity e x i s t s i n ste. The other three wastes are stored the system f o r storing rea

i n separate underground f a c i l i t i e s adjacent t o the processing c e l l s . Since the values i n the waste from the f e r t i l e stream

- 7Li,

thorium,

. . .

- -

- --

- 30 ORNL DWG. 65-3036

Fig. 10. Continuous Reduction Column. Barren salt and UF6 e n t e r the bottom of the column, which contains circulating L i F -13eF.2 -UF4. The UF6 dissolves i n the salt, aided by the presence of W4, and moves up the column where it is reduced by hydrogen. Reconstituted f u e l i s taken off t h e top of the column and sent t o t h e reactor core. Preliminary data indicate t h a t t h i s design is promising.

- 31 and 233U

- w i l l be worth

recovering a t some future time, some very

t e n t a t i v e ideas about how they may be recovered are presented a t t h e end of this section. NaJ? and MgF, P e l l e t s i

Spent N a F and MgF2 pellets, which r e t a i n the v o l a t i l e f i s s i o n and corrosion product fluorides from-the

m6 gas stream, are stored i n 10-in.-

diam by 8-ft-high stainless s t e e l cylinders i n a concrete vault adjacent t o the s t i l l - r e s i d u e waste tank.

(See Dwg. No. 58080~i n Appendix F.)

The cylinders are loaded inside the processing c e l l and transferred t o the underground area a t approximately gO-day intervals.

The vault is designed

t o contain a 5-year collection of cylinders (160 cylinders a t 106 p h n t f a c t o r ) ; after 5 years, the older cylinders are removed and transferred t o a permanent underground storage s i t e .

The integrated heat generation

when t h e vault contains 160 cylinders is about 1.73 x 10 Btu/hr.

Forced

circulation 'of about 11,300 scfm of a i r a t a temperature rise of l 2 5 O C is used t o remove t h i s energy.

The containers are constructed with a hollow

core, allowing coolant t o pass through t h e aylinder as w e l l as over the outside

Aqueous Waste from Off-s

Scrubber

This waste, f i s s i o n products i n a strong solution of KOH, w i l l be stored along with other aqueous wastes from the reactor systeq and represents an insignific waste

tribtrtion t o the total amount of aqueous

Fuel Stream Waste Residue from the vacuum stlll is stored i n bulk i n a f a c i l i t y similar

t o one evaluated previously by Carter and Ruch.14

A single, large tank

equipped with adequate cooling tubes and adequate f o r a 30-year accumulat i o n of waste is provide t h e expected lifetime of

ar capacity was chosen since that i s A f t e r f i l l i n g , the salt m i g h t

remain i n the tank f o r additional decay o r be disposed of by whatever method is currently acceptable.

- 32

Decay heat i s removed by forced a i r convection.

The heat load (Btu/hr)

continually increases over t h e f i l l i n g period but decreases r a t h e r sharply when no further additions are made (Fig. E-1, Appendix E ) .

The t i m e

behavior of the integrated s p e c i f i c heat generation rate (Btu hr'l

ft-3)

f o r a 5-year collection period i s shown i n Fig. 11. This is a smoothed curve f o r 4-ft 3 additions every 67.4 days, followed by an extended decay period during which no waste is added t o the tank.

The upper portion of

the curve m i g h t be extrapolated with l i t t l e e r r o r t o accommodate longer

Figure 11 shows the s p e c i f i c heat generation f o r the

f i l l i n g periods.

case of no d i l u t i o n with i n e r t salt; however, during the i n i t i a l stages of f i l l i n g , it is necessary t o add an i n e r t diluent, f o r example, NaJ?-KF e u t e c t i c m i x t u r e , t o lower the s p e c i f i c rate t o a tolerable value. It was calculated that 264 f t 3 of diluent is required f o r t h e 520 f t3 of LiF-fission product residue t o be collected over the 30-year period. f i s s i o n products being collected exhibit the decay behavior shown i n

The

Fig. 2, which i s representative of gross f i s s i o n products and does not account f o r those that have been removed by processing o r other mechanisms. F e r t i l e S t r e a m Waste The fertile,-stream discard is a l s o stored i n a large underground tank, adequate f o r 30 years of waste collection. and 13.5 f t high.

The tank is 13.5 f t i n diameter

Since uranium is removed from the blanket on a rather

f a s t cycle, the f i s s i o n rate i n the blanket i s low, making t h e waste a c t i v i t y several orders of magnitude less than that of the f u e l waste.

Cooling is provided by n a t u r a l a i r convection around the tank and through cooling tubes.

A l l metal surfaces are expected t o be coated with a layer

of frozen salt that w i l l furnish excellent corrosion protection. The integrated heat production rate due t o f i s s i o n products f o r t h e

30-year period is 5.91 x 10 Btu/hr. Since this energy i s associated with 1783 f t3 of LiF-ThF4 m i x t u r e , the moderate s p e c i f i c rate of 33.1 Btu hr-I f t - 3 presents no design problems.

When f i r s t removed from the

reactor, the heat production rate of the waste is about 1600 Btu hr'l but t h i s value decreases by a f a c t o r of 10 i n about 4 days.

ft-3

TIME AFTER DISCHARGE FROM REACTOR (days) I

Fig. 11. Specific Heat Generation Rate of Fuel-Stream Fission Products i n Waste Tar&. Waste i s accumulated i n 4 f't3 batches every 67.4 days. It is then held i n t h e processing c e l l another 67.4 days f o r further cooling before draining t o the waste tank. The reactor operates on a 58-day cycle a t 2160 Mw (thermal).

- 34 ,

Cooling System f o r t h e Waste I

A s mentioned above, the waste-storage system i s designed f o r cooling

by forced air draft.

A i r is easy t o handle, compatible with construction

materials f o r extremely long times, and presents a minimum hazard i n case of contamination by t h e waste.

Blowers, capable of supplying 76,000 scfm

a t a pressure drop of about 30 in. water, are located upstream from t h e waste vaults. The a i r is forced through the vaults and cooling tubes i n the waste tank and is exhausted t o the atmosphere via a t a l l s t a c k , which

a l s o disposes of gases exhausted from the reactor system.

The exit duct

contains the necessary radiation-monitoring instruments and absolute f i l t e r s f o r removing p a r t i c l e s .

Possible U l t i m a t e Treatment of Waste from F e r t i l e Stream Since a 30-year accumulation of waste from the f e r t i l e stream w i l l contain recoverable values ( 7Li, 233U,

and thorium), not too highly

contaminated with f i s s i o n products, it m y be worthwhile t o consider a recovery system before relegating t h i s waste t o permanent burial. Any s i g n i f i c a n t uranium value would probably be recovered by fluorination, but t h e recovery of thorium and lithium requires further process development.

A p o t e n t i a l method f o r Li-Th separation i s the incompletely

investigated BF d i s s o l u t i o n process, l5 based on the principle of leaching LiF f r o m ThF4 and rareearth f i s s i o n products with anhydrous hydrofluoric acid.

This process, however, leaves the thorium contaminated with f i s s i o n

products, making it necessary t o r e s o r t t o an aqueous system (solvent extraction by t h e Thorex process) o r t o develop a thorium recovery process that u t i l i z e s fluoride chemistry.

From a purely economic viewpoint,

thorium would be retained in t h e waste tank u n t i l i t s recovery became more economical than mining new thorium.

On a 30-year discard cycle, an e n t i r e fertile-stream inventory of TW4 and LiF w i l l accumulate i n the waste tank.

w i l l contain 2 I

In addition, the waste

3Y 233~a, ~ and f i s s i o n products i n amounts that depend on

the breeding ratio, efficiency of the fluorination step, and the blanket power. The uranium loss is based upon a 9 6 efficiency i n fluorination,

a value believed t o be conservative.

The l a r g e s t loss of fissionable

- 35 -

material, however, is through protactinium, which we assume t o be nonv o l a t i l e as the fluoride and which consequently is discarded in d i r e c t .ratio t o i t s concentration.

The amount of f i s s i o n products is calculated

f o r a blanket power of 62 Mw (thermal) and an

80$ plant factor; t o account

f o r those f i s s i o n products t h a t have v o l a t i l e fluorides o r w h i c h are plated out on parts of the system, a nominal figure of 8 6 is used f o r the fraction that f i n a l l y reaches the waste tank.

The 30-year inventory

of the waste tank i s shown i n Table 2.

Table 2. Inventory i n F e r t i l e S t r e a m Waste Tank A f t e r 30 Years of Collection Waste volume = 1783 f t3 Amount (kg)

Th (as m 4 )

7 ~ (as i LW) 233~ + 233~a(as U F +~ p u 4 ) Fission products (as fluorides)

Unit Value

- Wkg)

141,200

1.41

10,400

120

1.25

116

12,000

1.39

450

4.05 -

In v i e w of the figures In Table 2, the design presented here f o r the

fertile-stream.waste system i s not optimal.

Nearly 99.5$ of the 233U

23%a

value i n the waste tar& is a t t r i b u t a b l e t o 23%a discard; t h i s loss can be reduced t o negligible proportions by providing i n - c e l l decay storage followed by refluorination t

cover the daughter 233U.

1-year' holdup (about 60 f t31

Id allow more than

decay t o 233U.

For example, a

99.995% of the "P' a

.It probably would not be necessary t o provide additional lume because the refluorination could be

fluorination equipment f o r t h i s . scheduled i n existing equi

t during reactor shutdown. 1 of the S a l t Streams'

Flow control of the molten salt streams is by freeze valves coupled w i t h a controlled pressure drop.

This can be achieved by the simple freeze valve currently used i n the MSRE (Molten S a l t Reactor Experiment),

- 36 /

coupled with a flow r e s t r i c t i o n such as an o r i f i c e o r venturi.

A dynamic

freeze valve i n which a controlled layer of s a l t is b u i l t up i n a cooled section of the l i n e is being investigated; i f successful, it w i l l allow a greater ease of flow control. 27 Removal of Decay Heat

Heat removal i s a major problem i n a l l process vessels t h a t contain

In many cases, the heat flux and operating temperature w i l l be high, making it d i f f i c u l t t o use water o r a i r as primary coolants. Water has an additional disadvantage i n that it short-cooled, highly irradiated fuel.

is incompatible with the process fluids, creating a hazard should there be a leak i n the s a l t lines.

Therefore, a sodium-potassium e u t e c t i c

m i x t u r e , 22.3-77.7 w t $ Na-K, d s chosen as the primary coolant f o r process vessels a t temperatures above 5WoC where large amounts of decay This coolant a l s o has t h e capability of i n i t i a l l y heating the system t o 600'~ f o r s t a r t u p purposes. A i r was chosen as the

heat must be removed.

coolant f o r low heat fluxes a t temperatures less than 5OO0C. Sampling the S a l t and O f f -Gas Streams A rather complicated mechanism i s required t o remove a n a l y t i c a l

samples from a highly mdioactive molten-salt system.

A technique similar

t o that tested f o r the E K E will be used, 28 and the off -gas streams w i l l

be sampled conventionally.

Shielding, Maintenance, and Repair of Equipment A l l process equipment that handles material t h a t approaches t h e

radiation l e v e l i n the reactor core w i l l be shielded by about the same amount of shielding as f o r the reactor, and maintenance w i l l be indirect. Process vessels i n this area needing repair w i l l be removed and s e n t t o a decontamination f a c i l i t y before repair.

These include t h e fuel-stream delay tanks, fluorinator, sorption desorption systems, and d i s t i l l a t i o n system.

A l l other equipment f o r processing radioactive materials i s *

- 37 contained i n an area of d i r e c t maintenance w i t h less shielding.

I n the

direct-maintenance area, decontamination f o r maintenance i s achieved by

a molten-salt flush and an aqueous wash. Materials of Construction A l l process vessels and l i n e s i n contact With molten s a l t are made

of INOR-8 except the fluorinators, which are made of a special material,

Vessels and l i n e s t h a t contain fluorine o r fluoride -bearing Alloy 79 gases are made of Inconel o r Monel, and cold traps are made of copper. Other equipment is made of appropriately compatible material.

General Operating Policy The overall policy f o r operating the fluorinator, vacuum still, NaF

beds, and related equipment is 'based on the projected simplicity of opera-

t i o n and small s i z e of the equipment.

The system is designed f o r a

campaign -type operation of one- month * s duration, without shutdown except f o r emergency maintenance. during t h i s period.

There w i l l be no access t o the operating areas

A t the end of the operating period, the e n t i r e system

will be closed down, routine maintenance accomplished, feed hoppers replenished, acc&ulated waste transferred t o waste storage, etc. The operating cycle i s then repeated.

sign study was patterned as much a s

Process equipment f o r possible after previously

and tested equipment.

component was studied f o r

Each major

ication t o the continuous processing

requirements of the MSEB t o ensure a p r a c t i c a l deslgn.

Detailed designs

w e r e made t o the extent t h a t overall size, configuration, heat t r a n s f e r requirements, flow rates, etc., estimate of equipment cost.

defined 'to allow a reasonable

The waste s y s t e m was considered separately,

and liberal use was made of a previous, similar stu13.y'~ f o r i t s design

- 38 -

Fuel Stream ,

Design Criteria

Design c r i t e r i a f o r process equipment were based on past experience. Equipment capacity was based on handling a flow rate 30$ l a r g e r than that required f o r 106 continuous operation as specified by reactor physics calculations. Pertinent basic d a t a adopted f o r t h i s design are as follows: Salt flow rate:

' 15 ft3/day

Pressure: 2 a t m Density of pelleted sorbents, NaZ and MgJ?*:

1 g/cc

Temperature mnge f o r NaK.coolant: 1

coolant: 200'~ .As heat source: 800째C Coolant temperature rise i n heat t r a n s f e r operations: 3OO0C -1 f t P OF-1 Normal convective heat t r a n s f e r coefficient: 10 Btu h r AS

E l e c t r i c a l heat needed: 1.5 % per f t 2 of longitudinal area f o r temperatures higher than 500 C Fission product heat: 5 6 of t o t a l is removed w i t h v o l a t i l e fluorides i n the fluorination s t e p Major Process Equipment There are 41 major pieces of fuel-stream process equipment (Fig. 12). Most of the equipment design is straightforward and based on conventional techniques. 29-32

Each component shown i n the processing flowsheet

(Fig. 12) i s l i s t e d below, with i t s purpose, design basis, and description. The identifying number accompanying the equipment name corresponds t o the c i r c l e d numbers i n t h e figure. Design calculations t h a t involve unusual

techniques o r complexity are shown i n the appendices. 1. Flow Control

Purpose: The flow-control device meters t h e flow of a molten salt stream t o o r from a process vessel. Description: A piece of process pipe, 1/2-in., sched-40, 1 f t long. Pipe i s jacketed with 1-1/2-in. sched-40 pipe made of INOR-8 and has two coolant connections. 2.

Coolant Tank

Purpose: To allow d e l a y ' f o r fission-product decay of the molten-salt stream from the reactor core before i t s introduction t o the fluorinator.

-40

Design Easis: Type of tank: Heat exchanger w i t h well-mixed contents S a l t residence time: (See Appendix A) Temperature : 6 0 0 0 ~ Coolant: NaK Heat load: 5.14 x 104 Btu h r -1 f t - 3 Material of construction: I N O R . 4 Description: A tank having 22.5 f t 3 of l i q u i d volume, with bayonet coolers and two p i p e connections.

Fuel S a l t F l u o r i n a t o r Purpose: To remove a l l uranium and v o l a t i l e f i s s i o n products from the fuel salt by continuous f l u o r i n a t i o n Design Basis:

S a l t residence time: 2 hr Fluorine u t i l i z a t i o n : 33 -1/$ 2 M a x i m u m mass flow rate of F2: 0.277 slpm/in. Temperature: 6oo0c Coolant: NaK Heat load: 5.31 x 104 B t u h r -1 ft-3 Material of construction: Alloy 79-4 Description: A column made of 6-in. sched-40 pipe, 11-1/2 f t long, which is jacketed w i t h an 8-in. sched-40 pipe. Column i s expanded above jacketed s e c t i o n t o 8-in. sched-40 pipe, 1 f t long. Column i s supplied with.5 kw of e l e c t r i c a l hqat and has 5 pipe connections.

Surge Tank

Purpose: To allow molten-salt surge capacity between the f l u o r i n a t i o n and d i s t i l l a t i o n steps. Design Basis: Temperature : 6oo0c Surge capacity: 1 day's continuous flow (15 f t3 ) Coolant: NaK 4 Heat load: 4.4 x 10 Btu h r - l f t m 3 Material of construction: INOR-8

Debcription: A tank having 15 f t 3 of l i q u i d volume, w i t h bayonet coolers and two pipe connections. P

5. Movable -Bed Sorber Purpose: To separate some of the v o l a t i l e f i s s i o n products and corrosion products from uF6. Design Basis: Number of sorption zones: 3 Cooling load i n high-temperature t r a p : UF load based on 12-hr cycle: 13.5 kg N& loading: 0.5 g W6-g Na;F

5.52 x

lo5 Btu/hr C

----

."" .

- ....

. ..

. ., ...

- 41 -

Coolant: A i r Average temperature: 400째C i n bottom zone, 100 t o 150째C i n two top zones NaF Usage: 20$ of one zone volume per day Material of construction: Inconel Description: An annular column made of two concentric pipes. The outer pipe is 10-in. sched-b, 8 f t long; the inner pipe is 6-in. sched-b, 8 f t long with t h e bottom 30 in. finned. Bottom mechanical s o l i d s ejection; 34 kw of e l e c t r i c a l heat.

6. Purpose:

NaF Supply Tank To maintain a supply of NaF p e l l e t s f o r t h e movable-bed absorbers.

Design Basis: w

-NaF supply period: 30 days NaF usage: 10.75 kg/day Temperature: Ambient Material of construction: Stainless steel

Description: A tank, 1 4 2 f t i n diameter and 3 f t high, with a conical bottom and four 2-in. star valves.

NaF Waste Tanks and Cooling S t a t i o n

Purpose: To provide short-term waste storage f o r t h e s o l i d sorbents which. contain sorbed f i s s i o n products. Design Basis:

Tank capacity: 107.5 kg NaF Coolant: A i r Temperature : <400째C on surface Capacity of cooling station: 8 tanks 6 Heat rate of cooling station: 2.0 x 10 Btu/hr (Appendix B) Material of construction: Inconel Description: V e s s e l made of t w o concentric pipes; outer pipe is 10-in. schedS0 pipe, 8 f t long; the inner pipe i s 3-in. schedS0. Vessels supported i n a m e t a l frame.

8. Fission Product Trap Purpose: To remove t h e f i s by- sorption on M g 2 .

oduct

from a uF6 f i s s i o n product stream

Design Basis: Temperature : 150째C Coolant: A i r Capacity: 20 kg of M g 2 p e l l e t s Material of construction: Inconel Description: Vessel made of two concentric pipes; outer pipe i s 6-in. Vessel heated s c h e d a by 5-l/2 f t long; inner pipe i s 3-in. sched-b. by 5 kw of e l e c t r i c a l heat. L_

-42 9.

Primary Cold Trap and UF,: Vaporizer

Purpose: To serve as the primary cold t r a p f o r collecting uF6 from t h e sorber e f f l u e n t and t o be the UF6 feed s t a t i o n f o r t h e reduction column. Design Basis:

uF6 capacity: 1365 kg Temperature: -40 C t o l 0 0 O C Material of construction: Copper Description: An i n t e r n a l l y finned vessel, 5 in. i n diameter and 14 f t long, w i t h 15 kw of e l e c t r i c a l heat 10.

Purpose:

Refrigeration Unit To provide coolant f o r the f u e l - s a l t primary cold traps.

Description:

Freon-type,

rated a t 60,000 Btu/hr a t -45OC.

11. Secondary Cold Trap

To a c t as a backup t o the primary cold trap. Description: An i n t e r n a l l y finned copper vessel, 6 in. i n diameter and 6 f t long, with 5 kw of e l e c t r i c a l heat.

Purpose:

12. Refrigeration Unit Purpose: To provide coolant f o r the fuel-salt secondary cold traps. Description: Freon-type, r a t e d a t 8,000 Btu/hr a t -75OC.

13Purpose:

T ~ P To remove t r a c e amounts of uF6 that escape the secondary cold

trap.

Design Basis: Temperature: Ambient NaF Capacity: 20 kg Material of construction: Description:

Inconel

A vessel made of 4-in. sched-40 pipe,

4 f t long.

14. Fluorine Surge Tank Purpose:

To provide a surge capacity f o r the recycle F2 stream.

D e s i g n Basis:

Capacity: 1-hr supply Temperature: Ambient Material of construction: Description:

Inconel

A tank, 1-1/2 f t i n diameter and 4 f t long.

15. Fluorine Compressor Purpose: To recycle fluorine from the fluorinator off-gas back t o t h e f l u o r i n a t o r inlet.

Type: Diaphragm With remote head Capacity: 6 t o 10 ft3/hr Temperature: Ambient Material of construction: Monel Description:

A diaphragm pump with remote head. 1

16. Vacuum S t i l l Purpose:

To separate fuel-carrier s a l t from less v o l a t i l e f i s s i o n products.

Design Basis: Pressure: 1mm Hgo Temperature: 1000 C3 Liquid volume: 4 f t Heat transfer: Appendix D D i s t i l l a t i o n rate: 15 f t 3 of salt per day Coolant: NaK Operation: continuous addition of feed and continuous removal of d i s t i l l a t e ; periodic removal of residue.

Description: An INOR-8 vessel of shelland-tube design, 2.5 f t i n diameter and 2.3 f t high; 267 f t ? of heat t r a n s f e r area i n l i q u i d section and 6.9 f t i n condensing section; 45 kw of e l e c t r i c a l heat.

17.

S a l t Makeup Tank

Purpose: '20 prepare barren salt i n the nonshielded pperating area f o r use as s a l t makeup. Desi@pBasis:-

,=-

Temperature : ~ O O O C Capacity: 1 f t 3 Material of construction: INOR-8 Description: A tank, 1 f t i n diameter and 1.5 f t high, wi%h8 kw of e l e c t r i c a l heat.

18. L ~ I ?Makeup Tank Purpose: L W makeup f o r t h e d i s t i l l a t i o n unit. Design Basis: Temperature: 9 0 0 ~ ~ Capacity: 4 f t 3 Material of construc

INOR -8

diameter and 3 f t high, with 20 kw of

Description: A tank, 1.5 ' e l e c t r i c a l heat.

19. Coolant Tank f o r FuelIStrean Waste Purpose: To provide short-term storage and cooling of t h e waste stream from the vacuum s t i l l .

._ _ L _ -

I _ I _

_ I _

- 44 Design Basis: Temperature : ~ O O O O C Capacity: 4 f t 3 Heat removal: See Appendix D Material of construction: INOR-8 Description: A ,W having 4 f t3 of l i q u i d volume, w i t h bayonet coolers and 40 kw of e l e c t r i c a l heat. Makeup Tank Purpose: To provide surge capacity and barren salt makeup f o r the molten salt fed t o the reduction column. 20.

LIF-BeF,

Design Basis:

Temperature: 600'~ capacity: 8 hr of retention Material of construction: INOR-8 Description: A tank, 1.5 f t in diameter and 3.5 f t high, with 40 kw of e l e c t r i c a l heat. 21.

Purpose:

Vacuum Pump To provide vacuum f o r the vacuum s t i l l .

Design Basis:

Displacement: 40 cfm Pressure: <5O p Hg. Material of construction: S t e e l Description: Commercial, oil-pumped vacuum u n i t .

Reduction Column Purpose: To reduce the purified m6 in the s t i l l d i s t i l l a t e t o UF4. reconstitutes the fuel. 22.

This

Design Basis:

Temperature: 600'~ Residence time f o r d i s s o l h n g : 30 min Residence time f o r reduction: 1 hr m6 rate: 26.9 kg/day U F mass ~ flow rate: 7 g i n . 2 h r - l utilization: 5 ~ 4 t e r i a l of construction: IXOR-8

Description: A c o l m , made of =-in.. sched-40 pipe, 8 f t high, expanded a t top t o 18-in. sched-40 pipe by 18 in. long. Heated by 50 kw of e l e c t r i c a l heat. 23. Purpose:

MoltenSalt Pump

To recycle molten s a l t i n the reduction system.

Type : Open bowl with helium purge Temperature: 600'~ Capacity: 2 ft3/hr a t 20-ft head Material of construction: INOR-8 Description: 24.

Similar t o pumps developed f o r Molten S a l t Reactor Experiment.

Collection Tank f o r Recycle Fuel

Purpose: To provide surge capacity and a means of transferring molten salt t o the reactor dump-tank. '

Design Basis: Temperature: 600'~ Capacity: I 2 -hr retention Material ol construction: INOR-8

Description: A tank, 2 f t i n diameter and 2.9 f t high, with 27 kw of e l e c t r i c a l heat. 25.

Filters Purpose: To remove p a r t i c l e s from processed molten salt before i t s r e t u r n ' t o the reactor. Design. Basis:

2 F i l t r a t i o n area: 0.1 f t Material of construction: Porous nickel Description: Porous m e t a l f i l t e r i n a 2.5-in.-diam 0.1 f t 2 area and a 2-kw e l e c t r i c a l heater. 26. Purpose:

by 1-ft canister, w i t h

O f f -Gas Scrubbing Column

To remove v o l a t i l e fluorides f r o m waste-gas streams.

D e s i g n Basis: *

a .

Temperature: Ambient Type: Countercurrent packed column Packing: 1/2-in. Raschig rings Source of fluoride gases: 1 6 of F2 (derived from f u e l - s a l t fluorinator), a l l the HF from reduction column, a l l the HF f r o m blanket -salt spargin The 3F rates from these two sources are equal. ms mass flow rate: k s s than 50$ of flooding Aqueous purge: 1 gal of aqueous'KOH per hour Cooling: Water Material of construction

Description: A column, which an expanded section of 6-in i s Jacketed w i t h 5-in. sche Raschig rings.

sched-40 pipe, by 6 f t high, having d-40 pipe by 1 f t high on top. Column ipe,

6 ft high, and packed with 1/2-in.

-~~

~~~

~~~~~

~-~ - -

-46

27. Fission Product Hydrator Purpose: To hydrate a l l v o l a t i l e fission-product fluorides that leave the scrubbing column. Design Basis:

Temperature : 100째C Steam rate: Equivalent t o 1 gal of 5 0 per h r Material of construction: Monel !

Description: A vessel which is 4-in. sched-40 pipe by 2 f t high w i t h a 1-kw e l e c t r i c a l heater. 28.

Condenser

Purpose: Remove the condensable material from gases leaving the f i s s i o n product hydrator. Design Basis:

Temperature: Ambient Coolant: A i r Material of construction: Monel 2 Description: A 4-ft heat exchanger.

29. Absolute F i l t e r Purpose: To remove p a r t i c l e s from the noncondensable gases leaving the condenser. Design Basis: Temperature: Ambienjj F i l t e r area: 0.1 f t F i l t e r i n g medium: Fiberglas Material of construction: Stainless S t e e l Description:

A commercial u n i t .

30. Makeup Tank f o r KOH Purpose: To provide surge capacity and makeup volume f o r t h e scrubbing column. Design Basis: Temperature: Ambient Capacity: 8 h r of scrubber flow Material of construction: Monel Description:

31. Purpose:

A tank, 1.5 f t i n diameter and 2 f t high.

Pump f o r Solution i n O f f a s Scrubber To c i r c u l a t e the aqueous KOH i n the scrubbing system.

Design Basis: Temperature: Ambient Type : Canned r o t o r

- 47 Capacity: 7 gal/hr a t 5-ft head Material of construction: Monel Description: 32. Purpose:

Commercial u n i t .

Aqueous-Waste Tank To provide short-term holdup of aqueous waste.

Design Basis: Temperature: Ambient Capacity: 10 days Material of construction: Description:

Monel

A tank, 3 f t i n diameter and

f t high.

33. Fluorine Storage and Supply Purpose:

To provide t h e primary fluorine supply f o r the system.

Type: Tank and trailer. Description: Tank t r a i l e r containing gaseous fluorine.

34. Purpose:

Hydrogen Storage and supply

To provide the primary

5 supply f o r t h e

system.

Design Basis: Type: 200-scf cylinders with pressure-reducing s t a t i o n Capacity: 48 -hr supply Material of construction: S t e e l $scription:

Four high-pressure cylinders j commercial u n i t s

35.

Hydrogen Fluoride Storage and Supply

Purpose:

To provide H.F supply f o r the system.

Continuous supp temperature steam bath.

Design Basis:

Description:

rom 2OO-lb cylinders i n constant-

Two commercial cylinders.

36. NaK Cooler Purpose:

To remove heat from the NaK coolant.

Desigu Basis: Type : Outside, a i r -cooled NaK capacity: 5251 d / h r NaK temperature: Enter 500째C, e x i t 2OO0C Average air temperature: Material of construction Description:, Air-cooled heat exchanger with 9200 f t2 of heat t r a n s f e r area and a 60-hp fan.

-48

37. NaK Heater Purpose: To heat the NaK so it can be used as a heating f l u i d f o r s t a r t u p and checkout. Design Basis: Type : Heat exchanger using e l e c t r i c heat NaK capacity: 180 ft3/hr NaK temperature: E n t e r a t 500째C and leave a t 8OO0C Temperature on hot side of exchanger: gOO째C Material of construction: Stainless s t e e l Description: A heat exchanger w i t h 270 f t2 of heat t r a n s f e r area and 355 kw of e l e c t r i c a l heat. -

38. NaK Collection Tank Purpose: Design

To c o l l e c t NaK coolant and t o serve as surge capacity.

msis:

Temperature : 5CjOoC Volume: 200 f t Material of construction: Description:

Steel

A tank, 5 f t in diameter and 10.5 f t long.

39- NaK pump Purpose: To circulate t h e NaK through process equipment and heat exchangers. Design Basis:

Type: Centrifugal . Capacity: 650 gal/min with 5 O - f t head Temperature : 500'~ Material of construction: Stainless s t e e l Description:

40. Purpose: fh i d .

A comercial

centrifugal pump, powered by a l 5 - h ~motor.

NaK Supply Tank To provide surge capacity f o r NaK when it i s used as a heating

Design Basis: Temperature : 800'~ volume: 10f t 3 Material of construction: Stainless steel Description': A tank, 2 f t i n diameter and 4 f t long, with 36 kw of e l e c t r i c a l heat.

41. Purpose:

NaK Supply Tank .-

To supply NaK coolant t o process equipment * .

b,.

Description:

A tank,

3.5 f t i n diameter and 10 ft long. F e r t i l e Stream

Design Criteria Process equipment f o r the f e r t i l e stream was also designed f o r 3 6

106 continuous operation. The design rate is lo5 f t3 of f e r t i l e salt per day. Other basic design values are the same as l i s t e d above i n the design c r i t e r i a r o r t n e I u e i stream. . ,

There are 14 major items of process equipment; the function of each is shown i n Fig. 12. The l i s t i n g below gives a description of each i t e m , and the identifying numbers correspond t o the circled numbers of Fig. 12. \

42. Blanket Stream Fluorinator Purpose: To remove uranium from the blanket salt by continuous fluorination. Design Basis: S a l t residence time: 2 h r Fluorine u t i l i z a t i o n : 33 -1/34& WC;rate: , 15.15 g-moles/day

Coolant: NaK 2 mxirmun mass flow r a t e 02 F ~ : 0.277 slpm/in. Coolant load: 42 kw Resistance -heating load: Material of construction:

Description: A column, =-in. sched-40 pipe, 16.5 f t high, which i s jacketed with 14-in. sched-40 pipe, 16.5 in. high. Column i s enlarged a t top t o 24-in. sched-40 pipe, 18 in. high. Heated by 42 kw of lowvoltage a l t e r n a t i n g current applied t o 14 f t 2 of nickel electrodes; supplemental heat supplied by 4 kw units. .

43. Movable -Bed Sorber

Purpose: To separate some products from W6.

he v o l a t i l e f i s s i o n products and corrosion

Design Basis: Number of sorption zones: 3 uF6 load, based on 24 hr of operation:

221 g

NaJ? loading: 0.2 g m6 per gram of NaF Coolant: A i r Average temperature: h o c i n bottom zone, 100 t o 150째C i n two top zones NaF usage: 2 6 of one zone volume per day Material of construction: Inconel

Description: Column made of two concentric pipes, outer one i s 5-in. sched-40 pipe, 3 f t high; inner one i s l.5-in. sched-40, 3 f t high. Mechanical e j e c t o r f o r spent NaJ?; 5 k x of e l e c t r i c a l heat.

44. Fission-Product Trap Purpose: To remove f i s s i o n products from a UF6-fission-product sorption on w2.

stream by P

Design Basis:

Temperature : 150째C Capacity: 5 kg of w2p e l l e t s Material of construction: Inconel

Description: A vessel w h i c h i s 4-in. sched-40 pipe, 2 f t high, warmed by 2 kw of e l e c t r i c a l heat. \

45. Purpose:

R e f rigeration Unit

To provide cooling f o r the blanket-stream primary cold trap.

Description: i i

46. Purpose:

Freon -type with 24,000 Btulhr cooling capacity a t -45OC.

Refrigeration Unit To provide cooling f o r blanket-stream secondary cold trap.

Description:

Freon-type with 4,000 Btu/hr cooling capacity a t -75OC.

47. Purpose:

Product Receiver To receive the uF6 product from the blanket-stream cold traps.

Design Basis: ~.

Temperature : Ambient Capacity: 10 days' production Material of construction: Monel Description:

48. Purpose:

A vessel, 4-in.

sched-b pipe, 2 f t long.

UFc;Trap To remove t r a c e amounts of UF6 from the cold-trap off -gas.

Design Basis: N a F capacity: 5 kg Temperature : Ambient Material of construction:

Description:

Inconel

A vessel, 4-in. sched-40 pipe, 2 f t long

- 51 -

49. Makeup Tank

Purpose: To a c t as surge tank, molten-salt makeup tank, flush-salt makeup tank, and means of molten-salt t r a n s f e r f o r the purified blanket salt.

Design Basis: Temperature: 600'~ . Capacity: I 2 h r retention Material of construction: INOR-8 Description: A tank, 3.5 f t i n diameter and 6.5 f t high, with 105 kw of e l e c t r i c a l heat.

50. F i l t e r Purpose: To &move p a r t i c l e s from t h e processed molten salt p r i o r t o i t s return t o the reactor. .. -

Design Basis: F i l t r a t i o n area: 1 ft2 Material of construction:

Porous nickel

Description: A porous metal f i l t e r i n a 2.5-in. d i m by 1-ft-high canister, heated by 2 kw of e l e c t r i c a l heat. c

51. Waste Tank f o r F e r t i l e Salt Purpose:

Short-term storage f o r f e r t i l e salt waste.

Design Basis: Temperature : 600'~ Capacity: 10 days' retention Material of construction: INOR-8 Description: A tank, 14 in. i n diameter and 2.5 f t high, w i t h 14 kw of e l e c t r i c a l heat. 52. Makeup Tank f o r L i F - W 4 Purpose: For blanket-salt makeup i n the nonshielded operating area t o supply t h e main makeup tanks. Design Basis:

Temperature : 6 0 0 ~ ~ volume: 1 f t 3 Material of constructi Description: A tank, 1 f t e l e c t r i c a l heat.

and 1.5 f t high, w i t h 7.5 kw of

53. Makeup Tank f o r Aqueous Flush Purpose:

For makeup and supply of an aqueous f l u s h f o r decontamination.

Design Basis: Temperature: Volume: 62 f t Material of construction:

4mbient

Stainless s t e e l

- 52 A tank, 3.5 f t i n diameter and

Description:

54. Purpose:

6.5 f t

high.

Line Heater To heat sections of process l i n e s f o r transferring molten s a l t .

Design Basis:

me: Clam-shell e l e c t r i c a l

Rating:

k~ per 10 f t of 1/2-in.

line

Heaters furnished i n 10-ft sections.

Description:

55. Cooling A i r Blower Purpose:

To supply cooling air where needed i n t h e process. a

Design Basis: Capacity: 10,000 ftâ&#x20AC;&#x2122;/min 0 Pressure drop: 10 i n . 5 Material of construction: S t e e l Description: I2 in. %O.

Radial-flow blower w i t h a capacity of 10,000 ft3/min a t

PIANT DESIGN AND LAYOUT The inherent advantage of a fluid-fueled reactor i s i n close-coupling the reactor and processing plants, thereby reducing c a p i t a l investment,

inventory, and operating costs.

This conceptual MSBR processing plant i s

located i n two c e l l s adjacent t o the reactor shield; one contains t h e highradiation-level operations, and the other contains the lower-radiationl e v e l operations.

Each c e l l is designed f o r top access through a remov-

able biological shield having a thickness equivalent t o

6 f t of

high-

density concrete.

Both c e l l s are served by a crane used i n common w i t h

the reactor plant.

.Process equipment i s located i n t h e c e l l f o r remote

removal and replacement from above.

No access i n t o the c e l l s will be

required; however, it is possible with proper decontamination t o allow limited access i n t o t h e lower-radiation-level c e l l .

A general plan of

the processing plant and a p a r t i a l view of the reactor system is shown

i n Dwg. 58050D i n Appendix F

The highly radioactive c e l l contains only fuel-stream processing equipment:

the fluorinator,

and associated vessels.

still, waste receiver, N a F and MgF2 sorbers,

The other c e l l houses the blanket-processing

equipment, f u e l - and f e r t i l e -stream cold traps, UF6 reduction equipment, and f u e l - and f e r t i l e -stream makeup vessels.

v .

- 53 The high-radiation-level c e l l has a cross section of about 19 by 22 f t ; t h e less radioactive c e l l measures about 27 by 30 f t .

Each c e l l

is about 40 f t high, t h e same as t h a t of t h e adjoining reactor c e l l . A "cold" operating area, located along the face of the cells, has dimensions

12 by 62 f t and contains t h e cold makeup equipment, product receivers, and process-gas supply. A l l process operations are carried out remotely from the reactor control room. v

COST ESTIMATE

One of the major goals i n t h i s design study was to estimate the cost of an integrated f a c i l i t y f o r processing t h e f u e l and blanket salts f o r t h i s conceptual 1000-M~( e l e c t r i c a l ) MSBR.

The estimate arrived a t

includes t h e t o t a l fixed c a p i t a l cost and the annual operating cost which are, respectively, $5,301,510 and $787,790 per year.

I n general,

conventional-estimating practices were used except where past experience i n the nuclear energy f i e l d indicated changes. 29 -32

It is d i f f i c u l t t o separate t h e process building and its equipment from the t o t a l reactor f a c i l i t y f o r cost-estimating purposes, since the building m u s t be an i n t e g r a l part of t h e reactor i n s t a l l a t i o n .

determine the relationship between the costs of integrated processing plant and the reactor building, it was assumed t h a t the Molten' Salt Converter Reactor (MSCR) f a c i l i t y , designed by Sargent and Lundy Engineers,33 represented the nonprocessing part of an MSBR. Any required addition to the structure, f a c i l i t y , o r operating cost of the MSCR due t o the addition of the processing f a c i l i t y was considered a cost of the processing ' f a c i l i t y .

Process Equipment

The i n s t a l l a t i o n charge f o r a component t o be i n s t a l l e d i n an indirectaaintenance area was s e t a t 50$ of t h e s e l l i n g price and a t 25s f o r equipment t o b e i n s t a l l e d i n t h e direct-maintenance-and cold areas.

bi. t

A l i s t of a l l major i n s t a l l e d components and t h e i r costs i s given i n Table 3.

The t o t a l cost of the i n s t a l l a t i o n of major process equipment

is $853,730.

- 54 !Table 3.

Equipment Number From Fig. 12 1 2

3 4 5 6 7 8 9 10 11

I2 13 14 15

16 17

20 21 22

23 24 25 26 27 28

29 30 31 32 33 34 35 36 37 38 39 40 41

42 43 44

I n s t a l l e d Cost of Major Process Equipment

of Units

Instalied cost

$ 3,160 184,000 5,420 139,000 21,400 2,750 12,750

NO.

Description Flow control Cooling tank Fluorinator Surge tank Movable -bed sorber N a F supply tank N a F waste and cooling s t a t i o n F is sion -product t r a p Cold t r a p and UF6 vaporizer Refrigeration unit Cold t r a p Refrigeration u n i t U'F6 t r a p Fluorine surge tank Fluorine compressor Distillation unit Makeup tank Makeup tank Cooling tank Makeup tank Vacuum pump Reduction column Molten -salt pump Collection tanks Filters Scrubbing column Fission -product hydrator Condenser Absolute f i l t e r Makeup tank Pump Waste tank Fluorine storage and supply Hydrogen storage and supply HF storage and supply NaK cooler NaK heater Collection tank NaK Pump Supply tank Supply tank Fluorinator Movable -bed sorber Fission -product t r a p

1 1 1 2 1

3 1

820

31,930 6,300

12,280

1 2

5,000

420 1,910

1 1 1 1 1 1 1 1 1 2 2 1 1

L 1 1 1 1 2

4 2 1 1 1

1 1 '1 1 1 1

1,850 58,700 1,570 4,390 58,000 6,590 2,640 9,050 15, mo 11,600

900

1,240

400 800 200

1,410 200

4,540 13,800 780 1,280

60,500 38,670 6,380 2,000

'6,670 3,430 16,600 2, I20

430

- 55 -

Table 3 (continued)

Equipment Number From Fig. 12

Description Refrigeration u n i t R e f rigeration u n i t Product receiver m6 tmP Makeup tank Molten -salt f i l t e r Waste tank Makeup tank Aqueous -flush makeup tank Line heaters Coolingslir blower

45 46 47 48 49 50

No. of Units

51 52

53 54 55

Installed cost

1 1 1 1

3,600 3,300 200 250 63,300 1,060 2,940 1,570 ,

3 4 1 1 1

13 2

Total i n s t a l l e d equipment cost

7,400 6,500 4,760 $853,760

Several of the process vessels and a u x i l i a r y equipment are similar t o equipment previously purchased by ORNL o r previously estimated i n another design report.28

These estimates were updated, and i n some cases

conventional equipment s i z e s were extrapolated t o meet the requirements

1 1

of this l a r g e r plant.

anticipated yere considered t o have costs comparable t o that of the MSKE

d r a i n a n d -f ill tank.

ii -.

For vessels and tanks of conventional design, the cost was computed from the cost of material plus estimated charges f o r fabrication and installation.

! *

The process vessels i n which high heat fluxes were

A summary of the fabrication charges used is given i n

Table 4.

Additional estimating c r i t e r i a were:

I I

Vessel-wall' thickness: Vessel-end thickness :

Equivalent vessel weight f o r entrances and exits:

o r sched-b pipe dim, 1/2 in.

>36-in. dim, 3/4 in.

Cost of e l e c t r i c a l he

i !

i i

1/2-in.

<36-in.

U O O l b vessel,

10 lb 100-1000 lb vessel, 20 l b >lo00 l b vessel, 40 l b $lOO/kw i n s t a l l e d

- 56 Table 4.

29, 32

Summary of Fabrication Cost f o r Process Vessels

Fabrication Cost ($/lb) f o r a Vessel Weight of: Material

U O O O lb

>lo00 lb

Mild steel S t a i n l e s s s t e e l , nickel, Monel, and Inconel

2.00

1.00

3.50

2.50

INOR -8

4.50

Alloy 79-4

5 .oo

3-50

a .

4.00 \,

Structure and Improvements Structure and improvkments costs f o r t h e processing f a c i l i t y w e r e determined by assuming t h a t these costs are d i r e c t l y r e l a t e d t o corresponding costs developed f o r t h e MSCR by Sargent and Lundy Engineers. 33 Addition of t h e processing f a c i l i t y proposed here called f o r an additional

16.56 of f l o o r space i n t h e reactor building, and t h i s f a c t o r t o determine t h e cokt from the corresponding MSCR data. crane and h o i s t was a l s o increased by t h i s amount.

was used

The cost of the

This additional space

includes a lo$ increase i n a n a l y t i c a l and decontamination f a c i l i t i e s .

fie processing f a c i l i t y addition was 7.3s of t h e t o t a l building area; therefore, cost of the grounds and stack was increased by 7.3s.

The

t o t a l increase i n cost of s t r u c t u r e and improvements due t o t h e processing f a c i l i t y was $556,770, which was obtained as follows: structure Crane and h o i s t Grounds Stack

0.1656 x $2,932,400 I

0.1656 x 0.073 x 0.073 x

Total structures and improvements

= $485,610

195,000 = 501,500 = 31,000 =

32,290

36,610 2,260 $556,770

Interim Waste Storage The interim-waste -facili'ty cost was' estimated separately.

This

estimate was based on previous cost estimates f o r similar f a c i l i t i e s . 14332

- 57 Total waste f a c i l i t y cost was $387,970, with the following cost breakdown: Tank cost Excavation and backfill Concrete Transfer l i n e s Inst m e n t a t ion Cooling system S t r u c t u r a l steel

$213,200 32,350 18,070 9,000 10,000

Total

$387,970

loo, 650 4,700

Other Plant Costs

Significant cost items i n the indirect-maintenance area of the plant

are the process l i n e "jumpers", which have remote connectors necessary \

f o r i n d i r e c t connection of pipelines, instrumentation lines, and e l e c t r i c a l lines.

The following cost schedule, based on experience a t ORNL, was used

f o r estimating the cost of these connections: 32 Major pipelines Multiple pipe and i n s t m e n t a t i o n l i n e s E l e c t r i c a l heater connection, includiqg the heater

$1,5~/1ine

$1,700/set $2,000/set

Other process-piping cost schedules were:

Motor-operated control valves $500 each Coolant a i r ducts $lO/ft Major process l i n e s i n direct maintenance area (<20 f t ) as l i n e s i n d i r e c t maintenance area (<20 f t ) N ~ Kcoolant l i n e s (<20 f t ) The above cost schedule r e s u l t s i n a t o t a l process-piping cost of $155,800.

12i;

The e l e c t r i c a l auxiliaries consisted of the e l e c t r i c a l substation, switching gear, feeders, an

indirect connectors and jumpers.

Cost

schedules used f o r these auxiliaries w e r e : 29 E l e c t r i c a l substation Overhead feeders Underground feeders

8f;$F $12.2/ft

The t o t a l cost of e l e c t r i c a l auxiliaries is estimated t o be $84,300. Process instrumentation is estimated t o be $272,100,

radiation

monitoring t o be $100,000, and sampling connections t o be $20,000.

The costs of service l i n e s and high-temperature insulation are based on the i n s t a l l e d cost of the process equipment i n the main processing I

-58

areas and i n the waste storage f a c i l i t y .

The cost of service lines, taken

a t l5$ of the t o t a l i n s t a l l e d cost of t h i s equipment, amounts t o $128,060; t h e cost of the insulation, taken a t 6$ of t h i s i n s t a l l e d cost, amounts t o $51,220 Total Fixed Capital Cost The t o t a l direct plant costs are estimated t c r be $2,609,980 (Table 5). Past experience was used t o determine percentage costs of i n d i r e c t c a p i t a l These percentages are higher t h a n those f o r other chemical industries but represent a c t u a l cost experience i n other ORNL projects. 32

items.

Construction overhead is estimated a t 3 6 of t o t a l d i r e c t plant cost t o give a t o t a l construction cost of .$3,392,970.

Engineering and inspection

a t 25$ of t o t a l construction cost i s $848,240, which results i n a subContingency a t 25$ of the s u b t o t a l plant t o t a l plant cost of $4,241,210. cost is $1,060,300, and the t o t a l construction cost i s $5,301,510. Inventory costs include the cost of the molten salt held up i n the

system and the i n i t i a l cost of t h e NaK coolant. f o r t h i s system is 63 ft3.

A t $1,@O/ft3,

Total f u e l - s a l t holdup

t h e charge i s $89,460.

s a l t holdup is 120 ft3, and a t $560/ft 3 the cost i s $67,200.

cost i s then $196,660,

Fuel- and

b l a n k e t s a l t charges do not include t h e cost of f i s s i l e material.

400-ft3 holdup of NaK a t $loO/ft3 costs $40,000.

Blanket The

The t o t a l inventory

giving a t o t a l fixed c a p i t a l cost of $5,498,170. Direct Operating Cost

The direct operating cost includes the cost of operators, chemical consumption, waste containers, u t i l i t i e s , and maintenance materials.

The

number of operating and support employees i s based on a work schedule of four shifts of 40 h r each per week.

mese include immediate supervisory, operating, maintenance, laboratory, and health physics personnel, plus two people f o r routine c l e r i c a l and j a n i t o r i a l work (Table 6). N o attempt was made t o prorate the cost of higher supervisory, c l e r i c a l , o r plant protection personnel f o r t h e processing f a c i l i t y since some of t h i s cost is included i n labor overhead costs.

cd c

- 59 Table 5.

I n s t a l l e d process equipment

Total Fixed Capital Cost

Structure and improvements

Interim waste storage

387,970

Process piping Process instrumentation

155,800 272, loo

E l e c t r i c a l auxiliaries

84,300

Sampling connections U t i l i t i e s (156 of i n s t a l l e d process equipment)

20,000 ~8,060

I n s u t i o n (6$ of i n s t a l l e d process equipment)

51,220 100, 000

Radiation monitoring Total d i r e c t plant cost Construct ion overhead ( 3 6 of t o t a l d i r e c t plant cost)

782,990

Total construction cost Engineering and inspection (256 of t o t a l construction cost)

$3,392,970 848,240 $4,241,210

Subtotal plant cost Contingency (256 of subtotal plant cost)

1,060,300

Total construction cost

$5,301,510

Inventory cost Molten f u e l s a l t ( a t $1,470/ft3) c

Molten blanket salt ( a t $560/ft3) N ~ K( a t $100/ft3)

89,460 69,200 40,000

Total inventory cost

$ 196,660

Total fixed c a p i t a l cost

$5,498,170

- 60 Table 6.

Employment Costs

Product ion

32, ooo

96,000

Maintenance workers

48,000

Laboratory analysts

24,000

Health physics workers

E,000

Others

10,000

222,000

S h i f t supervisor Opem t o r s

Total

The cost of u t i l i t i e s , waste containers, and consumed chemicals i s

based on a 300 day/year operation f o r both t h e reactor and processing plant. Total d i r e c t operating cost f o r one year is $610,190 (Table 7); t h i s includes f u e l and f e r t i l e salt makeup.

Processing Cost The processing cost per year is estimated a t $1,447,570 (Table 8).

This cost i s obtained by combining the d i r e c t operating cost, the i n d i r e c t cost of labor overhead (So$ of d i r e c t labor cost), the fixed cost due t o depreciation ( 1 6 of fixed c a p i t a l per year), taxes (l$ of fixed c a p i t a l per year), and insurance (I$ of fixed c a p i t a l per year).

The percentage

used f o r the indirect labor cost i s arbitrary; however, it i s within the range of usual practice. On the basis of 300 days of operation per year f o r the 1000-M~ ( e l e c t r i c a l ) MSBR, the f u e l -processing cost is 0.201 m i l l / k w h r

The f u e l

cycle cost is composed of this cost plus the in-reactor inventory of fuel,

fertile, and carrier salts, plus makeup f e r t i l e and c a r r i e r salts, and

less the c r e d i t f o r excess 23%F6 produced. and c r e d i t w e r e not considered i n t h i s study.

In-reactor inventory, makeup,

- 61 Table 7.

Annual Direct Operating Cost

Labor

$222,000

Chemical consumption

$ 4,080

Fluorine (at $2.00/lb)

I I

1 *

KOH ( a t $O.lO/lb) Hydrogen ( a t $0. O l / f t 3 ) ( a t $0.26/1b) N a F p e l l e t s (at $l.OO/lb) MgF2 p e l l e t s (at $l.Oo/lb)

I n e r t gases (guess)

980 720 1,000

5, 780 420

830

Fuel -salt makeup ( a t $1,420/ft3)

28, 400

Blanket -salt makeup ( a t $560/ft3)

27,350

N ~ makeup X (guess)

830

-Waste containers

70,390 28,270

Utilities

E l e c t r i c i t y (at $0. Ol/kwhr )

i i ~

Others (guess)

73,300 71000

Maintenance materials S i t e (guess) Building ( a t 2$ of building cost) Service and u t i l i t i e s (at 4$ of service and u t i l i t i e s cost) Process equipment (at 15$ o equipment cost) Total annual d i r e c t operating cost

2,500 10,810

35,'W

160,040

209,230 $610,190

- 62 -

Table 8. Annual Fuel-Processing Cost

$ 610,igo

Direct operating cost Labor overhead (8oqd of labor cost)

177,600

Fixed costs Depreciation (lO$/year of fixed c a p i t a l )

549,820

Tax (1% of fixed c a p i t a l )

54,980

Insurance (1% of fixed c a p i t a l

54,980

Total

$1,'447,570

Within the applied ground rules, these costs are believed t o be a reasonably accurate representation of the cost f o r regenerating the f u e l and blanket i n an integrated MSBR processing plant.

A more thorough

study would have included detailed design of equipment and layout of t h e integrated processing plant, the reactor and i t s auxiliaries.

Such a

thorough analysis was beyond t h e scope of t h i s study. CONCLUSIONS AND RECOMMENaATIONs

The c e n t r a l issues i n t h i s preliminary study were t o analyze the f e a s i b i l i t y and cost of a conceptual system f o r continuously regenerating t h e f u e l and f e r t i l e streams i n the Molten S a l t Breeder Reactor.

Briefly,

the system consists i n (1)fluorinating, d i s t i l l i n g , and reconstituting

the molten fluorides used in t h e reactor core, and (2) recovering the

23% f r o m the molten breeder blanket by fluorination and using the uranium t o reconstitute the core salt.

The excess i s t o be sold.

The

power of the breeder reactor was set a t 1,000 Mw ( e l e c t r i c a l ) f o r t h i s study. A number of basic conclusions and e s s e n t i a l recommendations were

developed.

The conclusions relate t o the projected f e a s i b i l i t y and

anticipated costs i n terms of established technology and cost accounting practices, and the recommendations refer t o w h a t must y e t be learned w i t h respect t o technology and chemical data before a complete engineering analysis can be made.

It i s our opinion t h a t it w i l l be very useful t o

begin f i l l i n g these gaps i n the knowledge because of the promising simplicity and l o w cost shown by t h i s study. I n the conclusions and recommendations presented below, the recommendations are underlined.

The first four numbered paragraphs

r e l a t e t o genera1,characteristics of t h i s fuel-processing plant, and the others t o specific u n i t operations.

1. FEASIBILlTY. Fluorination followed by d i s t i l l a t i o n i s a feasible process f o r regenerating MSER f u e l (LiF-BeF2-UFb). Fluorination alone i s s u f f i c i e n t processing f o r t h e f e r t i l e s a l t (LiF -ThJ?4). Reactor physics calculations indicate that a t t r a c t i v e breeding r a t i o s can be obtained f o r such a process. Engineering problems i n the processing plant appear t o be amenable t o solution through a well-planned developmental program a t t h e u n i t -operations level. Fluorination-distillation should be developed as the processing method f o r the mBR. Concurrently, other a t t r a c t i v e processes should be investigated a t t h e laboratory and/or engineering stage as potential alternatives. 2.

INmWED PLANT. Integrating processing and reactor f a c i l i t i e s i s of primary importance i n lowering the processing cost. Comp l e t e advantage i s thereby taken of t h e ready adaptability of a fluid-fueled reactor f o r continuous processing w i t h corresponding minimum inventory. The r e l a t i v e l y s m a l l s i z e of t h i s sidestream processing plant, about 12 f t 3 salt per day f o r a 10004~. ( e l e c t r i c a l ) reactor, i s amenable t o integrated construction, thereby separating t h e economic dependence of the processing industry upon a large adount of i n s t a l l e d e l e c t r i c a l capacity. The same financing convention t h a t applies t o t h e power plant applies t o the processing plant; t h i s type of financing i s normally available a t a lower rate than i s available f o r a separated' processing ECONOMY. The estimated c a p i t a l cost, excluding inventory, of t h e plant i s $5,301,000, and operating costs are about $788,000 a year. It is significant t h a t the c a p i t a l investment in the integrated processing plant proposed here is only about 4% of the t o t a l cost of reactor system it serves. CORROSION. mere a t least three areas i n t h e chemical processing plant i n which corrosion behavior of construction materials should be studied. These are the vacuum s t i l l , the reduction unit, and f i l t e r s . The s t i l l temperature of about 1000'C is much hi@;her than has been contemplated f o r any other p a r t of the MSBR system, and th'e resistance of I N O R 4 and nickel t o corrosive a t t a c k a t t h i s temperature i s not known. A reducer, hydrogen, enters t h e reduction unit and probably . helps l i m i t corro on there,. but this should be verified. t o ' a t t a c k because of the large surface F i l t e r s are subje area exposed t o the f l u i d being f i l t e r e d .

- 64 ~

5. FLUORINATION DEVELOPMEXT. Batchwise fluorination of molten fluoride salts f o r uranium recovery has been rather thoroughly investigated a t ORNL; however, it i s recommended .that . - engineering _. development of a continuous fluorinator be given high p r i o r i t y . The need f o r continuous fluorination is evidenced by t h e requirement of low f u e l and c a r r i e r salt inventories i n the processing plant. The reactor f u e l system contains about 650 f t 3 of salt, and, without continuous fluorination, the out -of -pile inventory could possibly be as large as t h e in-pile inventory. This study indicates t h a t in a continuous fluorination-distillation process the holdup represents about loqd of the reactor f u e l volume.

It i s recommended that the frozen-wall concept f o r a , continuous f l u o r i n a t o r be developed and demonstrated. ThPs concept c a l l s f o r a 1/2- t o 3/4-in.-thick layer of frozen salt on the inside w a l l o f the f l u o r i n a t o r t o prevent the serious corrosive a t t a c k by the molten mixture during fluorination. Basic information needed includes fluorination rate data, process control i n continuous operation, and method of establishing and maintaining a frozen wall. i

Fluorination of the f e r t i l e salt introduces problems similar t o those encountered f o r the fuel stream. However, fertile-stream processing rates are 8 t o 10 times higher, and the fission-product a c t i v i t y is several orders of magnitude less. On the other hand, a lower fluorination efficiency can be t o l e r a t e d in blanket processing. Another method f o r continuous fluorination i s the gasphase continuous operation i n w h i c h fluoride microspheres are generated and fluorinated as they f a l l through a tower. This process should be recognized as a p o t e n t i a l a l t e r n a t i v e t o the continuous method of fluorination studied here, but i t s development should be subordinate t o that of the frozen-wall concept.

6. D I S ~ T I O N DEVELOPMENT. The vacuum-distilbtion concept f o r separating the LIF -BeF2 ( f u e l - c a r r i e r salt) f r o m f i s s i o n products is feasible from an engineering viewpoint. The t h e o r e t i c a l n e t discard of 7Li i n the s t i l l waste is low enough t h a t i t s cost i s insignificant compared with other f u e l cycle costs. Thermal problems require that s u f f i c i e n t volume be maintained f o r wetting a large heat t r a n s f e r surface, and t h e buildup of f i s s i o n products i n t h i s volume w i l l almost surely have an adverse e f f e c t on the decontamination f a c t o r of the d i s t i l l a t e . Relative v o l a t i l i t y data are needed f o r the multicomponent mixture, LW-BeFp-fission products, i n which compositions are i n the range 99.5-0.5-0 mol $I t o 84.5 -0.5 -15 mol L W -BeFp -rare earth fluorides It i s - s t r o n g l y recommended that a continuous vacuum s t i l l be b u i l t and operated t o demonstrate a workable design and t o obtain rate and entrainment data.

- 65 I ,

'7.

REDUCTION aF UF TO UF4. This reaction is quantitative when 'xF6 i s fed i n t o an $-F2 flame containing excess %, producing a powdery UF], product. The s o l i d product cakes and adheres t o vessel walxs, which should be avoided i f possible i n a remotely operated system. The liquid-phase reaction proposed here is more s u i t a b l e f o r remote operation and should be developed. The omratinn conditions that need study are: temDerature, UFI.concentration. reaction design. circulation rate.' _ _ rate. nozzle --, --- -- - , contactor design, and gas-liquid separation.

v--,

The reducing conditions t h a t exist i n t h i s operation are consistent with those required f o r purging nickel,'chromium, and iron corrosion products from the f u e l . Therefore the p o t e n t i a l i t y of using the reduction u n i t f o r simultaneously giving t h e f u e l a f i n a l cleabuD should be investimted.

. 8. I

SOLID -LIQUID SEPARATION. The general area of high -temperature, solid-liquid separation i n remote operations needs development. F i l t r a t i o n techniques should surely be i n v e s t i m t e d t o determine O D e r a b i l i t s and reliability i n molten-fluoride systems.

9- F I S S I O N PRODUCT REHATLOR. A better understanding of f i s s i o n product behavior throughout t h e processing plant is needed. In Darticular, data are needed on the w a x i n which the various nuclides Dartition i n the several Drocessinn stem and on the efficiency of removal. A more basic study i s concerned with the behavior of: the f i s s i o n products i n the reactor- environment t o determine whether o r not certain nuclides remain i n the reactor system. 10. URANIUM HEXAFLUORIDE PURIFICATION. The Nal? and MgF2 sorption u n i t s provide adequate decontaminatxon f o r UF6. The batchwise -its can be operated s a t i s f a c t o r i l y f o r both f u e l and f e r t i l e streams of the MSBR; however, a continuous, temperature-zoned system would reduce the frequency of c e l l entry. Probably the l a r g e s t uncertainties i n UFh purification are i n t h e removal of tellurium and ruthenium; means of removing them should be developed.

The portion of t h e f e r t i l e stream used as f u e l makeup need not be passed t h r o rption system because of t h e very small mount of f i oduct contamination. However, t h e n t u a l l y be handled by contact would uF6 product that need purification.

most s i g n i f i c a n t advancement i n fertile-stream processing can be made i n t h e development of a p r o c e s s ' t h a t removes protactinium. To be effective, t h e process m u s t remove protactinium f i v e t o ten times as fast as i t s decay rate; t h a t is, the blanket volume would have t o be processed every four t o eight days. Simplicity and ease of operation are obvious requirements. Thus, a process based on forming protactinium oxide by ion exchange appears promising and should be studied.

If. PRCYI!ACTINIUM REMOVAL. I

- 66 Effective removal would obviate the need of sending any portion of the f e r t i l e stream t o waste and possibly reduce makeup cost t o t h a t represented by thorium burnup alone. Such a process is not required f o r economic operation of t h e ME83 since fluorination and f r a c t i o n a l discard can adequately control the f i s s i o n product concentration; however, t h e potential advantage of reduced waste cost and improved breeding performance argues f o r basic development of a process. 12.

FUELSTREAM WASTE SYSTFM. The most economical method of waste management consists i n bulk storage i n a large heat exchanger tank. The heat generation problem i s so severe that the plant waste must be diluted with an inexpensive, i n e r t material, which complicates future processing fo recove of any contained i n the fluorinationvalues. The calculate'd l o s s of L i and d i s t i l l a t i o n process is small, being only 1.5 t o 2$ of the f u e l cycle cost.

The possible use of fission-product decay heat should not be overlooked in the evaluation of an MSBR. The accumulated waste generates about 4.5 Mw (thermal) throughout most of the f i l l i n g period of the waste tank. Thus salt-storage temperature can be maintained high enough t o make the waste tank a source of high-temperature energy.

13. FERTII;ES!I!REAM WASTE SYSTEM. The fertile-stream waste contains a significant inventory of valuable materials whose recovery i s probably warranted. A t t h e end-of the 30-year f i l l i n g period, t h e waste tank contains about 41,000 kg of Th, 10,400 kg of 7Li, and 116 kg of 233U; the 2333U i s isotopically pure material, having been formed almost e n t i r e l y out of t h e f i s s i o n zone. Although not considered in t h i s study, an i n - c e l l decay period of about s i x months followed by refluorination appears t o be advisable f o r greatly reducing the amount of 233Pa t h a t enters the waste tank.

14.

PROCESS CONTROL. "his aspect of plant operation was given only a cursory review i n t h i s study, and no areas of unusual control d i f f i c u l t y were observed. A flow-control device f o r the moltensalt stream i s needed, and the dynamic freeze-valve concept should be developed. Analytical and sampling requirements require a more thorough study than was given here.

W. G. Stockdale a s s i s t e d the authors i n t h e development of the cost

of the plant and equipment, and his help is' gratefully acknowledged.

- 67 i

REFEKENCES 1. E. S. B e t t i s , ORNL, personal comunication. I

H. F. Batman and E. S Bettis, OFiNL, personal communication on unpublished work s t i l l i n progress.

3. M. J. Kelly, OFiNL, personal comunication on unpublished work s t i l l i n progress. 4. E. D. Arnold, ORNL, personal communication on unpublished work. 5. R. E. Thoma, Rare Earth Halides, ORNL-3804 (May 1965), p. 42. 6. C B. Jackson (ed. ), Liquid Metals Handbook, Sodium (NaK) Supplement, 1955. 3d ed., U.S.A.E.C., Washington, D.C., 7. K. A. Sense, M. J. Snyder, and J.' W. ,Clem, Itme Vapor Pressure of B e r y l l i u m Fluoride, J Phys Chem. 58, 223 -4 (1954). 8. 0. R u f f , G. Schmidt, and S. Mugdan, "Arbeiten aus dem Gebeit Hoher

. .

Temperatur XV . D i e Dmpfdrucke der A l k a l i f luoride, Chem. 123, 83% (1922).

Z Anorg. Allgem.

9. H. von Wartenberg and H. Z

. f u r 'Elektrochemie 27, -

Schulz,

"Der Dmpfdrucke Einiger Salze. 11,

568 -73 ( 1921).

10. K. A. Sense and R. W. Stone, "Vapor Pressures and Molecular

Compositions of Vapors of the RbF-ZrF4 and LiF ZrF4 Systems,

J. Phys. Chem. 62, 1411%

(1958).

11. K. A. Sense and R. W. Stone, "Vapor Pressures and Molecular Compositions of Vapors of the Sodium Fluoride-Beryllium Fluoride System, J. Phys. Che 453 -7 (1958).

E. M. M. Popov, F. A. KO U r a n i u m Tetrafluoride

and N. V. Zubova, "Vapor Pressures of

. Neorg.

Khim.

4.1

1708-9 (1959).

13. W. R. Grimes, Reactor ghemistry Division Annual Progress Report f o r Period Ending January 31, 1g0, OFiNL-2931 (Apr. 29, 1960), p. 50-1. 14. W. L. Carter and J. B. Ruch, A Cost E s t i m a t e f o r B u l k Disposal of Radioactive Waste S a l t from Processing Zirconium-Uranium Fuel by t h e ORNL Fluoride V o l a t i l i t y Process, ORNL-TM-948 (Sept. 25,

1964)

(Confidential).

15. D.

0. ,Campbell and G.

unpublished data.

Cathers, ORNL, personal communication of

- 68 16. G. Burrows, Molecular Distillation, p. 18-31, Oxford University Press, Great Britain, 1960.

17.

W. H. Carr, S. Mann, and E. C. Moncrief,

"Uranium-Zirconium Alloy Fuel

Processing i n the ORNL V o l a t i l i t y P i l o t Plant," paper presented a t the A.1.Ch.E. Fuels,

Symposium, "Volatility Reprocessing of Nuclear Reactor

New York, N. Y. (Dee. 1961).

18. R. P. Milford, S. Mann, J. B. Ruch, and W. H. Carr, Jr., "Recovering Uranium Submarine Reactor Fuels, 'I Ind. Eng. Chem. 53, 357 (1961). 19. J. C. Mailen, "Volatilization of Uranium as the Hexafluoride from Drops of Molten Fluoride Salt," paper presented a t the A.C.S.

National

Meeting i n Chicago, Ill. (Sept. 2, 1964). 20.

G. I. Cathers, M. R. Bennett, and R. L. Jolly, The Fused Salt-Fluoride

V o l a t i l i t y Process f o r Recovering Uranium, oms2661 (1959). 21.

M. E. Whatley e t al., Unit Operations Section Monthly Progress Report,

September 1963, ORNL-m-785 (1964). 22.

E. C. Moncrief, Results of V o l a t i l i t y P i l o t Plant Cold Uranium Flowsheet Demonstration Runs T U 4 and TU-7, ORNL-CF -61-9-37 (1961).

23. 24.

W. W. P i t t , unpublished data, ORNL

(1965).

R. W. Kessie e t al., Process Vessel Design f o r Frozen Wall Containment

of Fused Salt, ANL-6377 (1961). 25. 26.

M. J. Kelley, ORNL, personal communication (1964). L. E. McNeese and C. D. Scott, Reconstitution of MSR Fuel by Reducing

UFC;

Gas t o

(1965).

27.

UF,, i n a Molten Salt, ORNL-IM-lO5l C. D. Scott, unpublished data (1965).

28.

W. L. Carter, R. P. Milford, and W. G. Stockdale, Design Studies and

29.

Cost E s t i m a t e s of Two Fluoride V o l a t i l i t y Plants, ORNL-'1M-522 (1962). John Happel, Chemical Process Economics, Wiley, New York, 1958.

30.

R. S. A r i e s and R. D. Newton, Chemical Engineering Cost Estimation,

McGraw-Hill, New York, 1955.

31. C. H. Chilton (ea.), Cost Engineering i n t h e Process Industries, McGraw - H i l l , New York, 1960. 32.

W. G. Stockdale, ORNL, personal communication

(1965).

33. Sargent and Lundy Engineers, Capital Investment f o r 1OOO-Mwe Molten S a l t Converter Reference Design Power Reactor, SL-1994 (1962).

CB i

- 69 6,

Octave Levenspiel, "Chemical Reaction Engineering,

'' Wiley,

New York,

1962.

35 36

66 - (16),

W. R. Gembill, "Fuse'd S a l t Thermal Conductivity," Chem. Eng.

B. C. Blanke e t al., Density and Viscosity of Fused Mixtures of

Lithium, Beryllium, and Uranium Fluorides, Mound Laboratory Report

2 '

m-1086 (Mar. 23, 1959).

S . I. Cohen, W. D. Powers,

and N. D. Greene, A Physical Property

Summary f o r ANP Fluoride Mixtures, ORNL-2150 (Aug. 23, 1

1956).

W. R. Grimes, Reactor Chemistry Annual Progress Report f o r Period

Ending January 31, 1963, ORNL-3417, p. 47.

39. M. E. Lackey, Internally Cooled Molten S a l t Reactors, ORNL-CF -59 -6 -89 ( 1959

. f

APPENDICES

- 73 APPENDIX A. DESIGN CALCULATIONS FOR FUEL SALT FLUORINATOR AND COOLING "K The u n i t operation of fluorination requires temperature control a t

about 550째C f o r the salt being fluorinated.

The f a c t t h a t t h i s s a l t i s

'highly radioactive introduces a problem because it i s necessary t o e x t r a c t t h i s decay heat through the walls of a fluorinator whose s i z e is fixed by other process requirements, such as throughput and residence time.

there i s i n s u f f i c i e n t heat t r a n s f e r surface' available f o r t h i s purpose, then the salt must be allowed t o "cool" before entering t h e fluorinator. The solution t o t h i s problem is t o i n s e r t a cooling tank immediately upstream of the fluorinator. The following calculations pertain t o t h e thermal design and s i z e of t h e f u e l - s a l t fluorinator and the s i z e of the cooling tank.

It was

determined that t h e maximum permissible heat f l u x f o r the fluorinator i s

5.31 x 104.Btu h r - l ft:3 and t h a t the s i z e is 4.75 i n . i n diameter by 10.3 f t high. The cooling tank requires a volume of 22.5 f t 3 A f u r t h e r

r e s u l t of t h i s calculation is a graph of heat generation rate a t the cooling tank e x i t as a function of elapsed time since discharge from the reactor.

The fluorinator design c r i t e r i a are:

15 ft3/a.ay 77-3 g-moles/day 334346 2

Fuel -salt flow rate Uranium rate, Fluorine u t i l i z a t i o n Maximum mass f l o w rate of g a s Maximum heat flux through the frozen w a l l Heat of reaction ( ~ + 4F2 = U F ~ ) Residence time of s a l t

0.277

s l p /in. 1.5 kw/ft 8 -162 kcal/g -mole 2hr

The F2 flow rate through t h e column is:

Fluorinator cross-sectional area is: D

A = -=*83 277

17.5 i n . 2

- 74 The t o t a l column volume needed is:

~24= 15 x 2

1.25 f t3

Therefore, the column height is:

The area of the frozen w a l l per l i n e a r foot of column i s :

4.75

e,=

The open volume per foot of column is: = 0.1204 f t

f t2

= 1.24

Therefore, the maximum heat removal r a t e i n t h e frozen-wall column is:

le25 '-5 0.1204

3413

5-31 x

104 Btu hr'l

ft-3

But, the heat of reaction contributes a t the following rate:

77.3 x 162 = 1620 Btu h r - l f t - 3 24 x 1.25 x.0.252

Thus, allowable f i s s i o n product heating i s :

5.15 x 104 Btu

h r - l ft'3

The Burge tank must have s u f f i c i e n t capacity t o allow t h e f u e l s a l t t o

cool t o a heat r a t e of 5.15 x 10 Btu h r - l ft'3.

If the tank i s assumed

t o be w e l l d x e d , and assuming t h a t t h e heat generation r a t e of t h e s a l t can be expressed as an exponential function of time, then, 34 (heat generation r a t e ) ,

H ( t ) = Ke-kt

E ( t ) = - e -t/T

(age distribution function),

(A -11

( A 4

where average residence time,

= V/F,

= f l u i d volume i n tank,

= volumetric f l o w rate,

= t i m e since e x i t from reactor core,

K,k = constants.

The teat generation r a t e of molten s a l t from the e x i t of the welllnixed

surge tank can be expressed as:

co E(t) H(t) dt 0

(A -3)

. .

The data on heat generation r a t e w i t h respect t o time elapsed since the s a l t has been removed from the reactor can be expressed as a series of

equations of the form of Eq. (A-1).

S i x i n t e r v a l s were chosen from t h e

data i n Figs. 2a and b (see t e x t ) .

I n each time interval, the constants

of t h e exponential equation approximation were determined by coupling the equation a t t h e two ends of the time interval.

This resulted i n a repre-

sentation of t h e heat generation data, which was always equal t o o r somewhat greater than the calculated heat generation rate.

The values of the

constants i n the, approximate equations are : Time I n t e r v a l (hd

0 -0.0167

5.98 x 105

0.0167 -0.167 0.167 -1

4.28 105 2.46 105 1.41 105 5.85 x 104 2.43 x 104

3 4 5 6

1-10 10 -100

100

24.2 3-98 0.651 0.098 0.0104 0.00154

After substitut-on of t h e constants, integration of Eq. (A-3) by time segments gives :

($ + 24.2)

4.28.

105

+ 1 (;F+ 3-98) ~

2.46 x 105 ( F + 0.651)

[e -0.01667

(1 s + 3.98)

1667 ( + + 0.651)

+ 1

+ 0.098)

- e

-0.1667 ( $ + 3-98)]

- e-(7

-lo(?+ 1 0.098) -e

1 (7 + 0.0104)

-1

00(

1 - e -LOO(?

0.00154)

+ 0.0104)

- 76 This equation was used t o determine heat generation rate of the

molten s a l t a t the cooling-tank exit i n terms of average residence time (Fig. A - 1 ) . A t the design heat generation t i m e , the average residence time was 36 hr.

Therefore, the volume of the cooling tank must be:

rr, J

- 77 -

-0

0 u)

0 *

0 c)

0 cy

, l1

- 78 A.PH3NDIX B. FISSION-PRODUCT HEAT QZNERA!ITON RATES IN TEE MOVABIX-BED SOEiBERS AND N U W A S T E

Heat generation rates f o r the following components are presented here: Movable -Bed Sorber Sodium Fluoride Waste Tanks Short qerm Cooling Station f o r Waste Sodium Fluoride InterimStorage of Waste Fission products, which are v o l a t i l i z e d i n the fluorinator, accumulate i n t h e movable-bed sorbers and create a heat source that must be considered i n the design of these units.

Excess heat must be removed so that it does

not i n t e r f e r e with control of bed temperature.

Actual removal of the heat

is not the problem here since sorbers, which accomplish t h i s end, have been designed-and used i n the ORNL V o l a t i l i t y P i l o t Plant. The e s s e n t i a l problem i s t o estimate the heat generation rate due t o radioactive decay of the f i s s i o n products present i n t h e system projected here.

T h i s was done by f i r s t assuming t h a t half the fission-product heat

generating capacity which reached the f l u o r i n a t o r would e x i t t o the sorber. Further, a l l t h e fission-product heat i s concentrated i n the lower zone i n t h e sorber, and it was removed i n the Na3 waste stream.

A s noted before,

one-fifth of t h e lower zone i s exhausted per day, and there are two p a r a l l e l sorbers a l t e r n a t e l y operating f o r 12 h r each.

This r e s u l t s i n an accumula-

t i o n of decaying f i s s i o n products o r of decreasing heat sources i n t h e sorber, i n the Na.F waste tanks, and f i n a l l y i n the interim waste-storage' facility. The heat generation rate f o r various process components was approximated by determining the average heat generation r a t e during a s p e c i f i c time period, and by assuming t h a t t h i s r a t e decayed as the t o t a l fission-product heat rate decayed as shown i n Figs. 2a and b.

The accumulated heat

generation rate could then be expressed as a r a t e c h a r a c t e r i s t i c of f i s s i o n products having an "average" age intermediate between the oldest and t h e most recently sorbed.

- 79 Movable -Bed Sorber

I n the movable-bed sorber, the heat load can be approximated by

assuming that the v o l a t i l e f i s s i o n productd from the fluorinator accumulate f o r 5 days and that those f i r s t accumulated w i l l decay o r cool as additional accumulation occurs.

A table can be prepared f o r accumulated heat genera-

t i o n rate, derived from the residence time of f i s s i o n products i n t h e sorber : Residence Time of Fission Products i n Sorber (d.aYs)

Heat Generation Rate (Btu/hr)

144,000

E O , 000

3 4 5

106,000

Steady-state heat generation rate i n movable-bed sorber

96,ooo 86,ooo 552,000 Btu/hr

Sodium Fluoride Waste Containers The NaF waste.tanks accumulate NaF and f i s s i o n products from the sorbers.

Each tank holds two complete bottom sorber zones from each of

two sorbers (10 days' accumulation of fission products in one-day inGre-

ments).

These zones exhaust t o t h e NE@' waste tank each day f o r 10 days.

. According t o the slope of the ission-product decay heat curve (Figs. 2a and b), the average residence time of the f i s s i o n products, and the average heat generation rate of the NaF bed material as it leaves the sorber, the

following heat generation rate exists i n the NaF waste tank a t the end of 10 days, a t which time it is f u l l .

- 80 Average Heat Generation Rate (Btu/hr )

Residence Time of NaF i n Container (days) . 1

110,300

83, 400 71,500 62,600 56,600

3 4

5 6 7 8 9

53, f500 50,700 47, 700 41,700 38,700

Heat generation r a t e i n N a F waste tank when f i l l e d

616,800 Btu/hr

Short -Term Cooling Station f o r Waste Sodium Fluoride The NaF waste containers a r e t o be cooled f o r 80 days, within the processing area.

This c a l l s f o r a cooling s t a t i o n with t h e capability of

cooling eight NaF waste containers whose average age varies from about

days t o

87 days. The following heat generation r a t e s apply t o the cooling

station: Identification Number of Containers 1 (about 7 days old)

Average Heat Generation R a t e (Btu/hr )

616,800 F

380,900 261,200 210,400

163,600 145,100

8 (e-mt 87 d ys old) Total heat generation a t cooling s t a t i o n

119,700 104,000 2,001,700

Btu/hr

- 81 InterimStorage of Waste As mentioned above, the wkte containers are t o be cooled f o r a

minimum of 80 days by forced a i r convection p r i o r t o t h e i r t r a n s f e r t o ~

t h e interim wasteatorage f a c i l i t y .

The average heat generation rate of

one container a t the end of t h i s time (based on the slope of the f i s s i o n -

product decay heat f o r t h i s period) will be 104,000 Btu/hr. An average of 28 containers are t o be sent t o t h i s wastestorage f a c i l i t y per year a t one-month i n t e r v a l s f,

per month.

- an average of 2.33

containers

When the c e l l i s opened f o r transfer, the most recently

f i l l e d container has 80-day-oid material i n it,' whereas, the first f i l l e d

On the average, the heat generating rate of t h e

i s about 110 days old.

transferred containers i s c h a r a c t e r i s t i c of one t h a t is 95 days old whose

rate i s 87,500 Btu/hr. The average heat generation rate of containers s e n t t o the interim storage f a c i l i t y is:

2.33 x 87,500 = 203,900' Btu/hr

The heat generation rate of these containers decays with storage 2

time, and t h e average rate i n t h e interim f a c i l i t y f o r one year's accumulation is : Residence Time of NaF Waste Container i n Interim F a c i l i t y (months )

Average Heat Generation R a t e (Btu/hr )

1 2

203,900 150,200

102,000

83,700 69,800 62,200 53,700 !

I I

42,900

37,600 33,300 29,000 26,800

1 #

1 1 12

I $

Average heat gene one -year accumulation of containers

i b

895,100 Btu/hr

4 -

I l l l l i x

-82

For f i v e years of waste accumulation, the t o t a l heat generation r a t e i n the interim f a c i l i t y is: Age of NaF Waste Tanks (years 1

Average Heat Generation Rate (Btu/hr ) \

895, mo 381,300 198,900 132,600 116,ooo

5 Heat generation rate from N a F waste i n interim f a c i l i t y

1,723,900 Btu/hr

-83

APPENDIX C. ESTIMATION OF DISTIILATICN RATE INVACUUMSTILL '

Here, calculations are given whereby the distillation rate of the vacuum still is estimated. The configuration of the still is shown in Fig. 9 of the text. The calculational method used is a modification of the procedure for calculating the rate in molecular distillation as given 16 by Burrows. For our still, it is estimated that at 1000째C and at a '. pressure of 1 mm of Hg, the distillation rates fqr LiF and BeF2 are, -3 respectively, 3.32 x 10 -3 and 2.02 x 10 g set -1 cm-2 . For molecular distillation, a still is designed so that its condensing surface is located quite close to its evaporating surface, thereby minimizing the transport distance for the vapor. If the separation distance is small enough and if the pressure is low enough, a molecule ~ leaving the liquid surface has a very high probability of reaching the condenser without colliding with another molecule. This is the essence of molecular distillation. Our still cannot be described precisely as a molecular still because its operating pressure is too high; however, the pressure is 16~ enough that conditions for molecular distillation are approached. For this reason, the calculationalmethods of molecular distillation, modified to apply to pressures slightly higher than those for true molecular distillation, are used here to estimate the distillation rate for our still. The theoretical rate ~of distillation of a single substance for such conditions can be derived from the kinetic theory of gases, shown by Burrows to be: $ -1 -2 g set cm 5.83 x 1O~~~~mm~ (c-1) vapor pressure in mm of Hg, M is the molecular *e= pmm is the equilibrium weight in grams, and T is the absolute temperature in OK. The resulting distillation rate is expressed as g set -' cm2 . The actual rate of distillation at low pressure will be less than the theoretical rate because there will be coll%sions in the vapor space. Burrows developed an expression for the factor by which the theoretical distillation

rate

should be multiplied

to get the actual /

rate.

Summarizing,

. - 84 ~

h i s treatment considers three events that can occur t o the molecule i n the

vapor space: 1. Some molecules can reach the condenser without a collision; t h i s fraction i s e q , where K i s a dimensionless f a c t o r depending upon

t h e distance between evaporator and condenser, the equilibrium mean f r e e path of the molecules, and the shape of the evaporating surface. 2. The fraction of molecules t h a t collide is (1 - e K ) , . and the fraction of these molecules t h a t reach the condenser i s (1 e -K )e 4s, approximately.

3. From purely geometrical considerations, the probability of a

molecule's s t r i k i n g the condenser after many collisions, which r e s u l t i n random motion, i s a f a c t o r F, the r a t i o of condenser area t o condenser area plus evaporator area. The fraction o f . these molecules that reach the condenser after many \ collisions i s ~ ( 1e X ) ( 1 e x ) .

The t o t a l fraction f of vaporized molecules t h a t reach the condenser i s given by the sum of the three fractions above: f =

= F = 1

- e-K)ex + ~ ( -1e -K )(1 - e x > , + (1 -F)(2eX-eZ) ,

- (1 - F ) ( l

-ex)*

When the t h e o r e t i c a l rate i s m u l t i p l i e d by f, defined t o be the evaporation coefficient, the rate of d i s t i l l a t i o n f o r a single component becomes : -2

5.83 x 10

fpmJm

g s e c - l cm

-2

(c-3)

The evaporation coefficient f is not readily calculated because of t h e d i f f i c u l t y in determining the proper value of K, which is defined t o be: d

K=Fi' where d = gap distance between evaporator and condenser,

%= k

mean f r e e path of molecules i n equilibrium vapor,

= suitable constant, which is 1 o r larger, used t o r e l a t e a c t u a l

conditions t o average conditions i n the gap.

Burrows16 reports

experimental values of k i n the range 3.5 t o 30. \

-85

However, as K increases, that Fig.

the value of f approaches the value of F, a number : is known,from the configuration of the still. For the still shown in 9 (see text), F ~0.58. An estimate of the size of K for this still

can be obtained'for

the two values of k given above and from the fact 24~

xE=

10 -2T

that (c-5)

Pmm o2

where a T

= molecular diameter, cm, = temperature, "K, = equilib-rium pressure, mm Hg. Pmm Substituting into Eq. (C-b), dp= K= For LiF at T

o2 .

2.3 x loso

(C-Q

= l273'K,

2 3.26 k 10~ CIII,

Pnurl= 1.0 mm Hg, and for this still, d 2 10 cm. Using these values, one finds that , , K= 12.1 for k = 30 , and . K= 105.7 for k = 3.5 ; In either case, K is sufficiently-large that Eq.~ (C-2) reduces to f ,-" F 2 0.58 . In processing the fuel stream of the MSBR, distillation must treat a multicomponent mixture of ale-BeF2-UF4-fission products. However, LiF and BeF2 constitute more than 99 mole $ of the mixture, making it possible to treat the mixture as a binary of calculating the distillation rate. Burrows has shown that for a binary the distillation rate, becomes: 5.83 x 10~

fFlxlAJT

mixture,

as shown.in Table C-l, solution for purposes Eq. (C-3), which gives

i g/set

(c-7)

-86

where

“1=

mole fraction of component 1 i n the l i q u i d m i x t u r e , 2 A = area of evaporating surface, cm ,

= equilibrium vapor pressure of component 1 a t temperatme T, mm Hg,

M1 = gram-molecular weight of component 1.

A more rigorous representation of Eq.

coefficient

(C-7) would include t h e a c t i v i t y

of component 1 instead of the evaporation coefficient f t o

account f o r deviations from i d e a l i t y .

However, t h e a c t i v i t y coefficients are not known f o r these salt solutions, s o f o r this study it will be assumed that Eq.

(C-7) gives a reasonably valid estimate of t h e d i s t i l l a t i o n rate.

Since t h e s t i l l operates a t constant volume, a material balance r e q d r e s t h a t BeF2 and LXF d i s t i l l a t t h e same rate a t which they e n t e r The vapor composition is therefore about 69-31 mole $I LiF-EH2, t h e same as it i s i n t h e stream entering the s t i l l . If the t o t a l pressure

the still.

i n t h e vapor space of t h e s t i l l is kept a t 1 m m Hg, t h e p a r t i a l pressures of LIF and EkF2 are 0.69 and 0.31 mm Hg, respectively, assuming t h a t there Table C-1.

Reactor D a t a and Approximate Composition of Fuel Stream a t Equilibrium

Reactor D a t a Fuel volume Cycle time Power

671 f t 3 58 days 2160 Mw ( t h e , m l )

Approximate Composition (mole f d c t i o n ) 233u

0.0029

235,

Negligible

Other U

0.0002

LiF

0.6840

BeF2 Fission products

’

0.3118

o.ooo8a

a Gaseous f i s s i o n products purged i n t h e reactor circulating loop, and noble f i s s i o n products t h a t are removed on a very short cycle by attaching themselves t o t h e INOR-8 walls, do not contribute t o t h i s value.

’

- 87 a r e no other v o l a t i l e compounds i n the s t i l l .

This assumption is valid

because the quantities of v o l a t i l e f i s s i o n products are quite small.

we assume that Raoultts l a w applies, the partial pressure of BeF2 is:

The vapor pressure of BeF2 a t 1000째C i s about 74 mm Hg; therefore the mole fractions i n the s t i l l l i q u i d are approximately: ahd

By Eq. (C-7),the rates a t which LiF and BeF2 evaporate from a surface having area A can be determined: w

5.83

x 10-2 fA[xpJ;MTT'ILIF

g/sec

LIF

and

The vapor pressure of LiF a t 1000째C i s 0.61 mm Hg. When t h i s value and

the values of the other quantities are inserted i n Eqs. (C-8)and ( C 9 ) , the specific evaporation rates are found t o be: 3.32 x

g sec -1 cm9

f o r LiF,,

and 2.02 x

10-3 g sec -1 cm2 f o r B ~ F ~ .

Adding these l a s t two equations and solving f o r t h e area of the evaporating surface :

1.85

2 ft

Now, the area of the pool surface i n the s t i l l (see Fig. 9 i n text) i s 2 However, it i s not safe t o say that the s t i l l Is overdesigned 4.9 f t

by a f a c t o r of about 2.5.

Better data on vapor pressures and p a r t i c u l a r l y

a c t i v i t y coefficients might make a considerable. difference i n t h e calculated d i s t i l l a t i o n rate.

It is strongly suggested t h a t the s t i l l considered f o r

t h i s study be viewed only as an approximate design that will probably change

as more i s learned about the d i s t i l l a t i o n process. \ V

3 t

-89

APPENDIXD. FISSION-PRODUCTACCUMULATIONAND REAT GENERATION RATE IN LITHIUM FLUORIDE POOL.IN'VACUUM STILL Nonvolatile fission products accumulate in the 'vacuum still as the LiF-BeF2 fuel carrier and volatile fission products'are distilled away. The still is initially charged with 4 ft3 of makeup 7LiF, evacuated to a pressure of 1 mm Hg or less and adjusted to a temperature of about 1000째C. Fluorinated fuel salt, containing fission products, is allowed\ to flow continuously into the pool of LiF; and the still is operated so that the There is no bottom rate of distillation is exactly equal to the feed rate. discharge, 'and the volume remains constant. Kelly's work 25 showed that the initial LIF and BeF2 distillate is decontaminated by a.factor of 100 to 1000 from rare.earths. Operation is continued at the above pressure and (1) Either the tempe.rature until one of two phenomena forces termination: solubility of fission products in the 4 ft3 of LIF is exceeded and troublesome precipitation occurs, or (2) the accumulated heat generation rate from fission-product decay begins to tax the'capacity of the cooling system. When the distillation is terminated, the L*-fission-product residue in the still is drained to a waste receiver and eventually to permanent storage. , The operating cycle is then repeated. The aim of the-following calculations is to determine the operating cycle for the still and the limiting conditions for the design shown in Fig. 9. Since one design criterion is to process the fuel-stream with

minimum R

u c

but-of-rekctor

holdup,

and

since

the

solubility

of the

rare

earths

in Lil? at 1000째C is about 50 mole $, it is apparent that condition (2) Decay-heat removal will bea seriousproblem above will be controlling. long before solub&lity limits are approached. It has been determined that dor about 67.4 days at a distillathe 4-ft3 stillcan operate continuously tion rate of 15 ft3/day, processing fuel that has the heat generation The significance of characteristics shown in Figs. 2a and b of the text. this number is that it represents the rate of 7Li discard to waste - 116 kg every 67.4 days. The heat generation rate at this time is 31 x 1Of; Btu/hr, the maximum that can be removed by-the cooling system. The calculations below are believed to give a conservative estimate The basic data on fission-product heat generation of the still performance.

-90

versus decay time represent gross values and hence do not exclude the contributions of those f i s s i o n qmducts removed by the reactor gas sparge, by deposition on metal surfaces i n the reactor system, o r by the fluorinat i o n s t e p i n chemical processing.

Furthermore, the l 5 - f t 3/day design rate

is probably excessive f o r a 1000-M~( e l e c t r i c a l ) MSBR; the economic optimum

rate i s perhaps nearer I 2 f t3/day.

Also, no c r e d i t was taken f o r periodic

interruptions i n processing due t o the reactor's operating a t less than 100$ plant factor.

A l l these factors tended t o shorten the s t i l l operating

cycle and increase t h e discard cost. The calculations are arranged as follows:

F i r s t , an a n a l y t i c a l

expression i s derived f o r the heat generation rate of f i s s i o n products i n the s t i l l as a function of elapsed time since discharge from t h e reactor.

Second, t h i s equation i s then used t o evaluate t h e s t i l l design. Analytical Expression f o r Heat Generation R a t e

When irradiated f u e l s a l t is discharged from t h e fluorinator, it has been out of the reactor about 38 hr.

A cooling tank downstream from the

f l u o r i n a t o r adds another 24 h r of holdup, s o t h a t t h e salt is about 62 h r old When it reaches the s t i l l .

The heat generation c h a r a c t e r i s t i c s of

t h e salt during t h i s period were calculated by assuming an i n f i n i t e l y mixed system; these calculations are described i n Appendix A, and the

results are shown graphically i n Fig. A-1. I

Continuous operation i s

assumed throu&out the system, and, by the time the f u e l reaches the

s t i l l , i t s heat generation (Fig. A-1) is about 4.45 x 10 Btu hr-'

The s t i l l is a sink f o r f i s s i o n products that exhibit decay behavior like that shown i n Fig. 2a and b.

t h e 4 f t3 of LiF i n

ft-3. L

I n i t i a l l y a t t = 62 h r = 2.58 days,

the s t i l l i s fresh material having a zero heat genera-

t i o n rate, but t h i s condition changes rapidly when t h e s t i l l i s put on-

stream. Flow i n t o the s t i l l is continuous, and t h e rate of heat generation w i l l rise u n t i l flow is stopped. The magnitude of t h i s rate a t any future t i m e T i s therefore an integrated quantity over the accumulation period:

to= 2.58 days t o T.

To describe the behavior of the s t i l l during t h i s

period, define t h e quantities \

w .

= s p e c i f i c heat generation rate of f u e l entering still, =

4.45

x 10 Btu h r ' l

ft-3,

q ( t ) = s p e c i f i c heat generation rate a t time t, Btu h r - l ft-3, F

= flow rate of f u e l i n t o still, =

'25 ft3/day,

v ( t ) = f u e l volume processed a t time t,

= time that s t i l l has been operating, days, = volume f u e l pool i n still, =

4 f t3.

The heat generation i n t h e s t i l l pool a t any time t i s Vs q&t),expressed

as But/hr, which is t h e difference between w h a t the rate would have been i f there had been no de,cay and t h e mount the rate has decreased because

of decay.

That is,

Noting t h a t dv = F d t

The limits of integration extend from to= 2.58 days t o t =T days, where T denotes t h e t i m e after discharge from t h e reactor a t which t h e heat P

generation rate is desired. .

To treat Eq. (D-3) analytically, the function q ( t ) i s A

btained from

Figs. 2a and b by representing the curve on t h i s graph by four straightl i n e segments i n t h e range 2 t o 400 days. Fig. D-1.

The procedure i s diagrammed i n

The general form of the equations i s : ( t ) =-ktn

The slope n of each segment i s determined from values of q ( t ) read d i r e c t l y from Figs. 2a and b.

The i n i t i a l conditions of the f u e l entering t h e s t i l l ,

namely, t = 2.58 days, q = 4.45

10 Btu hr'l

ftm3, were introduced i n t o

t h e equation f o r t h e f i r s t segment t o determine t h e constant kl.

The

constant k2 was determined similarly by using the end condition of t h e

ORNL DWG 65-3023Rl

TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. D -1. Schqnatic Logarithmic Curve Showing Approximations t o Fission-Product Decay. I n the mathematical model f o r computing t h e heat generation rate, it i s convenient t o divide the time scale i n t o four p a r t s .

-93

first segment at t = 10 days. This stepwise procedure was followed for segments two and three to evaluate k and k4. The four equations so 3 determined are: _ _ 6.69 x lo4 tl4'43

2 s t1 s.10 days )

9.686

10 c t2 S 20 days )

(D-5)

q3 = 21.2 x 104 t 3 -0.854

20 5 t3 5 100 days ,

(DA)

‘14 = 194.8 x lo4 t4-1*335

100 S t4 S 400 days .

C+

lo4

t2-o'59

In these equations, q is in Btu hr -l ft-3, and t is in days. The function q(t) in Eq. (D-3) is replaced by the four separate functions give:

of Eqs. (D-4 to (D-7),

vs $W

and the integration

= 15 x 104 [4.45 (T - 2.58) - 23.62(t20*41

- 11.74

is carried

(t;-57

- 1oOo41) - 145.2(t30*146

out to

- 2.58°*57) - 200.146)

The units Btu's/hr.

of I, tl, t2, t , and t4 are-days; the units of Vsqs are 3 The,restrictions on the several t's are those specified for '~ 'Eqs: (D-4) to (D-7). In solving Eq. (D-8) for Vs qs(T), a value is chosen for T, and this value is assigned‘to either tl, t2, t‘ or t4 in the 3’ following way: For 2;58 < T 5 10 days: let t; = T ; 10 < T 5 20 days, let t2 = T ; 20 CT ~2 100 -days, let t' =T; 3 100 < T 2 400 days, let t4 = T . ' All ti C T assume their maximum values; all ti > T, of course, are not, considered. F+ure D-2 is a plot of Eq. (D--8); it gives the integrated heat generation rate for times between 2.58 and 400 days after discharge Note that the a,ccumulation time for fission products from the reactor. in the still

is (T - 2.58)

days.

- 94 -

ORNL DWG 65-1821R2

O'at 2.58d

TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. D 2 . Heat Generation Rate i n the L i F Pool Resulting from ' Fission-Product Accumulation i n the S t i l l . The s t i l l i s charged w i t h 4 f t 3 of fresh, molten LiF a t the beginning of the d i s t i l l a t i o n cyole. Barren f u e l - c a r r i e r (2.58 days old) flows i n t o t h i s LiF a t a rate of 15 ft3/day, and LW-BeF2 d i s t i l l s a t t h e same rate, keeping t h e volume constant. Accumulating f i s s i o n products cause a rapid incre'ase i n the heat generation rate.

ib,

Evaluation of VacuumStill Design

Two principal conditions had t o be s a t i s f i e d in the s t i l l design:

(1)t h e evaporating surface had t o be s u f f i c i e n t f o r obtaining t h e required

l 5 - f t 3/day d i s t i l l a t i o n rate, and (2) maximum heat t r a n s f e r surface

had t o

be provided t o minimize t h e frequency w i t h w h i c h t h e s t i l l is drained t o the waste tank.

The choice of dimensions was somewhat arbitrary; a tube\

sheet diameter of 2.5 f t was chosen because it gave an evaporating surface I

that was about 2.5 times t h e calculated area, and a closely spaced arrangement of 2.5-ft-long to-volume r a t i o .

by 1/2-in.-diam

tubes was used f o r high cooling-surface-

The primary unknown operating condition is t h e length of

time t h a t f i s s i o n products .can be accumulated before t h e integrated heat generation taxes t h e capacity of the cooling system; an estimate of t h i s time is determined i n the following calculations. The schematic diagram iniFig. D-3 depicts t h e s t i l l operation and summarizes calculated performance and physical data.

1 l

The calculations w h i c h a r e described below indicate that the s t i l l can accumulate f i s s i o n products f o r about 67.4 days i n a 4-ft 3 L W pool when the s t i l l is operated continuously a t a feed rate of 15 f t3/day.

The

discard rate f o r LiF i s therefore a very small fraction of t h e processing

rate, being about 0.4$. A s mentioned above, the heat generation rate predicted by Eq. (D-8) i s probably excessive because gross instead of n e t fission-product data were used; therefore t h i s accumulation time can be -<

t r e a t e d a6 a lower limit. Physical Data f o r

stili.

Applicable data are given on Fig. D-3. H e a t Transfer Characteristics , The 1000째C temperature of t h e LiF pool i n the s t i l l was chosen t o

achieve adequate d i s t i l l a t i o n rate and fission-product solubility. pool t r a n s f e r s heat t o t h

: u

The

w a l l s by natural convection, and t h i s heat

is picked up on t h e outside of t h e tubes by NaK'(22.3-77.7 mole f Na-K e u t e c t i c ) under forced convection.

The NaK coolant enters the s t i l l a t

- 96 -

ORNL O W 0 65-3040

LiF FROM OUTSIDE

STILL DISTILLATE 15 ft3/doy

AIR OR NaK

LiF-BeF2 DISTILLATE 15ft3

NoK (55OoC) COOLANT

FEED TO UF6UF4 REDUCTION COLUMN STILL AND WASTE RECEIVER DATA DRAIN TO UNDERGROUND STORAGE L i F e FP

. .

Fig. D-3.

8l7-hin. a 16qo. INOR-8 TUBES TUBE SURFACE '267 112 TUBE SHEET '2.511. Dia.aL5ft High STILL CONDENSER 8624 ft.* HEAT LOAD (mo1)*31 a IO6 btulhr EVAPORATING SURFACE a4.9 1t? N E E SPACING * 38 ~n A CENTERS

Schematic D i a g r a m of Vacuum S t i l l Operation.

- 97 55OoC and e x i t s a t 75OoC, giv'ing a logarithmic mean temperature difference of 3 b o C (612%). Heat t r a n s f e r characteristics of LiF were calculated by

using the Nusselt equation f o r natural convection, i n conjunction with the physical-property data given i n Appendix F.

A Nusselt-type equation f o r

flow normal t o banks of tubes was used f o r calculating t h e heat t r a n s f e r properties of the coolant.

Since the t o t a l heat dissipation requirement

of the s t i l l was unknown (because fission-product accumulation time was not known), t h e coolant heat t r a n s f e r coefficient w a s expressed as a function of the heat generation rate.

It was determined that the s t i l l

6 could dissilpate about 31 x 10 Btu/hr f o r t h e NaK flow conditions shown i n Fig. D-3. The overall heat t r a n s f e r coefficient f o r maximum heat flux -2 O F 4 i s about 190 Btu hr-' f t From Fig. D-2, the time after discharge

from t h e reactor corresponding t o the above integrated heat generation is 70 days.

The fission-product accumulation time is 2.58 days less o r 67.4

days because of t h e time l a g before f u e l reaches the d i s t i l l a t i o n step.

LU?, BeF,,

W,,, and Fission-Product Discard Rates.

- A t the end of

t h e 67.4-day cycle, t h e s t i l l contents are primarily LiF and f i s s i o n The equilibrium BeF2 concentration was estimated above t o be

products.

only 0.4 mole $ because of i t s r e l a t i v e l y high vapor pressure a t 1000째C. If it is assumed that the fluorination s t e p is' 99.7$ e f f i c i e n t for uranium

removal, then i n 67.4 days about 3.64 kg of U w i l l have entered the s t i l l .

U r a n i u m t e t r a f l u o r i d e has a vapor pressure a t 1000째C about 2.3 times that of LiF, so s portion of the IF4 w i l l be recovered i n the distillate. The, amount recovered cannot be calculated u n t i l more is'knm a i o i t the vaporliquid e q u i l i b r i a of multicomponent m o l t e n s a l t mixtures. Fission-product

accumulation dur'ing t h i s period is approximately 190.1kg; most of the RbF and.CsF of t h i s inventory will distill because of relatively hi& pressurep

vapor

An estimate of the inventory releg8ted t o waste every 67.4 days i s given i n Table D-1. calculation.

A reactor plant-factor of lo@

was assumed i n the

These values are based on ideal-solution behavior, p a r t i c u l a r l y

with regard t o LiF and %F2;

t h i s i s almost surely not t h e case f o r t h i s

mixture. As more i s learned about t h e a c t i v i t i e s of t h e components, it can be expected that the compositions of Table D - 1 w i l l be d i f f e r e n t from those

Shown

- 98 i)

Table D-1. Estimated Composition of t h e Vacuum S t i l l A f t e r 67.4 Days' Operation a t 15 ft3/Day S t i l l volume = 4 f t3

Weight (kg) 233, 235,

<3 33

Other U

<o .26

<o .002

a 0 5

LiF

0.850

195-9

BeF2 Fission productsa

a Molecular weight

Male Fraction

2.1

0.005

190.1

0.143

= 150, assumed.

Heat Removed by S t i l l Condenser. The load on t h e s t i l l condenser is r e l a t i v e l y small, consisting only of the l a t e n t heat i n 15 f t3/day of a 69-31 mole $I LiF-BeF2 mixture plus some radiation from t h e pool surface. The l a t e n t heat of vaporization of LiF was estimated from vapor pressure data t o be 44,000

was used f o r EW2.

cal/g -mole; an experimental value''

of 50,100 cal/g -mole

Radiative heat t r a n s f e r amounted t o about 62, TOO Btu/hr,

giving a t o t a l condenser duty of 198,300 Btu/hr. For smooth, nonflashing d i s t i l l a t i o n , t h e condensate i s t h e LiF-BeF2 e u t e c t i c having a melting point about 50O0C, s l i g h t l y higher than t h i s i s satisfactory.

and a condenser temperature I f there are deviations from

. )

i d e a l operation so t h a t a higher-melting composition d i s t i l l s , t h e condenser temperature could be adjusted t o temperatures s l i g h t l y above the

melting points of t h e pure components, 8 0 3 O C f o r BeF2 and 845OC f o r LiF.

- 99 APPENDIX E. DESIGN WUIATIONS FOR WAS'ESTORACZ SYSTEM Separate storage is provided f o r f u e l and f e r t i l e stream wastes as fluorides i n underground tanks designed f o r bulk accumulation over a 30/

year period.

Waste management i n the post-30-year period was not conHowever, as mentioned i n the text (see Table 2),

sidered i n t h i s study.

it will probably be desirable to'reprocess t h e fertile-stream waste, which i s only mildly radioactive, a t some future,time f o r recovery of thorium, lithium, and uranium values. About 116 kg of 233U will be present. Fission products, separated i n t h i s recovery, could be stored longer, depending on the activity. I n the fuel-stream waste, t h e only significant value, other than the possible future value of individual f i s s i o n products, 2s 7L i .

However, as

explained below, it is n e c e s s a r y t o add a m i x t u r e of NaF-KF t o t h i s waste t o f a c i l i t a t e heat transfer.

Since' these compounds are chemically similar

t o LW, recovery of the lithium i s d i f f i c u l t .

In any event, a t the end of

t h e 30-year period, the d e s i r a b i l i t y of recovery would have t o be analyzed

in l i g h t of t h e prevailing costs. Two basic problems must be solved i n designing the waste storage sptem:

(1)the integrated rate of heat generation by fission-product

decay m u s t be determined, and (2), using t h e results of (1),the most econonic design f o r t h e prevailing conditions m u s t be found.

The heat

generation rate is computed from the fission-product,decay behavior c

exhibited in Figs. 28 and b (see text), and t h e r e s u l t s are shown i n

Figs. E-1 and E-2. A previous study by Carter and Ruchl' examined a similar wasteatorage problem, and, i n accord with t h e i r recommendation f o r economic waste management, bulk storage i n large, heat exchanger' tanks i s adopted f o r our waste shown on Dwgs. 58080 D

eptual designs of the waste f a c i l i t i e s are C i n Appendix F.

Over the 30-year p This volume includes 520 f t

ion-product mixture drained from

the processing cell, plus 2

XF diluent.

f u e l e t r e a m waste are collected.

f t i n diameter and

6.33 f t high.

The storage tank i s

The corresponding volume of f e r t i l e -

stream waste i s 1783 ft3; this is stored i n a tank 13.5 f t i n diameter by

- 100 -

Most of the following calculations are concerned w i t h the design of the waste tank f o r fuel-stream e f f l u e n t because the most d i f f i c u l t design

problems occur f o r t h i s waste.

A f t e r the decay c h a r a c t e r i s t i c s of t h e

f i s s i o n products and integrated heat generation rate are determined, the calculations examine, i n t h e following order, the basic features of tank design, maximum allowable heat generation rate, volume of diluent, waste-

tank design, and the underground storage f a c i l i t y . Once these calculations are made, computations f o r the fertile-stream waste are almost incidental. FuelStream Waste System Decay Characteristics of t h e Fission Products The i n i t i a l problem i n designing the waste tank i s t o determine t h e

decay i n terms of the time-related c h a r a c t e r i s t i c s of t h e salt being added t o t h e tank.

That is t o say, the s t i l l bottoms, which represent an accumula-

t i o n of f i s s i o n products having ages i n the range 2.58 t o 70 days, generate heat a t a rate c h a r a c t e r i s t i c of f i s s i o n products having an "average" age

somewhere between these values.

The time-related behavior of t h e integrated,

s p e c i f i c heat generation rate i s shown i n Fig. 11 of t h e text f o r the-case of no i n e r t salt dilution.

The graph covers a 5-year collection period of

4-ft3 batches every 67.4 days.

The i n i t i a l point on the graph begins a t

about 140 days; t h i s i s t h e accumulated time i n the processing plant with a

reference time of zero taken as t h e day the f u e l i s discharged from the reactor.

An i n - c e l l delay period of 67.4 days a f t e r removal from the s t i l l

i s included i n t h i s time t o permit some i n i t i a l cooling before draining t o the underground tank.

The decay curves of Figs. 2a and b can be used t o determine t h e decay

behavior when t h i s "average" age has been found.

The volume of f u e l salt

from which f i s s i o n products have accumulated is:

15 ft3/day x 67.4 days = 1011 f t3

It was determined i n Appendix D that the heat generation r a t e a t age equal

t o 70 days out of the reactor is 31 x 10 Btu/hr.

Therefore on the basis

of unit volume of core s a l t the average heat generation rate is: , 6 - 31 lo = 3.066 x 104 Btu h r - l fts3 Qavg 1011

- 101 Define t h e average heat generation,

The processing rate F, i n f t3/hr, is steady, so i n

where v is t h e volume. t h e time i n t e r v a l dt,

dv = F d t

(E4

It was shown i n Appendix D t h a t the instantaneous heat generation rate is

represented by an equation of the form:

, Btu h r â&#x20AC;?

q = k t 31 Combining Eqs

. (E-2) and (E -3)

i n (E-1): t-â&#x20AC;? d t

k Ir

p.vg

ft-3

where k and n ark c h a r a c t e r i s t i c constants for t h e decay curve; to (= 2.58 days) is t h e age of the f i s s i o n products a t the beginning of d i s t i l l a t i o n : the t i m e corresponding t o 4,

The desired average age It w i l l be recalled

t h e discussion of Appendix D ( i n p a r t i c u l a r

Fig. D-1) that the q-vers

curve is b e s t represented by four segments

over the range of i n t e r e s =

6.69 x i o4 tl4.43

q1 +C = 9.686 x

i o4 t2 -0.59 4

four equations are:

= 21.2 x 10 t

854

q4 = 194.8 x 104 t4 -1.335

2 s tl s 10 days, 10

20,s t 100

(E-5 1

20 days,

(E4

s 100 days,

(E-7)

t4 s 400 days.

- 102 \

l i e s i n the range 2 t o 10 days, then i t s value can be found d i r e c t l y

If t

avg 4 from Eq. (E-4) by u t i l i z i n g k = 6.69 x 10 and n = 0.43 from Eq. (E-5).

However, i f t i s i n the range 10 t o 20 days, Eq. (EA) contains a second a% 4 term i n the numerator involving t h e constants k = 9.686 x 10 and n = 0.59 Similarly f o r t i n t h e range 20 t o 100 days, t h e numerator avg of Eq. (E-4) contains a t h i r d term involving the constants of Eq. (E-7). of Eq. (EA).

The solution i s easily determined by trial, and f o r t h i s case the proper

form of Eq. (EA) contains the constants k and n from Eqs. (E-5) and (E-6):

The solution t o t h i s equation is:

11.6 days.

An average age near

the lower end of t h e time scale would be expected because of t h e l a r g e r

contribution by the "younger" f i s s i o n products t o heat generation.

The

quantity t10= 10 days used i n Eq. (E-9) denotes the upper l i m i t of Eq. (E-5) and the lower l i m i t of Eq. (E-6). Integrated Heat Generation i n Waste Tank When the vacuum s t i l l is taken off stream and drained, i t s 4-ft 3

volume i s generating heat a t 31 x 10 Btu/hr,

but t h i s rate is decreasing

i n a manner c h a r a c t e r i s t i c of 11.6-aap01d f i s s i o n products, as shown on A t t h i s point, the change i n rate i s rather rapid, indicating

Fig. 2b.

that a short delay time i n the c e l l before draining t o underground storage w i l l appreciably a l l e v i a t e design requirements f o r the large tank.

I n s t a l l a t i o n of a second 4-ft 3 vessel quite similar t o the vacuum s t i l l i s a convenient way of providing a 67.4-day holdup, during t h i s time t h e heat generation rate decreases by a f a c t o r of about

4.5.

A decay curve was calculated f o r a t y p i c a l batch of s t i l l residue by

assuming t h a t t h e curve was parallel t o t h e decay curve of Fig. 2b f o r MSBR f i s s i o n products a f t e r 11.6 days.

The 67.4-day i n - c e l l delay makes the

accumulated time from reactor discharge t o underground storage equal t o Y

- 103 137.4

days; a t t h i s time t h e heat generation rate of the

6.8 x 10 Btu/hr.

4-ft < 3batch i s

The curves of Fig. E-1 are plotted t o show t h e thermal

history of t h e underground waste tank over a 5-year collection period during which 27 batches a r e added; a l s o t h e decay i n t h e post+-year collection period is shown.

The curve with positive slope represents the

buildup of heat generation rate from a l l batches accumulated up t o the indicated time.

This curve i s actually a stepped curve, but f o r convenience

it has been drawn smooth through the maximum point of each step. The waste tank i s designed f o r 30 years' collection, and the accumulated heat generat i o n a t t h i s t i m e can-be obtained with l i t t l e e r r o r by extrapolating the positive-sloped curve of Fig. E-1. Figure E-2 has been included t o show t h e integrated heat generation

rate i n the underground waste tank when the s t i l l residue i s drained i n t o the tank immediately upon completion of the d i s t i l l a t i o n cycle.

Figures E -1 and E 2 represent an upper l i m i t f o r fission-product heat generation.

As pointed out earlier i n t h i s report, the values are f o r

gross fission-product decay, which includes nuclides t h a t have been removed p r i o r t o vacuum d i s t i l l a t i o n . reactor plant f a c t o r of'

Also the curves were calculated f o r a

lo@.

Basic Features of Tank Design A previous study14 showed that bulk storage i n large tanks i s the most

economical management f o r fluoride wastes.

A 30-year period was chosen

because it coincided with the amortization period of t h e reactor plant, and

a single tank i s sufficie

ecause t h e overall waste volume i s small. For

economy and r e l i a b i l i t y ,

ing by forced a i r d r a f t was adopted, and an

upper l i m i t of 60 f t / s e c

ssigned t o the velocity.

Waste temperature

i n the tank is not t o exceed 75OoC.

Maximum Alluwable Heat Under t h e above ground

l i m i t t h a t can be t o l e r a t e d that the waste w i l l ha

I n 30 years a t

it is apparent that there i s an upper e volumetric heat production rate, and t o w e t s u f f i c i e n t cooling surface.

80$ reactor plant f a c t o r there w i l l

be 520 f t3 of radioactive

- 104 -

째

5 I I

10 I

I l l l l

TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. E-1.

Decay Curves f o r FuelStream Fission Products Cooled

67.4 Days Before Draining t o Waste Tank. The t o t a l heat generation rate i n the underground waste tank i s s i g n i f i c a n t l y decreased by the 67.4-day holdup i n the processing c e l l . Compare with Fig. E-2. Waste i s from an MSBR operating on a 58-day cycle a t 2160 Mw (thermal).

w. c

- 105 -

ORNL DWG 65-1838

I I l i

I I l l

ACCVMULATED.

Fig. E 2 . Decay Curves for FuelStream Fission Products Sent t o Waste Tank Without Prior Cooling n the still bottoms are w i n e d directly t o the underground waste tank, t o t a l heat generation rate i s about 2.8 times that shown on F i g . E which a 67.4-daY cooling i s aU0wed. The cycle t i m e and power of the reactor is the same as for Fig. E-1.

- 106 -

waste salt accumulated whose heat generation rate is about 1.76 x 10 Btu/hr, obtained by extrapolating the upper curve of Fig. E-1. Allowing the coolant t o undergo a temperature r i s e of l25'C,

lb "F 0.25 Btu

1.76 x i o7 Btu -X hr

t h e a i r requirement is: 1

3.13 x 105

(1.8)(125)OF

lb/hr

This is equivalent t o 85,190 f t3/min a t an average temperature of 87". The required cross section f o r a i r flow i s :

85,190 ft3 60

sec

sec ==

2 23.7 f t

This area can be obtained w i t h 894 tubes 2.5 in. i n diameter, w i t h a 9-gage

w a l l ; i n the f i n a l design, Drawing 58080 D, Appendix F, 937 tubes were used.

These tubes, arranged i n a U-tube configuration, can be accommodated i n a

16-ftdiam tank. Waste salt i s stored on the s h e l l side i n the tank.

The available

volume per foot of tank is: 2 mixink

[

(No. tubes )atube

=4 1 -

n2 "tank

The heat t r a n s f e r surface per foot of tank height is:

Therefore, the surface available t o each cubic foot of waste is 7.88 f t2/ f t 3

The mean temperature difference between coolant and s a l t i s 65ioc, and the o v e r a l l heat t r a n s f e r coefficient is estimated t o be 6.2 Btu h r - l f t S

??-I. The maximum allowable heat generation is then: 6*2 Btu x 7*88 ft2 x (651)(1.8) OF = 5.75 x 104 Btu hr'l ft-3 h r f t 2 ?l? ft3

A t a l l times during waste accumulation, there must be s u f f i c i e n t volume

present s o t h a t the integrated heat generation (Btu/hr) divided by t h e t o t a l volume ( f t3) does not exceed t h i s figure.

+ D

- 107 -

kd Volume of Diluent The t o t a l volume of f l u i d i n the waste tank a t any time is the summation of individual batch volumes from the process plus the required diluent.

Expressed mathematically,

N V(t) =

c ( 4 +van>

n=l where Van is the required diluent volume f o r the n t h batch, and t h e summation i s carried out over N batches. 'The above paragraph shows t h a t the t o t a l 4-

volume a t any time i s the quotient of: integrated heat generation rate (Btu/hr ) allowable volumetric heat generation (Btu hrl' f t -3)

-Mt) 5975 x IO4 '

The required d i l u e n t volume f o r each batch can now be determined by solving N

& =

5.75

x 10

c (4

+Vh>

IF1

The value of q ( t ) correspondi o t h e nth batch is *ad from Fig. E-1. It is apparent t h a t the largest diluent volume is required when t h e f i r s t batch is drained t o t h e waste tank.

Eventually t h e tank contains

s u f f i c i e n t volume so that no f u r t h e r i n e r t diluent i s required. behavior i s shown on Fig. c

lllaking the 3O-year volume

This

264 f t 3, ste plus diluent equal t o 784 f t3 In The t o t a l volume of diluent is

a c t u a l practice, t h e t o t a l diluent volume would probably be added a t the

beginning of waste collection r a t h e r than i n d i s c r e t e steps, as shown i n the figure. ,

Waste Tank Design The 937 U-tubes ar interior

- leading CI)

l l e d i n the tank with hone end open t o t h e

of the vault, and the other end welded i n t o an

t o t h e stack. Ai orced i n t o t h e vault passes over the outside of the tank before entering the U-tubes. This design provides about 2 6100 f t of tube cooling surface, which i s about 2.5 times the calculated

- 108 -

TIME AFTER DISCHARGE FROM REACTOR (days)

Fig. E-3. Proportions of Diluent (NaF-KI?) Required t o Insure Suitable Heat Transfer from Waste Tank. F i r s t batch of fuel-stream waste requires largest proportion; no d i l u t i o n required a f t e r 744 days.

requirement.

- log -

The r e l a t i v e l y large number of tubes and the small waste

volume lead t o a tank t h a t has a low height-todiameter r a t i o .

The tank

is 16 f t i n diameter and 6.33 f t high and has a storage volume of about 860 ft3. Monel was chosen as t h e s t r u c t u r a l material. Underground Storage F a c i l i t y The undergrouna storage area is shown on Drawing 58080-D, Appendix F. In addition t o t h e vault f o r t h e fuel-stream waste tankj the area contains

a storage vault f o r s o l i d NE@ and M C 2 wastes from the uF6 sorption step. The design of t h i s portion of the waste system is discussed i n Appendix B.

F e r t i l e Stream Waste System

Design bases use1 f o r fertile-stream wasteatorage were: 1. Thirty-year accumulation i n a single waste tank. 2. F e r t i l e stream power of 62 Mw (thermal). 3. Only one blanket volume (1783 ft3) discarded i n 30 years. 4. Fission-product heat generation as shown on Figs. 3a and b. 5. Coolin$ by natural a i r convection.

It was estimated t h a t t h e integrated heat genekation rate a t t h e end of the 4 30-year f i l l i n g period would be only 5.9 x 10 Btu/hr, and, if the storage temperature is allowed t o be as high as goo%, t h i s heat can be dissipated 2 by about 40 f t of cooling surface. Therefore it is only necessary t o

place cooling surfaces over the tank cross section i n 1ocations.that shorten ~

the path f o r heat conduction through t h e salt.

Twelve b-in.-diam pipes

equally spaced over the cross section a r e provided t o remove i n t e r n a l heat. Regions of s a l t most d i s t a n t from a cooling surface m i g h t be molten during some period

i n tank lifetime, b u t t h i s w i l l not present a corrosion problem

because cooling surfaces w i l l always be covered with a frozen salt layer.

13.5 f t

&e storage tank i s about 1900 f t3 of storage

lume.

construction because on

casionally will molten LIF -ThF4-fission product

m i x t u r e contact a m e t a l s

f t i n diameter and

high, providing

S t a i n l e s s steel can be used i n t h e The tank i s contained i n a n underground

concrete vault as shown on Drawing 580814, Appendix F. /-

- llo APPENDIX F.

PHYSICAL-PROPERTY DATA AND DRAWINGS

This Appendix contains the following information: Table F-1

Thermal Data f o r LiF, Bel?*, Na, K, and NaK

Figure F-1

Calculated Density of MSBR Fuel S a l t and LiF

Figure F-2

Calculated Density of MSBR F e r t i l e S a l t

Figure F-3a and b

Vapor Pressure-Temperature Curves f o r Several Metal Fluorides

Figure F-4

Vapor Pressure of NaK

Figure F-5

Viscosity and Thermal Conductivity of LIF

Figure F-6

Viscosityqemperature curve f o r N ~ K(22.3-77.7 w t $) Alloy

Figure F-7

Properties of NaK (22.3-77.7 wt $) Alloy

Figure F-8 Figure F-9

Process Flowsheet f o r Fuel and F e r t i l e Streams Underground Storage System f o r Fuel Stream Waste

Figure F-10

Underground Storage System f o r F e r t i l e Stream Waste

Figure F-11

Arrangement of Processing Equipment f o r Fuel and FertLle Streams

Table F-1.

Thermal Data f o r LiF, BeF2, Na, K, and NaK

Latent Heat of Vaporization (cal/g)

LiF

1690 ( r e f 38; calculated from vapor pressure data)

1070 (ref 11)

BeF2 Na

: 1038 (ref

NaK (22.3-77.7 wt $):

6) 496.5 (ref 6) 617 (calculated) Heat Capacity (cal/g 'C)

LiF

. c

1.88

r P r

1.68

L I

LiF 68.5 mo! % BeF2 31.2 mol % UF4 0.31 mol %

MELTING PblNT = 500'C

I \

TEMPERATURE ('C)

F i g . F-1.

Calculated Density of MSBR Fuel Salt and L U .

-112

4.10

71 mol % (17.11 w t %) ,(82.79 " ) 33U 0.012 It (0.035 I' 33Pa 0.022 (0.064 " )

TEMPERATURE ("C)

Fig. F -2.

Calculated Density of MSBR F e r t i l e S a l t .

- 113 -

5.5

hd r*'

6.0

Fig. F -3a. Vapor Pressure Fluorides

7.0

7.5 1 0 ~ 1em ~

8.0

8.5

9.0

- Temperature Curves f o r Sever& Metal I

Fig. F-3b. Fluorides.

Vapor Pressure

- Temperature

Curves f o r Several Metal

- 115 i,

ORNL DWG 65-2994 R1

E E Y

..- 3

12.â&#x20AC;&#x2122;

(OK) i

.k,

Fig. FA. .Vapor Pressure of NaK Eutectic.

- 116 -

O R N L DWG 65-3049

4.04

4.02

4.00

3.98

3.96

e TEMPERATURE ("C)

TEMPERATURE ("C)

Fig. F-5. Viscosity and Thermal Conductivity of L i F .

u r

- 117 I

0.54

00%

0.44

0.42

h n

0.38

c C 0

7 0.333 e743 p/T

0.34

0.30

0.26

0.22

r(cp) = 0.1Mj p

P = s/cm T = OK

0.18

0.14 EMPERATURE ("C)

Fig. F-6.

Viscosity-Temperature Curve for NaK (22.3-n.7 w t $) Alloy.

(Data from Reference 6.)

- 118 -

TEMPERATURE ("C)

Cl?

c P

4 I ,

TEMPERATURE ("C) j

Fig. F-7. Reference 6. )

Properties of NaX (22.3-77.7 w t $) Alloy.

(Data from

- 119 -

f&f?.ECYCLcr

\ .

RECYCLE

rL-----

I I

--I--

I I

1 I

Volwne U-LSS Pa-2SS

* It0 Kg

= 141, t S O K

LiF

I788 Ft'

-62.7Kg

38,765 $9

= 4.4 s

ravg Volume U-235 U-255

'1.93

Other U

= 58.1 Kg 19533 Kg

LiF

rc%

' 11.g We33 l2.G68Kg/d O)horUWe35 0.191 I.OOi?Kg/d Kg/d

FUEL STREAM

FERTILE STREAM

FlSSldN WWER

91pp"

I .I

NET

NET Fl35lON WWER

I I

-671Ff.J ' 736. S Kg '11.1 Kg

LiF BsF,

277.6 Ky/d I

*16099Kq C Fucl 4. 56700 Kg

F*Mgr(

eCFe

GONTINIIOUS FLUORINATOR

* 5 5 6 . 8 Kg/d

1000.F.

FLUORINATION

f93XURmowl) TANK

I 1

I I

t I

Th 15.75 Kg/d LIF* 3.54 K g / z

b40

' I

UUZSS t 3 3 *le.CWKs/d * 0.100 Kg/d

1 I

Fig. F-8.

Process Flowsheet f o r Fuel 1 and F e r t i l e Streams.

RwERma?-

--I

r m

I-

< L

k 0

E '

a, a A

- 122 -

PgI

PROCESSING

CELLS

WPPLY AND

MAKEUP AREA

-t

aowt

PRELIMINARY EQUIPMENT LAYOUT

- t W M C . .I'-O.

Fig. F -11. Arrangement of Prc :essing Equipment f o r F u e l and F e r t i l e Streams. The highly r a d i o a c t i v e 01 m t i o n g i n fuel-stream processing are c a r r i e d out i n t h e smaller c e l l (UJ ,er l e f t ) . The otHer c e l l houses equipment f o r the f e r t i l e stream ar , t h e c o o l e r f i e l d stream operations. 'C Y

123

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