E ’,_ Zf f | E v En ORNL-3791 UC-80 — Reactor Technology TID-4500 (46th ed.) PRELIMINARY DESIGN STUDY QF A CONTINUOUS FLUORINATION—VACUUM%DISTILLATION SYSTEM FOR REGENERATING FUEL AND FERTILE STREAMS IN A MOLTEN SALT BREEDER REACTOR C. D. Scott W. L. Carter RELEASED FOR ANNOUNCEMENT IN NMUCLEAR SCIENCE ABSTRACTS Top S OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document. ORNL~-3791 Contract No. W-7L05-eng-26 CHEMICAL TECHNOLOGY DIVISION PRELIMINARY DESIGN STUDY OF A CONTINUOUS FLUORINATION- VACUUM-DISTILLATION SYSTEM FOR REGENERATING FUEL AND FERTILE STREAMS IN A MOLTEN SALT BREEDER REACTOR C. D. Scott W. L. Carter JANUARY 1966 OAK RIDGE NATIONAL ILABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.8. ATOMIC ENERGY COMMISSION Bons iii CONTENTS Abstract =--ecmam et 1 Introduction wmeccmmm e e e - 3 The Molten Salt Breeder Reactor System --cemmeccmmcmm e e e o 5 Design Criteria —-----cmmmmmm e e e e 7 Basic Consideralions =-e-mmcmmmmm e e e e e e It Process for the Fertile Stream m--c-cemmcm e e mm e e e e 9 Process for the Fuel Stream =ccememrcme o m e e e e e e e e e 9 Waste STOrage --mmmmm e e e e e e e e - 10 Operating POliCY ==mermm o s e e e e e e e e e e e 10 Process Dats —ceececm e e e — 11 Description of ProCess memememmcmcm o e e e e e e 11 Summary of the Process Flowsheet —=-em-mmemcommmcme e 16 FLlUuorination =--—== == oo o o e e 16 Purification of Uranium Hexafluoride by Sorption and Desorption --- 19 Vacuum Distillation ==mecmm e e e 25 Reduction of Uranium Hexafluoride and Reconstitution of the Fuel -- 28 Off -Gas Processing ==c-mmme e e e e e e 29 Waste Storage =--=-cmm oo e e e e e 29 Flow Control of the Salt Streams -—-—--cccmmmmcm e e e e e 35 Removal of Decay Heat --m--emmcm s cm oo e e e e e 36 Sampling of the Salt and Off-Gas Streams -----m;-—cmcccmrcc s e 36 Shielding, Maintenance, and Repalr of Equipment ------ceeeweeeoeou- 36 Materials of Construction =-meemecmemem e e e e 37 General Operating POliCy =me=cmcmmmmm e e e e e e e e e e e e 37 Process DesSign =—r-mm e e e e e e e e e 37 Fuel Stream -——c-ememmm o e e e 38 Fertile STream -—----cmcemm oo e e L9 Plant Design and Layoub =—=-c-mme oo e e 52 Cost Estimate —---romcmc e e 53 Process Equipment ==cecommmm e e e e e e e e e 53 Structure and IMprovements == --mce e m e c e e e 56 Interim Waste StOrage ———---em oo oo e e 56 Other Plant Costs =--remmeme e e e e 57 Total Fixed Capital Cost —--mmmmmem o e e 58 Direct Operating Cost —mmmm o e oo c e e e 58 Processing COBT ==mm e s o oo e e el 60 Conclusions and Recommendations ==—=mee—mmm oo oo oo - 62 Acknowledgement ===-em-m—m e e e e - 66 References ==---mmmm o e s e e e e ————————— 67 Appendix A. Design Calculations for Fuel Salt Fluorinator and Cooling Tank m--mmc o e e e e e 73 Appendix B. Appendix C. Appendix D. Appendix E. Appendix F. iv Fission Product Heat Generation Rates in the Movable - Bed Sorbers and NaF Waste Tanks ---scemecmccceccccmcmcman-— 78 Movable -Bed SOrber mmeemcmmccccc e e e e ————— 79 Sodium Fluoride Waste Contalners -—-meeccemccccmacmeeaa 79 Short-Term Cooling Station for Waste Sodium Fluoride-- 80 Interim Storage of Waste —weeemmcmmm oo 81 Estimation of Distillation Rate in Vacuum Still ~-weeeee- 83 Fission Product Accumulation and Heat Generation Rate in Lithium Fluoride Pool in Vacuum Still --cceemmeemaaaan 89 Analytical Expression for Heat Generation Rate -=--w-- 90 Evaluation of Vacuum-5till Desigh «=-=mmccmccmmcmm e 95 Design Calculations for Waste-Storage System ----ceemeew- 99 Fuel Stream Waste System —cecmmmmmmmc e e 100 Fertile-Stream Waste System ==e-cwcccccmcc e 109 Physical -Property Data and Drawings ==e=ceceecceeoaceaaoo 110 PRELIMINARY DESIGN STUDY OF A CONTINUOUS FLUORINATION— VACUUM -DISTILIATION SYSTEM FOR REGENERATING FUEL AND FERTILE STREAMS IN A MOLTEN SALT BREEDER REACTOR C. D. Scott W. L. Carter ABSTRACT The purpose of this study was to make a preliminary design and an engineering evaluation of a conceptual plant for treating the fuel and fertile streams of a molten-saltl breeder reactor. The primary requirements of the process are to recover the unburned fuel (233UFM) and fuel -stream carrier salts (LiFéBng% from the fuel stream, and the LiF -ThF), plus the bred 33U from the fertile stream. Both streams must be sufficiently decontaminated for attractive breeding performance of the reactor. The plant was designed to operate continuously as an integral part of the reactor system, Titting into two relatively small cells adjacent to the reactor cell. In this study, the plant capacity is based upon treating 15 ft3/day of fuel salt and 105 ft3/day of fertile salt removed as side streams. These capacities are adequate for a 1000-Mw (electrical) power reactor. As to the fuel stream, bvasically it is purified by fluorination and vacuum distillation. The first step removes uranium as volatile UFg; the second recovers the Li¥ -BeFo by simultaneously volatilizing these two compo - nents from the less volatile fission products. Fortunately, the fission products so separated are primarily the rare earths, which are the most serious neutron poisons. The UFg from the fluorinator is accompanied by some volatile fission product fluorides, primarily those of Mo, Te, Ru, Zr, and Nb, which are removed by sorption on granular NaF and MgFo,. Finally, the UFg is reduced to UF) by hydrogen and recombined with the decontaminated LiF -BeFp carrier in a single operation. Fission products are removed from the plant by discard of NaF and MgF, sorbents and the still residue, which is a highly concentrated solution of the rare earths in Li¥F. Wastes are permanently stored underground. With respect to the fertile stream, the process consists only of fluorination followed by decontamination of the UFg on NaF and MgFo sorbents. It is only necessary to remove the bred ©33U sufficiently fast to keep 2 low concentration in the blanket, thereby ensuring a low fission rate and negligible poisoning by fission products. Discard of the barren fertile stream at a slow rate suffices to keep the fission product concentration at a tolerable level. The chief conclusions of this study are: (1) that the Tluorination-distillation process for the fuel stream and the fluorination process for the fertile stream comprise a compact and relatively simple system that can be engineered with a normal amount of developmental work, and (2) that integration of the processing plant into the reactor facility is both feasible and economical and the logical way to take advantage of the processing possibilities of a fluid-fueled reactor. The nominal cost of this plant is presented in the following summary of major items: Process equipment and building space $5, 302, 000 Fuel salt inventory 89, 500 Fertile salt inventory 69,200 NaK coolant inventory 40, 000 Direct operating cost 788, 000 /year These costs contribute about 0.2 mill/kwhr to the fuel cycle cost when the reactor operates at an 80% plant factor and ~ capital charges are amortized at 10%/year. This cost is sufficiently low to add to the incentive for developing the molten-salt breeder reactor. Some of the steps of the evaluated process are based on well -established technology, whereas others are based on extrapolations of laboratory and small -scale engineering data. - Fluoride volatility and assoclated UFg decontamination by - sorption are well -known operations, having been demonstrated ~in a pilot plant. However, vacuum distillation and liquid- phase reduction of UFg to UF) have been demonstrated only at the bench. Certainly, more development of these steps is required for a complete process demonstration. A singularly serious problem is the corrosive nature of the fluorine-molten salt mixture in the fluorinator. However, this study shows that this and other inherent processing problems can be solved by proper design and operation of equipment. | -3 - INTRODUCTION A reactor concept that has a high potential for economic production of nuclear power and simultaneous breeding of fissile material on the thorium- uranium fuel cycle is the molten-galt breeder reactor (MSBR). This reactor utilizes two fluid streams. For this study, the stream compositions are: (1) & fuel stream consisting of an LiF -BeF, (69-31 mole %) carrier that contains the fissile component, and (2) a fertile stream, which surrounds the fuel stream, consisting of a 71-29 mole % LiF -Th¥) mixture. In each stream the lithium is about 99.995 at. % 7Lia The above compositions are 39 not unique; other MSEBR designs might use different compositions. The con- figuration of the system allows a relatively high neutron leakage rate from the fuel stream into the surrounding fertile stream, where capture by thorium breeds additional 233U fuel. The reactor is operated at an average temperature of about 650OC; fission energy is recovered in external heat exchangers through which the fuel and fertile streams are circulated. Sustained breeding performance of the MSER depends on the removal of fission and corrosion products from the two fluid streams so that parasitic neutron capture is kept to a tolerable rate. Portions of the two circu- lating streams are continuously removed, processed for removal of fisslon products, fortified with makeup Tissile and fertile materials, and returned to the reactor in a cyclic operation. A primary consideration of any process for recycling reactor fuel is that minimal losses of all valuable fuel componente be attained without intolerable capital investment cor operating expense. In the MSBR system, this requirement applies to both fuel and carrier components. As mentioned above, the fuel is 233U, I and one carrier component is the highly enriched Ii isctope. The other major constituents, beryllium and thorium, are less expensive; yet large losses of these cannot he tolerated either because of the adverse effect on fuel ~-cycle cost and fuel conservation. The process evaluated here accomplishes the obJjectives of conservation while providing fission-product removal sufficient for a successful breeding system. The work reported here is unique in that it examines a processing plant integrated directly into the reactor system, which, in effect, accomplishes on-stream processing. This method obviates the cumbersome Lo and expensive transfer of highly radiocactive material by carrier shipment; furthermore, common use can be made of services and equipment necessary to the reactor, thus avoiding the duplication that results in a separate processing building. Also, since the spent fuel flows directly into the processing plant, there is minimal out-of -pile inventory of valuable fuel components. Another interesting feature of this study is the use of the relatively recent cdfiéept of wvacuum distillation as a means of purifying the carrier salt. A modest vacuum of only 1 to 2 mm Hg is required, but the tempera - ture (about lOOOOC) is higher than any normally encounteréd in handling molten fluoride salts. The operation was explored first by Kel]_ey3 in laboratory exPefiments that indicated LiFéBeFe decontamination from fission products by factors of 102 to 103. The attractiveness of the process lies in the fact that it involves only a physical operation that is easily controlled and that can be made continuous. Fission products can be concentrated in the still residue (primarily 7LiF) by a factor of about 250 by using the decay heat of the fission products to volatilize LiF—BeFE. This high concentration factor ensures a low discard rate for the valuable 7L:'L. Cyclic operation of the still was assumed, allowing the fission product concentration to increase with time. The corresponding increase in the rate of decay heat generation limited the cycle time to about 68 days. Although limited experience with the distillation step indicates that high-nickel alloys are satisfactory structural materials for use at this relatively high operating temperature, a more extensive investigation is needed to define the design limitations. A novel idea has also been studied in the evaluation of liquid-phase reduction Qf UF6 to UFA by hydrogen. Initial bench-scale experimentse6 have,given promising data. The reaction is carried out by absorbing UF6 in a molten mixture of LiF—BeFQ—UFA at about 600°C followed by contacting with H2 to reduce UF6 in situ. This technique avoilds the troublesome problem of remotely handling solids (small UFA particles) that would be net if the customary gas-phase reduction of UF6 were used. Aside from indicating the feasibility and economy of the process, this study also uncovered important design and engineering problems assoclated with the scaleup of laboratory and batchwise operations to 4 o _5_ larger, continuous ones. In this regard, recommendations are presented at the end of the text, along with important conclusions. The most note- worthy recommendation is that the key operations, vacuum distillation and continuous fluorination, be given high priority in development. The material that follows is arranged in this sequence: TFirst, a brief description of the MSBR system is given to put the study in per- spective; second, design criteria and ground rules are stated for each phase of the study; third, a process flowsheet and a description of each unit operation is presented; fourth and fifth, a description and pertinent design data for each major component and the processing cells are listed; sixth, equipment and building-cost data are presented; seventh, an over- all evaluvation of the process is given in a set of enumerated conclusions and recommendations arranged according to plant characteristics and process operations. Six Appendices, giving detailed data and calculations, are attached. THE MOLTEN-SALT BREEDER REACTOR SYSTEM 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 designl of a power reactor capable of producing 1000 Mw (electrical) with a thermal efficiency of 45%. Basically, it consists of a graphite matrix enclosed in a cylindrical Hastelloy N vessel for containment. GCraphite occupies about 79 vol % of the core, fuel salt about 15 vol %, and fertile salt about 6. The flow passages are such that the fuel and fertile streams do not mix. The core is surrounded radially and axially by a 3.5-ft blanket of LiF-ThFM 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 16 ft in diameter and 25 ft high. Fission energy is recovered in a battery of external heat exchangers through which the 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. ©OSmall side- streams of fuel and fertile fluids are continucusly withdrawn from the ORNL DWG 65-3017 _ COOLING FUEL _ ' rju"’S-SA‘\LT PUMP(4eq) BLANKET FUELBANKE —gp PUMP{(2ea) — 3 it o ; CONTROL ROD{leq) 7 REFLECTOR: : Pb SEAL- EXPANSION JOIN s 1 4 /T-\‘\ : TR Il 1 . INT / % ] (.‘_, : MODERATOR— | : _ _ _ | _ R uu N : ! FUEL HEAT § ' ! L+ EXCRANGER ' ' §§ 1 , CHANGER N H IR =2 v ' - A : . | WA BE e NAATT ;QJV}[ \ific——‘/ o LS - /‘I}{’ N - — — oLAN ST 5 Few — | BLANKET HEAT EXCHANGER (2eq) SECTION "A-A" Fig. 1. Conceptual Design of the Molten-Salt Breeder Reactor. - ._7... 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 fluids are returned to the reactor via the fuel-makeup system. The processing cycle is selected to give the optimum combination of fuel-cycle cost and breeding gain. The data presented in Table 1 are typical for the MSBR and were the bases for this study.2 Since the reactor concept is undergoing engineer- ing and physics evaluation, these data represent no fixed design and are subject to change as the studies progress. DESIGN CRITERTA The following discussion delineates ground rules and arguments for the particular choice of process and design used in this evaluation. Choices were made on the basis of existing knowledge and data. The study presented here is expected to verify basic assumptions or indicate Jjudicious alternatives. Basic Considerations One basic consideration concerns the fuel yield (the fraction of fissile inventory bred per year) which, for a breeder reactor, is inversely proportional to the total inventory of the reactor and chemical plant systems. This characteristic is essential to the design of an MSBR pro- cessing plant and suggests close-coupling of the reactor and processing plant to give minimal out-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. Thegse include mechanical equipment, compressed gases, heating and ventilat- ing equipment, electricity, etc. The cost savings for an integrated -8 - Table 1. Typical Characteristics of & Molten Salt Breeder Reactor General Reactor power, Mw (electrical) Thermodynamic efficiency, % Reactor geometry Core diameter, ft Core height, ft Blanket thickness, ft Moderator Volume fraction of moderator in core Volume fraction of fuel in core Volume fraction of fertile stream in core Reactor containment vessel Fraction of fissions in fuel strean Plant factor Breeding ratio Fuel Stream Composition, mole % LiF (99.995 at. % TLi) BGFE UF), (fissile) Inventory at_equilibrium Volunme, £t LiF, kg BeFo, kg 233y (as UFY), kg 235U (as UF),), kg Other U (as UF)y), kg Cycle time, days Power, Mw (thermal) Liquidus temperature, C Density (calculated) o (g/em3) = 2.191 - 0.000k t (°C) for 525 < t < 1200°C Fertile Stream Composition, mole % LiF (99.995 at. % [Li) ThE), Inventory at equilibrium Volume, ft3 LiF, kg Th (as ThF)), kg 233y (as Ugu), kg 233pa (as PaF)), kg Cycle time, days Power, Mw (thermal) Liquidus temperature, C Density (calculated) o (g/em3) = 4.993 - 0.000775 £ (°C) for 565 < t < 1200°C Steam Conditions Pressure, psia Temperature, I Condenser pressure, in. Hg abs ® 1000 L5 Cylinder 8 Ly 3.5 Graphite 0.79 0.15 0.06 INOR -8 0.972 0.8 ~1.08 68.5% 31.22 0.31 3515 1000 1.5 Fpasic composition of carrier salt is 69-31 mole % (LiF-Bng). Equilibrium composition for this cycle. -9 - facility are immediately apparent when one considers the large amount of equipment and facility duplication required for separate plants. A further basic consideration is that there will be no large extrapolation of technology in the process design. Accordingly, the process is based on treating the molten salt by fluorination and distilla- tion, with the supporting operations of UF6 sorption on and desorption from beds of pelletized NaF, followed by reduction of the UF6 to UFM' A large amount of dats is available for each step except for the distilla- tion and reduction operations, both of which have been demonstrated in the laboratory.3’26 However, this study does assume the necessary engineering extrapolations to convert from the current batchwise operations to continuous operations. Process for the Fertile Strean The two streams of the bfeeder reactor require different processing rates and must be treated separately to prevent cross contamination. The first step in process for the fertile stream consists only of continuous fluorination, which removes the bred uranium as the volatile hexafluoride. No other treatment is needed if this step is designed to maintain a low uranium concentration. To accomplish this, the stream is required to go through the processing plant on a relatively short cycle, for example, once every 20 to 50 days. The cycle time for this study is 22 days. The fission rate in the blanket is low, and the fission products are kept at a tolerable level by periodic discard of barren LiF-ThFM salt. A 30-year discard cycle suffices. 1In the second step, the volatilized UF6 is sorbed on NaF beds, desorbed, and finally caught in cold traps. Process for the Fuel Stream The fuel stream of the reactor is processed by fluorination and vacuum distillation to recover both uranium and carrier salt sufficiently decontaminated of fission products. A cycle time of 40 to 7O days is required to maintain the fission-product concentration at a low enough level for attractive breeding performance. The calculations of this study - 10 =~ are for a 58-day cycle. The UF6 is recovered by NaF sorption and cold traps, Just as for the fertile stream. Decontaminated fuel and carrier are recombined in a reduction step that converts UF6 to UFM and. further purifies by reducing corrosion products (iron, nickel, chromium) to their metallic states. Makeup fuel and carrier are added at this point, and the stream is returned to the reactor. The time that the fuel stream spends in the processing plant is kept as short as practicable to minimize out-of -reactor inventory. Waste Storage The chemical plant provides its own storage system for process wastes. Incidental wastes, such as slightly contaminated aqueous solutions and flush salts, are assumed to be handled by the reactor waste system, thus such facilities are not duplicated for the processing plant. OSeparate storage is provided for fuel-and fertile-stream wastes, which are primarily LiF plus fission products, and Li¥F -ThF,, respectively. The facilities are designed for a 30-year capacity and afe located underground a short distance from the chemical processing area. The fuel -stream process also produces a less radiocactive waste than the LiF~fission-product mixture. This waste is in the form of pellets of sodium fluoride and magnesium fluoride pellets used for decontaminating uranium hexa - fluoride. Interim storage of 5-year duration is provided for these solids. Fission-product decay heat is removed either by forced air or natural convection, as required by the heat load. Operating Policy Certain ground rules consistent with convenlent and safe operation were adopted for this study. Maintenance operations are facilitated by assigning unit-process steps to either a high- or a low-radiation level cell. Operating and maintenance personnel, who are not required on a full - time basis, are to be shared with reactor operation. No water or agueous solutions are to be admitted to The process cells; fluids required for heat ‘I’ transport will be either air or sodium-potassium eutectic (NaK). A barren - 11 - fluoride salt, for example, an NaF -KF mixture, would be an acceptable substitute. Process Data The primary concern of a processing cycle for short-cooled fluoride mixtures is in dissipating fission product decay energy so that process operations can be controlled. A maximum of about 6,5% of the total energy of the system is associated with beta and gamma energy in the fission products; this amounts to about 140 Mw (thermal) in the MSBR fuel stream. Most of this energy i1s emitted quickly, decreasing about 82% in 1 hr and 95% in 1 day. The data for this study were calculated for the reactor described in Table 1, using the PHOEBE Code, which computes gross fission - product decay energy as a function of exposure and time after discharge from the reactor. The data are presented graphically in Figs. 2a-b and 3a-b for fuel and fertile streams, respectively. The graphs give an upper limit for heat generation because the calculations do not account for possible intermittent reactor operation attributable to the 80% plant factor. In addition, the graphs include decay energy assoclated with gaseous products that are sparged in the reactor circulating loop and with those fission products that might depos;t on surfaces throughout the system. It is difficult to separate this energy from gross energy until more is known about the behavior of fission products in molten fluorides. A process flowsheet showing material balances for the fuel and fertile streams in the processing plant is included in Appendix F. DESCRIPTION OF PROCESS The processing facility must have the capability of removing the major portion of fission products from the molten fuel salt and returning the purified salt to the fuel system after necessary reconstitution with 233U and carrier salts. As to blanket-salt processing, the facility must achieve recovery of the major portion of the bred uranium for recycle to the fuel ‘stream or for sale. These goals can be met with present technology or with processes that can be reasonably extrapolated from current technology. ORNL DWG 65-2987 R1 A ER SRR IR I 1 i i 10 SEC w0 W oW @ ™ < (g, ™8) 31V¥ NOILV¥INIO 1¥3H o oy 2 " TIME AFTER DISCHARGE FROM REACTOR (days) in'MSBR Fuel Sfiream. i o s Md @] (TGN O g op~ a0 @ =} O =3 O —~ 3 = = & RE Ay IRy T = = >y g O O\G O i —Qd © o o o O & Oy 5 jien] 5 -lf [43] 0w o D B \O QW&, &8 &0 o (& stream vol ORNL DWG 65-2988R1 6 7 891 T = Tt rim ] Fr T 25 B 5 MANRRE] 30 -~ £ i 1 ) il 5 6 7 9]04 ? | it e A @ T 3 4 5 G0N © oW B o~ n,c—u._: nig) 3LV NOILV¥INIO LVIH 102 10! TIME AFTER DISCHARGE FROM REACTOR (days) Fuel - MSBR Fuel Stream. in a y Heat ; pover, 2160 Mw (thermal); cycle time, 58 days. Fission ~Product Deca 671 £13 Fig. 2b. stream volume, 7 891 ORNL DWG 65-2989 R1 T 6 7'-':_;.9 1 - 3 [ R e Fertile- | ile-sfiream;~_ ‘cycle time, 22 days. AR AR LR ot TIME AFTER DISCHARGE FROM REACTOR {(days) 5 6 7 8 5 i 1 il v Heat in MSBR Fert power, 62 Mw (thermal) e 2 . AR IRIRANARSL] Fission-Product Deca 1783 £t3 ] 3a. stream volume, oL 3 T M Fig R il TIHL b0 seC A_m,a?_f nig) 3LVY NOILY¥INO LV3H HEAT GENERATION RATE (Btu hr-]fl_s) 10 ORNL DWG 65-2986 R1 6 7891 9 EEEag i _: : ==asa8ss = = £ =2 == BEaT T 8 i L" :::;:— st 53t == =gt sieed & = = z . 7. == Eas o H S5 =S + ; = o eI SeSsR st Tk B S If‘ ‘ = f T & T T . 5L "u‘ E t = T T T - | 6 : ' ' : . = = i wal — II T - # i t ] S i i 1 ANNETY r - I I | 1 Hil I 4 H L N === H o S :A:::{;}tf:;;_ - H ve @ = ses H : = 1 1 : e ni == S = ”. R - i it 1 UREAD LAY 1o e B i i S : ; 2 -1 Tr I { i i f Il () | T 4 I L i i Eh.Y 1l T 1 T 1 I 1M | ir ' T EaREN i i i AR R e | i ! HHEREH ]0]1 . i 1 I | A 9. i £ ‘ ESEE; : = = HE 8 B : = 5 = = N =i e o B SEmaT Es EEHEE HH H 7 s = f SN R auaaa it T a1 ) o o o o 1 jEa s o T 8 ik t ] ! = i : Eheus + g : T FrifE ! — b i s : 5 & S | T ANER BN P T Bais b HARHE FEL e e 1 1 T It T AESR REN 1 s : Hr 4 T T T I I H F Tr 1 - 1 Yol == = T e H e : = e amnoSesee : = & = == : : ERs s 2 = Ho t 3 T t T T i T T .7 T b 1 3 E 2 L = FEERRLE R ERSREER Nt Lk ik ; : =5s: = EaaEsn T S i i T : e s " i e I Hi i HHHHE T ” pheit o i it it i Hit i e i r ¥ ! 1] I 1 ! ‘;é‘ . i ] | i [T TR i i k ] = I I R i 0 i CHI | i | L _1 i} 100} R T _ 1071 SRR E : 8 = = =i £ : : 2 == === e SSses k A 22t - T t T H asdsins b T 11 6 ' i , = Thi=es i i 5 - it e it T i £f I l‘ll ‘I f I 4 T } Hh : I ‘ ] } 1 M E e T i 8 I e S : . L = = : 3ifis rf e ? t + 3 = i e : ot . i ! i it = : HHER ! FE R i - f 2 iy L; : B - i 1 T Rd i | i M i 1 H HEsIHH 1 : ] | I | |||1 i ] ]|]| i HH HH fl i LI { T ) Y \q_'_rr . 30 5 !,] T L 101 E A 1 R A ARSI Hulnflfl 'rmmmmmmmfl*mrrm 1 mmmmmmmmnmmmmu VA0 840 RS RURRTARRD 3 4 € 8 9]02 5 6 7 89 5 2 4 7809 10 Fig. 3b. stream volume, 1783 ft3; 103 6 78 Yt TIME AFTER DISCHARGE FROM REACTOR (days) Fission-Product Decay Heat in MSBER Fertile Stream. Fertile- power, 62 Mw (thermal); cycle time, 22 days. <1 - 'Sfimmary of-the_Process Flowsheefi In Fig. b4 it can be seen that fuel—stream proce851ng is achleved by 233 first removing the uranium (as UF ) and volatile fission products from a side-stream of the molten salt by fluorlnatlon. Then the carrier salt is vacuum-distilled frbm the residual fission products. Next, the UF6, further purified by a sorption-desorption process based on the use of NaF pellets, is dissolved in the carrier salt. TFinally, the UF6 is reduced to UFA by hydrogen, thereby reconstituting the fuel salt. Blanket salt is processed concurrently with the fuel salt by fluorinating a side-stream. The UF6 gas 1is separated from the volatile fission-product fluorides by sorption and desorption, using beds of NaF pellets as mentioned above. The blanket salt needs no further purification, 233 and a portion of the UF6 is sent to the fuel stream by dissolving it in the carrier salt and reducing it with hydrogen. The excess over that needed to refuel the reactor is sold. The two chemical reactions (fluorination and reduction) in the process are simple, fast, and quantitative. The other interactions are physical and require only heat and mass transport; however, in the case of distillation, a rather high temperature (about lOOOoC) is involved. The salt is extremely stable at any temperature anticipated in this process, and other physical properties, primarily vapor pressure and solubility for fission products, are in accord with process requirements. Fluorination As noted above, uranium and volatile fission products are removed from both streams by fluorination. A batchwise, molten-salt fluoride - volatility pfocess for recovering uranium has been under development for several years. Currently the process is in the pilofi plant phase at ORNL.lY’18 One of the steps in this process is batch fluorination of the molten fluorides to remove uranium hexafluoride. As noted before, uranium ~in these streams is in the form of UFl¥ | This uranium and some»of the flSSlon products are converted by fluorlne ‘to higher-valent, volatile - fluorldes Whlch leave the salt and go to the uranium hexafluorlde N ABSORPTION CoLD TRAP ORNL DWG &5-3016 NON CONDENSABLE GAS TO STACK SORBED FP IN NaF WASTE L BLANKET SALT REACTOR FPUFg,Fy | | ] L= ¢ /2] ._ ud FLLUORINATION % 3 m [} u & o - 0. SALT WASTE STREAM DISTILLATION SORBED FP iN NaF WASTE FLUORINATION BARREN SALT STILL BOTTOMS TO WASTE ot H2 HF, FP Fz FP OFF~GAS TREATMENT CAUSTIC SCRUBBER PURIFIED FUEL SALT CAUSTIC WASTE I} CAUSTIC MAKE-UP < FUEL SALT ] Fig. 4. Major Steps in MSBR Fuel and Blanket Processing. separated from blanket and fuel salts by fluorination (blocks 1 and 2), further purified by sorption, and collected in cold traps. MAKE—UP Uranium is The barren fuel ~carrier salt is distilled away from the bulk of the fission products (block 3). The cold-trapped UFg and fuel-carrier salt are recombined in reduction step (block 4) and returned to the core. LT - purification system. The reaction'for_producing UF6'in the present_batch processing 1ls the same as that_for.the proposed continuous system and may be represented as follows: flFh + F2 w>GF6 . Certaln fission- product fluorldes are also volatile. The principal ones are: Ru, Nb, Cs, Mo, and_Te. Z;rconlum is volatilized to a lessér degree. RIS o - o Contlnuous proce881ng of the MSBR fuel salt can be best achleved 1f the flu@rlnatlon process 15 made contlnuous, preferably W1th counter- : . current contactlng of the molten salt and fluorlne. Such a_process_has Vnot'béefi_deyéloped;.and the @resefit_volatility process at ORNL is entirely a a batch proééss. Tberefore} dévelopment work will be necessary to provide for continuous fluorination. - o o _ Rate data for molten—salt fluorlnatlom are limited and conflicting, . aithough the reported rates_have been sufficient for the batch_process, 19 -Mallen 8 data bn the flubrination'of falling'droplets of molten salt support the view that the reaction itself is very fast; whereas, the . slower rates that result from sparging a pool of molten salt with fluorine .(Cathers_gz_§£:,20 Pltt,gl and Moncr1ef2 ) can be explained by assuming a massatransfer—controlling rate mechanism due to inefficient mixing of the _gas_and liquid phases. Contact times (fluorine and molten salt) of 1 to 2 hf:have been.shown to be adequate for uranium removal down to 10 ppm in su¢h'batch tests. The countercurrent, continuous operation, énvisioned here, wéul@ probably give better contact. It is difficult to predict.the fluofiné'utilization in a continuous fluorinator; however, it should be better'than that for batch fluorination, which suggests that a utilization of . 33-1/3% will probably be achieved. A relatlvely low mass Tlow rate of gas must be malntalned in the fluorinator to prevent salt entralnment in the off-gas. The highest mass flow rate that has been used successfully w1thout enfiralnment 1s 0. 28 Slpm per in. 2 of fluorlnator cross sectlon 23 | Experience in the Volatility Pilot Plant at ORNL showed that - corrosion on bare metal walls (L-nickel) in the_fluorinator.1s_relatively 'fhigb. Kessie §§_§£:2M-developed_a technique for keeping a frozen -19 - protective layer of salt on the metal wall, and we adopted their approach for this study. This frozen wall of salt, which is kept so by a high cooling rate on the metal wall, prevents gross corrosion. Internal heating keeps the rest of the salt molten. For the fuel stream, ficsion- product heat is sufficient for this purpose, but for the blanket stream it may be necessary to supplement decay heat by using suspended electrodes for resistance heating. Since a maximum of 1.5 kw of heat per ftg of wall area can be removed from such a system, the fuel stream will have to be delayed in a cooling tank until the heat generation due to decay heat is at a sufficiently low level. As to the blanket salt, decay heating will be ingignificant. The continuocus fluorinator is a tall column into which molten salt is fed at the top and flows to the bottom; fluorine is introduced at the bottom and passes to the top, accompanied by thé volatile fluorides (Fig. An expanded deentrainment section is added to the top, and the body of the column is Jacketed with a coolant to maintain the frozen salt wall. A gravity leg is used in the molten salt outlet to hold a constant salt leve in the column. Purification of Uranium Hexafluoride by Sorption and Desorption The UF6 from both streams is purified in the same way before being returned to the fuel stream. OSince the UF6 that leaves the fluorinator contains volatile fission products, it is purified by a series of sorption steps. These are batch steps, but the process is made continuocus by using parallel beds alternately. The first separation occurs in a NaF absorption system where the gas stream passes through fixed beds of NaF pellets. This system consists of two distinct zones, one held at L00°C and one at about 100°C. In the higher -temperature zone, corrosion products, entrained salt, and niobium and zirconium are irreversibly removed from the fluorinator off-gas while the UF, and other fission products pass through (Fig. 6). In the second NaF zone (lOOOC), UF6 and some of the molybdenum are held up by sorption while the remaining fission products pass through. At this point, the UF6 has been separated from all the fission products except molybdenunm. 5)- 1 - 20 - ORNL DWG 65-3037 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 attack by the molten-salt—fluorine ‘ mixture. F2, Ru, Te, Cs F2, UFg, Mo 100°C DURING SORPTION, < NaF 150°C DURING ABSORBER DESORPTION \ 400°C FLUORINATOR ey OFF-GAS OR Fy ) v NaF WASTE (Nb,Zr) Fig. 6. Fission Products. one for the fertile. ¢ ORNL DWG 65-3015 F2 < Jl NaF I¢ UF, TRAP SECONDARY COLD TRAP, prk ~60°C ABSORBER PRIMARY COLD (150°C) TRAP, -40°C UFg, Fo o N ;_I ¥ MgF2 UFg TO WASTE REDUCTION UNIT (Mo) Purification System for UF6 with Disposition of Volatile Two such systems are used, one for the fuel stream and - 20 Molybdenum fluoride is removed from the UF6 by isolating the 100°¢ ‘ NeF zone, desorbing the UFg and molybdenum fluoride (by raising the temperature to >lBOOC and passing fluorine through the bed), and passing desorbed gas through a fixed bed of MgF,, which is held at about 150°¢C. The MgF, sorbs the molybdenum but allows the purified UF, to pass through into the cold traps. The NaF pellets used in the high-temperature zone must be replenished periodically since they accumulate fission products and corrosion products. This discafd constitutes one of the waste streams. The two NéF sorption zones may be integrated into a single unit, one zone on the other, and, as Nal' is discharged from the lower zone, it can be replenished by Nal' from the low-temperature zone, which is in turn fed with fresh NaF pellets. Such a system has been used effectively in pilot plant operations, and a similar system would be desirable for the MSBR processing facility.l7’18 In our concept, the system has a movable bed of NaF pellets, and, after each sorption cycle, some of the lower NaF is mechanically ejected to waste. Annular design with alr cooling would probably be necessary to allow removal of fission product heat (Fig. 7). As the UF6 leaves the last sorption trap it must be collected and ultimately used as feed for the fuel -reconstitution step. This is done by collecting the UF6 in cold traps. Two cold traps are connected in series. The first, or primary trap, is operated at about -hOOC, and the second, or backup trap, is operated at about -6OOC. The principal design consideration is the heat transfer rate. Conventional designs are avail- able for such units in which there are internal cooling fins for collecting UFg, and Calrod heaters for vaporizing the UFg for removal (Fig. 8). Uranium hexafluoride from the heated cold trap is fed directly to the reduction process. This calls for at least three primary cold traps for continuous operation: one for collecting UF6, one for feeding UF6, and. one in transition between these two functions. A fixed bed of NaF at ambient temperature very effectively removes trace amounts of UF6 from gas streams. Such beds are used as backup UF6 traps in the fluorine exhaust from the cold traps and in other process streams that might; contain UF6. Uranium is recovered from such traps by . using the NaF as charge material for the main absorption beds. - 23 - ORNL~LR~DWG B50451R-3 YNOF CHARGING CHUTE 1Y5-in. NPS, SCHED -40 ( AIR COOL- ING AND THERMOGOUPLES ) TO UFg TRAPS =—p DESORPTION CYCLE 5-in.NPS, SCHED-40 INCONEL 100°0OR 400°C. TRANSITION ZONE / INCONEL-X PISTON 5-in. NPS, SCHED-80 INCONEL HYDRAULIG CYLINDER 1! ~ 35° WASTE RECEIVER Fig. 7. Movable-Bed Temperature-Zoned Absorber. When the lower zone of the bed becomes loaded with fission products, the hydraulic cylinder operates the piston to discharge that portion of the bed into the waste carrier. Fresh NalF is added at the top. This apparatus has already been tested in the ORNL pilot plant. ORNL-LR-DWG 1909! R-I QUTLET THERMOCOUPLE OUTLET END HEATER FILTER CARTRIDGES — REFRIGERANT TUBES (4} 5-in-SPS COPPER PIPE 12 0 12 24 Ewm—m_mom) ‘ SCALE (in.) This design has already been Fig. 8. Cold Trap for UFg Collection. successfully used in the ORNL pilot plant. - 1—(8_ - 25 - Since there is excess fluorine in the fluorinator off-gas (33-1/3% utilization), fluorine is recycled. This recycle contains some fission products, so it is necessary to remove a side-stream (10%) Lo prevent their buildup. Fresh fluorine is used for the desorption step, for all processing in the blanket section, and for fluorine makeup. Vacuum Distillation The vacuum distillation step applies to the fuel stream only and is used to separate the carrier salt from the fission products after the uranium is removed by fluorination (see above ). The LiF and BeF2 volatilize, leaving fission products in the still bottoms. This residue consists largely of rare-earth fluorides. If the relative volatilities of the fission products, compared with the volatilities of LiF and Bng, are low, then a good separation can be achieved in a single-step distillation without rectification. At the average operating temperature (sbout 650°C) of the MSBR itself, the solubility of rare-earth fluorides in fuel salt is only a few mole percent; however, at lOOOOC, the solubility in LiF alone is about 50 mole % for the more insoluble compounds, for example, L&F3, Pr¥F_, and CeF Other rare-earth (EE) fluorides have even higher solubilities 2t this teiperature. Physics calculations on the 58-day-cycle MSBR indicate that at equilibrium the molar ratio of L:L'F:(RE)F3 in the fuel is about 1400, a number consid- erably greater than the 1:1 ratio permitted by the solubility limit at 1000°C. It is therefore apparent that, based on solubility data alone, distillation at about 1OOOOC can tolerate an extremely large rare-earth concentration factor before precipitation occurs. The design of the distillation unit is concerned primarily with determining the appropriate configuration that will permit a large fission-product concentration factor and, at the same time, provide adequate heat-removal capability for the short-cooled Tuel. The still design developed for this study is shown in Fig. 9. It is charged initially with L ft3 of Lil that has tThe same isotopic composition as that in the reactor fuel; this volume Tills the tubes to a depth of about an inch above their tops. The pressure above the LiF pool is reduced p) - 26 - ORNL DWG 65-1802R2 34" ID - j «—-‘fw\ AIR COOLANT |~ B il fiy— | AIR OUTLET * NaK OUTLET © 817 TUBES 1/2" x 16 GAUGE INOR-8 1/2" WALL 1/2" INOR-8 TUBE SHEET W FUEL SALT , INLET e==xo) =& NoK COOLANT LiF WASTE DRAIN - 30" 1D - Fig. 9. Vacuum Still for MSER Fuel. Barren fuel-carrier flows continuously into the still, which is held at about 1000°C and 1 mm Hg. LiF -BeF, distillate is removed at the same rate that salt enters, thus keeping the volume constant. Most of the fission products accumulate in the still bottoms. The contents are drained to waste storage when the heat generation rate reaches & prescribed limit. This concept of the vacuum still has not been tested. ' - 27 - to 1 o 2 mm of Hg by evacuating the product receiver (see Fig. D-3 in Appendix D), and the temperature is adjusted and held at sbout 1000°C. Fluorinated fuel salt 1s continuously admitted to the LiF pool in the still, and distillation is allowed to proceed at the same rate as the inlet fuel rate so that there is no net volume change. The operating principle 1s to allow the rare-earth fission products, which have a much lower vapor pressure than either LiF or BeFE, to con- centrate in the still. The liguild pool in the still reaches an equilibrium concentration in LiF and BeF and the Two components then distill at the P rate at which each is enteriig the still. The condenser is a conical region located Jjust above the evaporating surface; it is kept at about 85000 by forced convectlion of heated air. Distillate 1s collected in a circular trough and drained to a product receiver. The still is operated in this way until the heat-generation rate due to fission-product decay reaches the heat-removal capability of the NaK cooling system. This point occurs after about 67 days of continuous operation, at which time the fission product concentration in the still is about 1& mole ¢, a value considerably less than the approximately 50 mole % solubility limit at 1000°C. At this time, the contents of the still are drained to a permanent waste receiver, and the cycle is repeated. These calculations are conservative since they are based upon gross fission-product heat release and do not subtract the effect of Tthose fission products removed or deposited in the reactor before chemical processing. The attractive feature of carrying out the distillation in this way is that it minimizes the volume of expensive LiF relegated to waste. Since distillation is carried out after fluorination, less than 0.1% of the uranium removed from the reactor should enter the still, and at 1000°C the vapor pressure of UFM is fTavorable to the recovery of a significant portion of this fraction, reducing the overall uranium loss. Beryllium fluoride losses should be insignificant because the vapor pressure of Bell, ot 1000°C is about 100 times that of LiF. Distillation probably will not effect decontamination from CsF and RbF; each of these com- pounds has a greater vapor pressure than either BeFE or Li¥. Because rare earths are concentrating in the still as a function of time, their decontamination facltor in the product will decrease with time. It is not possible with existing data Lo asgess the magnitude of this effect. - 28 - Detailed calculations for the still design are given in Appendices C and D. Vapor pressure data for principal components of the fuel streanm are included in Appendix F. Distillate from the still is collected in an evacuated tank operating at still pressure. When filled, the receiver is isolated from the still, and the LiF -BeF, mixture is transferred by gravity flow or pressurization to the reductioi and fuel -makeup operations. An in-cell waste receiver is provided for the initial cooling of the still residue before transfer tb the underground waste-storage facility. The tank has a b-fto volume, allowing a one-cycle delay (about 67 days) inside the cell where heat is conveniently removed by the circulating NaK coolant. During this delay, the heat generation rate decreases from 3.2 % 10/ Btu/hr to 6.8 x 106, imposing less stringent design requirements in the permanent waste recelver. The interim recelver is a shell-and- tube type, similar Lo the still; hovever, no condensing surface or provisions for air cooling are needed. Reduction of Uranium Hexafluoride and Reconstitution of the Fuel The combined UFé;streams,'that form the fuel sgalt and the Tertile salt, are reduced to UFM’ and only a sufficient amoung3§o maintain criticality is returned to the reactor. The excess U from the fertile stream is sold. The usual method for this reduction has been by reaction with the excess Hy in an HgéFg flanme: (H2 + Fg) . UF6 + H2 > UFu 4+ 2HF This reduction is carried out in a btall column into which UF6 and H2 are introduced into an HQAFE flame, and dry UFM powder is collected as the product. However, according to our proposal, a more convenient method for preparing UF) for the MSBR is by reducing the UFg to UF) with H, ggzgz_the UF6 is dissoclved in the molten salt. There are some experimental data indicating the feasibility of such a process; however, the kinetics of the absorption and of the reduction must be further investigated.26 It is possible that this two-step process could be carried out continuously - 29 - in a single column in which the molten salt flows upward, the UF6 is introduced and dissolved in the bottom of the column, and the H2 is introduced at an intermediate point to reduce the UF6 (Fig. 10). Some of the reconstituted salt has to be recycled to the column to provide enough dissolved uranium for proper UF6 gbsorption. Off -Gas Processing Most of the off-gas from the processing plant comes from the continuous Tluorinators; smaller amounts are formed in various other processing vessels. The gases are processed o prevent the release of any contailned fission products to the atmosphere. Excess fluorine used in the fluorinators is recycled through a surge chamber by a positive displacement pump, and a small side-stream of the recycling fluorine is sent through a caustic scrubber to prevent gross buildup of fission products. Each of the processing vessels and holdup tanks have off -gas lines which lead to the scrubber for removing HF, fluorine, and volatile fission products. The scrubber operates as a continuous, countercurrent, packed hed with recirculating agqueocus KOH. A small side-stream of KOH solution is sent to waste, and the scrubber off-gas is contacted with steam to hydrolyze fission products such as tellurium. A filter removes the hydro- lyzed products. The noncondensable Tission products are sent to the off- gas facility for gases generated by the reactor. Wagte Storage Four waste streams requiring storage leave the processing facility: (1) NaF and MgF, sorbent from the UFy purification system, (2) aqueous waste from the KOH scrubber, (3) molten-salt residue from the distilla- tion unit, and (4) molten salt from fertile stream discard. The aqueous waste stream is small, and it is assumed that adequate capacity exists in the system for storing reactor waste. The other three wastes are stored in separate underground facilities adjacent to the procegssing cells. Since the values in the waste from the fertile stream - TLi, thorium, ORNL DWG. 65-3036 Fig. 10. Continuous Reduction Column. Barren salt and UFg enter the bottom of the column, which contains circulating LiF—BeF24UFu. The UFg dissolves in the salt, aided by the presence of UF), and moves up the column where it is reduced by hydrogen. - Reconstituted fuel is taken off the top of the column and sent to the reactor core. Preliminary data indicate that this design is promising. - 31 - 2 and 33U -~ will be worth recovering at some future time, some very tentative ideas about how they may be recovered are presented at the end of this section. Nal and MgF . Pellets Spent NalF and MgF2 pellets, which retain the volatile fission and corrosion product fluorides from the UF6 gas stream, are stored in 1lO-in. - diam by 8-ft-high stainless steel cylinders in a concrete vault adjacent to the still-residue waste tank. (See Dwg. No. 58080D in Appendix F.) The cylinders are loaded inside the processing cell and transferred to the underground area at approximately 90-day intervals. The vault is designed to contain a 5-year collection of cylinders (160 cylinders at 100% plant factor); after 5 years, the older cylinders are removed and transferred to a permanent underground storage site. The integrated heat generation when the vault contains 160 cylinders is about 1.73 x 106 Btu/hr. Forced circulation of about 11,300 scfm of air at a temperature rise of 12500 is used to remove this energy. The containers are constructed with a hollow core, allowing coolant to pass through the cylinder as well as over the outside. Aqueous Waste from Off -Gas Scrubber This waste, fission products in a strong sclutlon of KOH, will be stored along with other aqueous wastes from the reactor system and reprecsents an Insignificant contribution to the total amount of aqueous wasgte. Fuel -Stream Waste Residue from the vacuum still is stored in bulk in a facility similar to one evaluated previously by Carter and Ruch.lh A single, large tank equipped with adeguate cooling tubes and adequate for a 30-year accumula - tion of waste 1s provided. The 30-year capacity was chosen since that is the expected lifetime of the reactor. After filling, the salt might remain in the tank for additional decay or be disposed of by whatever method is currently acceptable. - 32 - Decay heat is removed by forced air convection. The heat load (Btu/hr) continually increases over the filling period but decreases rather sharply when no further additions are made (Fig. E-1, Appendix E). The time behavior of the integrated specific heat generation rate (Btu ™t ffi"3) for a b-year collection period is shown in Fig. 11. This 1is a smoothed curve for L-ft° additions every 67.4 days, followed by an extended decay period during which no waste is added to the tank. The upper portion of the curve might be extrapolated with little error to accommodate longer filling periods. Filgure 11 shows the spedific heat generation for the case of no dilution with inert salt; however, during the initial stages of filling, it is necessary to add an inert diluent, for exampie, NaX' -KF eutectic mixture, to lower the specific rate to a tolerable value. It was calculated that 26k ft3 of diluent is required for the 520 ft3 of LiF ~-fission product residue to be collected over the 30-year period. The fission products being collected exhibit the decay behavior shown in Fig. 2, which is repregsentative of gross fission products and does not account for those that have been removed by processing or other mechanisms. Fertile -Stream Waste The fertile-stream discard is also stored in a large underground tank, adequate for 30 years of waste collection. The tank is 13.5 ft in diameter and 13.5 ft high. Since uranium is removed from the blanket on a rather fast cycle, the fission rate in the blanket is low, making the waste activity several orders of magnitude less than that of the fuel waste. Cooling is provided by natural air convection around the tank and through cooling tubes. All metal surfaces are expected to be coated with a layer of frozen salt that will furnish excellent corrosion protection. The integrated heat producfion rate due to fission products for the 30 ~year period is 5.9l x 3_0lL Btu/hr. Since this epergy is associated with 1783 ft3 of LiFmThFu mixture, the moderate specific rate of 33.1 Btu hrwl ft"3 presents no design problems. When first removed from the 1 reactor, the heat production rate of the waste is about 1600 Btu hr’ ft-3, but this value decreases by a factor of 10 in about 4 days. -3 ft ORNL DWG 65-1844 R2 -1 SPECIFIC HEAT GENERATION RATE (Btu hr E = 5-YEAR HEHEEEEL e i COLLECTION PERIOD il : %” e L * e I ] i | e j T il i i E e 1 ' == 10 = | 1 YEAR L 2[5 ST By 15 130 B0 i o0 | , I mlgjfs il I . i T e EMHIE] 1 i 5 L] RRAE NG R N 4] I (E) ] EWREAIR . i s slizilells 10 108 10 10 10 TIME AFTER DISCHARGE FROM REACTOR (days) Fig. 11. Specific Heat Generation Rate of Fuel-Stream Fission Products in Waste Tank. Waste is accumulated in 4 ft° batches every 67.4 days. It is then held in the processing cell another 67.4 days for further cooling before draining to the waste tank. The reactor operates on a 58-day cycle at 2160 Mw (thermal). ¢e - e Cooling System for the Waste As mentioned above, the waste-storage system_is designed for cooling by forced air draft. Air is easy to handle, COmpatible with construction - materials for extremely long tlmes, and presents a minimum hazard 1n case. ~of contamination by the waste. Blowers, capable of supplylng 76 000 scfm at a pressure drop of about 30 1n. water, are located upstream from the waste vaults The alr is forced through the vaults and coollng tubes in the Waste tank and 1is exhausted to the atmosphere via a. tall stack, which also dlSposes of gases exhausted from the reactor system The ex1t duct contains the necessary radlatlonamonltorlng 1nstruments and absolute ' fllters for remov1ng partlcles -Pés’sible'Ultimate Treetment of Waste from Ferfiile'smeam : Slnce a 30 -year accumulatlcn of wagte from the fertlle stream w1ll 33U, and thorlum), not too highly contain recoverable values (TLl,: contaminated with fission products, it may be worthwhlle_to consider a - recovery system before relegatlng this waste.té permahent burial. Afiy. significant uranium value would probably be.recévefed.by fluorination, but the recovery of thorium and lithium requires:further process develop- ment. A potential method for Li-Th separation is the.incompletely 15 investligated HF -dissolution process, based on_the:principle of leaching LiF from THFA and rare-earth fissiqn products.with_ahhydrous hydrofluoric acid. This process, however, leaves the thorium contaminated with fission products, making it necessary to resort to an aquebus system (solvent extractlon by the Thorex process) or to develop a thorlum recovery process that utlllzes fluorlde chemlstry From & purely economlc v1ewp01nt, thorium would be retalned in the waste tank until its recovery became more economlcal than mznlng new thorxfim On a 30-year dlscard cycle, an entire fertlile -stream lnventory of : ThFh and LiF will aecumulate 1n the waste tank. In addltlon; the waste 33 233, will contaln "Pa, and flSSlon products in amounts that depend on the breedlng ratlo, eff1c1ency of the fluorination step, and the ‘blanket pOWer. The uranlum loss is based upon a. 90% efficiency in fluorlnatlon, A value belleved to be conservatlve,_ The_largest loss of fissionable - 35 - material, however, is through protactinium, which we assume to be non- volatile as the fluoride and which consequently is discarded in direct ratioco to its concentration. The amount of fission producis is calculated for a blanket power of 62 Mw (thermal) and an 80% plant factor; to account for those fission products that have volatile fluorides or which are plated out on parts of the system, a nominal figure of 80% is used for the fraction that finally reaches the waste tank. The 30-year inventory of the waste tank is shown in Table 2. Table 2. Inventory in Fertile-Stream Waste Tank After 30 Years of Collection Waste volume = 1783 ft3 Amount Unit Value Value (kg) ($/ke) (10° $) Th (as ThFu) 141,200 10 1.h41 14 (as LiF) 10, 400 120 1.25 233y 4 233pg (as UrF), + PaFM) 116 12,000 1.39 Fission products {as fluorides) 450 - 4.05 In view of the figures in Table 2, the design presented here for the fertile-stream waste system is not optimal. Nearly 99.5% of the 233U + 233Pa 233 value 1in the waste tank is attributbtable fo Pa discard; this loss can be reduced to negligible proportions by providing in-cell decay storage 233, followed by refluorination to recover the daughter 1-year holdup (about 60 ft3) would allow more than 99.995% of the 33pg, o 233 U. . For example, a decay to It probably would not be necessary to provide additional fluorination eguipment for this volume because the refluorination could be scheduled in existing equipment during reactor shutdown. Flow Control of the Salt Streams Flow control of the molten salt streams is by freeze valves coupled with a controlled pressure drop. This can be achieved by the simple freeze valve currently used in the MSRE (Molten Salt Reactor Experiment), - 36 - coupled with a flow restriction such as an orifice or venturi. A dynamic freeze valve in which a controlled layer of salt is built up in a cooled section of the line is bveing investigated; if successful, it will allow a grester eage of flOW'control.ET Removal of Decay Heat Heat removal is a major problem in all process vessels that contain short-cooled, highly irradiated fuel. In many cases, the heat flux and operating temperature will be high, making it difficult to use water or | air as primary coolants. Water has an additional disadvantage in that 1t is incompatible with the process fluids, creating a hazard should there be a leak in the salt lines. Therefore, a sodium-potassium eutectic mixture, 22.3-77.7 wt % Na-K, was chosen as the primary coolant for process vessels at temperatures above SOOOC where large amounts of decay heat must be removed. This.codlant also has the capability of initially heating the system to 600°¢ for startup purposes. Air was chosen as the coolant for low heat fluxes at temperatures less Than 5OOQC. Sampling the Salt and Off-Gas Streams A rather complicated mechanism is required to remove analytical samples from a highly radioactive molten-salt system. A technique similar to that tested for the MSRE will be used,EB and the off-gas streams will be sampled conventionally. Shielding, Maintenance, and Repalr of Equipment All process equipment that handles material that approaches the radiation level in the reactor core will be shielded by about the same amount of shielding as for the reactor, and maintenance will e indirect. Process vessels in this area needing repair will be removed and sent to & decontamination facility before repair. These include the fuel-stream delay tanks, fluorinator, sorption-descorption systems, and distillation system. All other equipment for processing radioactive materials is - 37 - contained in an area of direct maintenance with less shielding. In the direct-maintenance area, decontamination for maintenance 1s achieved by a molten-salt flush and an aqueous wash. Materials of Construction All process vessels and lines in contact with molten salt are made of INOR-8 except the fluorinatocrs, which are made of a special material, lloy 79-&.28 Vessels and lines'that contain Tluorine or fluoride -bearing gases are made of Inconel or Monel, and cold traps are made of copper. Other equipment is made of appropriately compatible material. General Operating Policy The overall policy for operating the [luorinator, vacuum still, NaF beds, and related equipment is based on the projected simplicity of opera- tion and small slze of the equipment. The system is designed for a campalgn -type operation of one month's duration, without shutdown except for emergency maintenance. There will be no access to the operating areas during this period. At the end of the operating period, the entire systenm will be closed down, routine maintenance accomplished, feed hoppers replenished, accumulated waste transferred to waste storage, etc. The operating cycle is then repeated. PROCESS DESIGN Process equipment for this design study was patterned as much as possible after previously designed and tested equipment. Fach major component was studied for its application to the continuous processing requirements of the MSBR to ensure a practical design. Detailed designs were made to the extent that overall size, configuration, heat transfer requirements, flow rates, etec., were defined %o allow a reasonable estimate of equipment cost. The waste system was considered separately, and liberal use was made of a previous, similar :sfi:uéLyll‘L Tor its design and cost dats. - 38 - Fuel Stream Degign Criteria Design criteria for process equipment were hased on past experience. Equipment capacity was based on handling & flow rate 30% larger than that required for 100% continuous operation as specified by reactor physics calculations. Pertinent basic data adopted for this design are as follows: Salt flow rate: 15 ftB/day | Pressure: 2 atn | o o Density of pelleted sorbents, NeF and MgF,: 1 g/cc Temperature range for NaK .coolant: As coolant: 200°C . As heat source: 800 C o Coolant temperature rise in heat transfer operations: 300 C Normal convective heat transfer coefficient: 10 Btu he T rp e Opt Electrical heat needed: 1.5 kw per ft2 of longitudinal area for temperatures higher than 5OOOC Fission product heat: 50% of total is removed with volatile fluorides in the fluorination step Major Process Lguipment 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 in the processing flowsheet (Fig. 12) is listed below, with its purpose, design basis, and description. The ldentifying number accompanying the equipment name corresponds to the éircled numbers in the-figure. Design calculations that involve unusual technigues or complexity are shown in the agppendices. 1. Flow Control Purpose: The flochontrol device meters the flow of a molten salt stream to or from a process vessel. Description: A piece of process pipe, l/2-in., sched -40, 1 £t long. Pipe is Jacketed with 1-1/2-in. sched-bO pipe made of INOR-8 and has two coolant connections. ' 2. Coolaflt Tank Purpose: To allow delay for fission-product decay of the molten-salt stream from the reactor core before its introduction to the fluorinator. 39 - ANMONIUM FLUSH OXALATE SaLT AND WATER = TO PROCESS HOT Fg.FP COOLING Fp SURGE| |et \ 2 TANK J b sauEops T ASWERS (15)COMPRESSOR FLUSH F : ALT FLUSH o Fa.FP o MAKE-UP MAKE 1P ( ANK TARK Fo UG FP 4 7 FLI TO PROCESS Fp UFg FF ConTROL VESSELS FLOW NeF ™ CONTROL UFg TRAP TO PROCESS VESSELS — NaF —ag— Fadifg,FP g HEBTER Mg Fz*!' o, 70 SOREERS = 'S M3 o FP TRAP F2,UFg,FP b ié UF4 FP @ 4 NoF —={33) ‘;;" o & REFRIGERATION coro i i c & MgF FO TRAP 5 9 Ay £ I | TRAP w x o @ 5] HE@ER 5 © | s o o Ple F2 i REFRIGERATION ® . 3 s 3y D B g > - | = / i & e Lfiflfl c TO WASTE g > “ STORAGE & w = w ¥ & LiF, ThPg PRODUCT O 1 l CONNER ‘ g G 0000000 o N : MAKE-UP MaF WASTE TANKS AND =] \ TANK Al COOLING STATION el A - =4 FLOW {ne TER CONTROL. - Hp HF 0] ) || TO WASTE T X STORAGE Ha,XE £ @ BLANKET Hp FP wgnz FP HYDRATER TANK REACTOR 2 3 J ) TOBLANKET . _ TO WASTE My HF,UFg HEATER STORAGE <5 4 1 scausaNme UM M%KE-’;UP Efrve @ ® - 2 H He : > ¢ y NaF 7O ) L‘ SORBERS . oy VACOUM PUMP UFg ¥ , Hp Be = ' Supply Sopoly - Y TQ WASTE N Tank Tark . OWALI I MS,UFz STORAGE el Fa.FP k Process ‘ ‘ CONTROL > 7 et . NaF €9 @ ] @ o P Hot Coot 2 @ } . WAETE Nk N COLLECTION Co | HERVER g HOLD-UP Pry §2 £ ‘‘‘‘‘ 2 < . & 5 Eo | ur ® C°¥GC'=0" 5 2 il il - REFERENCE DRAWINGS RO. onk oy < =® 1 dem 0OAK RIDGE NATIONAL LABORATORY 3 ) . ! g@ KOH,FP TO OPERATED BY - ] ek WASTE STORAGE UnioN CARBIDE NUCLEAR COMPANY FUEL )% DMSION OF UNION CARBIDE CORPORATION E SgLT TO g ~ Pégp 0AK RIDGE, - DUMP TANK — LIMITS ON DIMENSIONS UNLESS |MSBR Integrated Processing Ptont BLOG. 4500 HF STORAGE £ O, ! pump JF STORAG €5 STORAGE @ : N GTHERWSE SPECEFIED: FRACTIONS 10, REVISIONS pATE JaPeD| APPD MSR FUEL AND BLANKET ARMER | 3-11-69 : i ANBLES % SUBIHTTED APFRGTED FPPROVED £ FATE IO | RS | DA ’ - - DATE | TTDATE | APPROVED | DATE SCALE: NO SCALE APPROVED I l 58025| D‘ 1 E) i Fig. 12. Major Components in Fuel- and Fertile -5tream Process. - bho - Design Basis: Type of tank: Healt exchanger with well-mixed contents Salt residence time: (See Appendix A) Temperature: 600°C Coclant: Nak L 1 -3 Heat load: 5.14 x 10 Btu hr — ft Material of construction: INOR-O 3 Description; A tank having 22.5 £t~ of liduid volume, with bayonet coolers and two pipe connections. 3. Fuel-Balt Fluorinator Purpose: To remove all uranium and volatile fission products from the fuel salt by continuous fluorination. Design Basis: Salt residence time: 2 hr Fluorine utilization: 33-1/3% o Maximm mass flow rate of F: 0.277 slpm/in. Temperature: 600°C Coolant: Nak i _ 3 Heat joad: 5.31 x 10 Btu hr ki Material of construction: Alloy 79-4 1 Description: A column made of 6-in. sched-40 pipe, 11-1/2 £t long, which is Jacketed with an 8-in. sched-4O pipe. Column is expanded above jacketed section to 8-in. sched-k0 pipe, 1 ft long. Column is supplied with' 5 kw of electrical heat and has 5 pipe connections. 4., Burge Tank Purpoge: To allow molten-salt surge capacity between the fluorination and distillation steps. Design Basis: Temperature: 600°C Surge capacity: 1 day's continuous flow (15 ft3) Coolant: NaK L 1 3 Heat load: L4.4 x 107 Btu hr — ft Material of construction: INOR-3 Description: A tank having 15 ft3 of liquid volume, with bayonet coolers and two pipe connections. 5. Movable-Bed Sorber Purpose: To separate some of the volatile fission products and corrosion products from UF6- Design Basis: Number of sorption zones: 3 Cooling load in high-temperature trap: 5.52 x 10” Btu/hr UFg load based on l2-hr cycle: 13.5 kg NeF loading: 0.5 g UF6/g NaF - b1 . Coolant: Air Average temperature: 400°C in bottom zone, 100 to 150°C in 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-40, 8 £t long; the inner pipe is 6-in. sched-Lo, 8 ft long with the bottom 30 in. finned. Bottom mechanical solids ejection; 3% kw of electrical heat. 6. NaF Supply Tank Purpose: To maintain a supply of NaF pellets for the movable-bed absorbers. Design Basis: NaF supply period: 30 days NaF usage: 10.75 kg/day Temperature: Ambient Material of construction: Stainless steel Description: A tank, l~1/2 £t in diameter and 3 £t high, with a conical bottom and four 2-in. star valves. 7. Na¥ Waste Tanks and Cooling Station Purpose; To provide short-term waste storage for the solid sorbents which contain sorbed fission products. Design Basis: Tank capacity: 107.5 kg NaF Coolant: Air Temperature: 36-in. diam, 3/L in. Equivalent vessel weight for entrances and exits: <100 1b vessel, 10 1b 100-1000 1b vessel, 20 1b . >1000 1b vessel, 40 1b Cost of electrical heaters: $100/kw installed 5 56 B 29, 32 Teble L., Summary of Fabrication Cost for Process Vessels Fabrication Cost ($/10) o _ - for a Vessel Welght of ‘Meterisl o <1000 1 >1000 1b Mi1d steel R S 2.0 1.00 Stalnless steel nlckel Monel, and Inconel L 3;50 ~2.50 INOR-8 | , - o - ko 3.50 Mloy 794 - © 5.00 4.00 Structure and Improvements . Structure and improvements costs for the processing facility w ere ~determined by assuming that these costs are directly related to corre- sponding costs developed for the MSCR by Sargent and Lundy Engineers. 'Adflition of the procesgsing facility proposed here called for an additional 16.56% of floor space in the reactor building, and this factor was to determine the cost from the corresponding MSCR data. The cost o 33 used T the crane and hoist was also increased by this amount. This additional space includes a 10% increase in analytical and decontamination facilitie The processing facility addition was 7.3% of the total building are SB a; therefore, cost of the grounds and stack was increased by 7.3%. The total increase in cost of structure and improvements due to the processing ~ facility was $550,770, which was obtained as follows: Structure 0.1656 x $2,932, 400 = $485,610 Crane and hoist | 0.1656 x 195,000 = = 32,290 Grounds = 0.073 x 501,500 = 36,610 ‘Stack | 0.073 x 31,000 = = 2,260 Total structures and improvements 8 o $556, 770 - - Interim Weste Storage The 1nterlmwwaste«fac1llty cost was estimated separately. _'Thi estlmate was based on previous cost estlmates for similar fac1lmt1e g . o 4,32 - 57 - Total waste facility cost was $387,970, with the following cost breakdown: Tank cost $213, 200 Excavation and backfill 32, 350 Concrete 18,070 Transfer lines 9, 000 Instrumentation 10, 000 Cooling system 100, 650 Structural steel 4, 700 Total $387, 970 Other Plant Costs Slgnificant cost items in the indirect-maintenance area of the plant are the process line "Jjumpers", which have remote connectors necegsary for indirect connection of pipelines, instrumentation lineg, and electrical lines. The following cost schedule, based on experience at ORNL, was used 2 for estimating the cost of these connection8:3 Major pipelines $1,500/1ine Multiple pipe and instrumentation lines $l,700/set Electrical heater connection, including the heater $2,000/set Other process-piping cost schedules were: Motor -operated control valves $500 each Coolant air ducts $10/f% MaJjor process lines in direct maintenance area (<20 ft) $400 each Gas lines in direct maintenance area (<20 ft) $200 each NeK coolant lines (<20 ft) $800 each The above cost schedule results in a total process-piping cost of $155,800. The electrical auxiliaries consisted of the electrical substation, - switching gear, feeders, and indirect connectors and Jjumpers. Cost schedules used for these auxillaries were:29 Electrical substation and switching $36.60/kw Overhead feeders $6.1/ft Underground feeders $12.2/f% The total cost of electrical auxiliaries is estimated to be $84, 300. Process instrumentation is estimated to be $272,100, radiation monitoring to be $100,000, and sampling connections to be $20,000. The costs of service lines and high-temperature insulation are based ‘I' on the installed cost of the process equipment in the main processing . 5 - areas-and in the waste storége facility. The cost of service lines, taken at 15% of the total installed cost of this equipment, amounts to $128,060; the cost of the insulation, taken at 6% of this installed cost, amounts to $51,220. Total Fixed Capital Cost The total direct plant costs are estimated to be $2,609,980 (Table 5). Past experience was used to determine percentage costs of indirect capital ' items. These percentages are higher than those for other chemical indus - tries but_re@resent actual cost expéerience in other ORNL projects.32 Constructiofi overhead is estimated at 30% of total direct plant cost to give a total construction cost of $3,392,970. Engineering and inspection at 25% of total construction cost is $8L83,2L40, which results in a sub- total plant cost of $h,2hi,216. Contingency at 25% of the subtotal plant cost is $1,060,300, and the total construction cost is $5,301,510. Inventory costs include the cost of the molten salt held up in the system and the initial cost of the NaK coolant. Total fuel-salt holdup for this system is 63 3. At $1,Aeo/ft3, the charge is $89,460. Blanket- salt holdup is 120 ft3, and at $56o/ft3 the cost is $67,200. TFuel- and -~ blanket -salt charges do not include the cost of fissile material. The 400 £t holdup of NaX at $lOO/ft3 costs $40,000. The total inventory cost is then $196,660, giving a total fixed capital cost of $5,L98,170. Direct Operating Cost The direct operating cost includes the cost of operators, chemical consumption, waste containers, utilities, and maintenance materials. The number of operating and support employees 1s based on a work schedule of four shifts of LO hr each per week. These include immediate supervisory, operating, maintenance, laboratofy, and health physics personnei, plus two people.for routine clerical and janitorial work (Table 6). No attempt was made to prorate the cost of higher supervisory, clerical, or plant protectlon personnel for the procesgsing facility since . some of this cost is included in labor overhead costs. _59... Table 5. Total Fixed Capital Cost Installed process equipment Structure and improvements Interim waste storage Process piping Process instrumentation Electrical auxiliaries Sampling connections Utilities (15% of installed process equipment) Insulation (6% of installed process equipment) Radiation monitoring Total direct plant cost Construction overhead (30% of total direct plant cost) Total construction cost Engineering and inspection (25% of total construction cost) Subtotal plant cost Contingency (25% of subtotal plant cost) Total construction cost Inventory cost Molten fuel salt (at $l,h70/ft3) Molten blanket salt (at $560/ft3) NeK (at $100/£%°) Total inventory cost Total fixed capital cost $ 853,760 556,770 387,970 155, 800 272, 100 8, 300 20, 000 128, 060 51, 220 100, 000 782,990 848, 240 1, 060, 300 89, 460 69, 200 10, 000 $2, 609, 980 $3, 392,970 $h, 241, 210 $5, 301, 510 $ 196,660 $5, 498, 170 ....60 - ! Table 6. Employment Costs | | Cost ‘No. ($/year) Production | ; . : Shift supervisor | b 32, 000 Operators o S 16 96, 000 .'Malntenance workers - 8 | L8, 000 Laboratpry analysts ok 24, 000 Health physics workers 2. 12,000 Others | 2 10, 000 Total 36 222, 000 The cost of utilities, waste contalners, and consumed chemicals is 'f:'based on a 300 day/year operation for both the reactor and processing plant. | | Total direct operating cost for one year is $610,190 (Table 7); this includes fuel and fertile salt makeup. Processing Cost The processing cost per year is estimated at $1, 447,570 (Table 8). This cost is obtained by combining the direct operating cost, the indirect cost of labor overhead (80% of direct labor cost), the fixed cost due to depreciation (10% of fixed capital per year), taxes (1% of fixed capital per year), and insurance (1% of fixed capital per year). The percentage used. for the 1nd1rect labor cost is arbltrary, however, it is w1th1n the range of usual practlce On the basis of 300 days of operation per year for the 1000-Mw (electrical) MSBR, the fuel -processing cost is 0.201 mlll/kwhr. The fuel- cycle cost is composed of this coéfi plfis the in-reactor inventory of fuel, fertile, and carrier salts, plus makeu@ fertile and carrier salts, and 233 less the credit for excess UF6 préduced In—reactor inventory, makeup, and credlt were not conSldered in thls study. - 61 - Table 7. Annual Direct Operating Cost Labor $222, 000 Chemical consumption Fluorine (at $2.00/1b) $ 4,080 KOH (at $0.10/1b) 980 Hydrogen (at $0.01/7t°) 720 HF (at $0.26/1b) 1,000 NaF pellets (at $1.00/1Db) 5, 780 MgF, pellets (at $1.00/1D) 420 Inert gases (guess) 830 Fuel-salt makeup (at $l,h20/ft3) 28, L0oo Blanket -salt makeup (at $560/ft3) 27, 350 NaX mekeup (guess) 830 70, 390 Waste containers , 28, 270 Utilities Electricity (at $0.01/kwhr) 73, 300 Others (guess) 7, 000 80, 300 Maintenance materials Site (guess) 2, 500 Building (at 2% of building cost) 10, 810 Service and utilities (at 4% of service and utilities cost) 35, 880 Process equipment (at 15% of process equipment cost) 160, 0Lo 209, 230 Total annual direct operating cost $610, 190 Table 8. Annual Fuel~Processifig Cost Direct operating cost = $ 610,190 Labor overhead (80% of labor cost) B | 177,600 Fixed costs - | L o Depreclation (lO%/year of flxed capital) o 5&9,820 Tax_(L% of fixed capital) | R 54,98@ _insfirance.(l% of £ixed capital . : '5h;980 Total B B : g1, W7, 570 Within the applled ground rules, these costs are belleved to be a _reasonably‘accurate representatlon of the cost for regeneratlng the fuel and blanket 1n an 1ntegrated MSBR processing plant. A more thorough o stbudy would have_lncluded_detalled design of equipment and layocut of the g._integrafied processing plant, the reactor and its auxiliaries. Such a _'thorough analysis was beyond the scope of this study. CONCLUSIONS AND RECOMMENDATIONS The central issues in this preliminary study were to analyze the feasibility and cost of a conceptual system for continuously regenerating the fuel and fertile streams in the Molten Sall Breeder Reactor. Briefly, the system consists in (1) fluorinating, distilling, and reconstituting the molten fluorides used in the reactor core, and (2) recovering the 233U from the molten breeder blanket by flucrination and using the uranium to reconstitute the core salt. The excess is to be sold. ~ The power of the breeder reactor was set at l,OOO Mw (electrlcal) for thls study . | | | | A number of basmc conclusions and essential recommendatlons were developed The conclu81ons relate to the projected fea81blllfiy and. | ant1o1pated costs in terms of establlshed technology and cost-accountlng pracfilces, and The recommendatlons refer to what must yet Dbe learned W1th_ reSpect to technology and. chemical data before a complete englneerlng . analy81s can be made It is our op;nlon that it w1ll.be.very_useful_to - 63 - begin filling these gaps in the knowledge because of the promising simplicity and low cost shown by this study. In the conclusions and recommendations presented below, the recommendations are underlined. The first Tour numbered paragraphs relate to general characteristics of this fuel-processing plant, and the others to specific unit operations. 1. FEASIBILITTY. Fluorination followed by distillation is a feasible process for regenerating MSBR fuel (LiF-BeF,-UF)). Fluorination alone is sufficient processing for the fertile salt (LiF -ThF),). Reactor physics calculations indicate that attractive breeding ratios can be obtained for such a process. Engineering problems in the processing plant appear to be amenable to solution through a well -planned developmental program at the unit-operations level. Fluorination-distillation should be developed as the processing method for the MSBR. Con- currently, other attractive processes should be investigated at the laboratory and/or engineering stage as potential alternatives. INTEGRATED PLANT. Integrating processing and reactor facilities is of primary importance in lowering the processing cost. Com- plete advantage is thereby taken of the ready adaptability of a fluid -fueled reactor for continuous processing with corresponding minimum inventory. The relatively small size of this side- strean processing plant, about 12 13 salt per day for a 1000 -Mw. (electrical) reactor, is amenable to integrated construction, thereby separating the economic dependence of the processing industry upcn a large amount of installed electrical capacity. The same {inancing convention that applies to the power plant applies tTo the processing plant; this type of financing is normally available at a lower rate than is available for a separated processing facility. ECONOMY. 'The estimated capital cost, excluding inventory, of the plant is $5, 301,000, and operating costs are about $788, 000 a year. It is significant that the capital investment in the integrated processing plant proposed here is only about W% of the total cost of the reactor system it serves. CORROSION. There are at least three areas in the chemical processing plant in which corrosion behavior of construction materials should be studied. These are the vacuum still, the reduction unit, and filters. The still temperature of about 1000%C is much higher than has been contemplated for any other part of the MSBR system, and the resistance of INOR-8 and nickel to corrosive attack at this temperature is not known. A reducer, hydrogen, enters the reduction unit and probably helps limit corrosion there, but this should be verified. Filters are subject to attack because of the large surface area exposed to the fluid being filtered. - 64 - FLUORINATION DEVELOPMENT. Batchwise Tluorimation of molten - fluoride salts for uranium recovery has been rather thoroughly investigated at ORNL; however, 1t is recommended that engineering development of a continuous fluorinator be given high priority. The need for continuous Tluorination is evidenced by the require- ~ment of low fuel and carrier ealt.inventories'in_the;proeessing" plant. The reactor fuel system contains about 650 ft3 of salt, and, without continuous fluorination, the out-effplle inventory could possibly be as large as the in-pile inventory. This study indicates that in a continuous fluorination-distillation process the holdup represents about lO% of the reactor Tuel : volume. - Itlsrammmmmdtmmimefmmahwfllcmmqm:mra contlnuous Tluorinator be developed and demonstrated. ThlS concept calls for a 1/2- to 3/%-in.-thick layer of frozen salt - on.the 1n51de wall of the fluorinator to prevent the serious Corr081ve attack by the molten mixture during fluorination. _Ba81c information needed includes fluorination rate data, process control in contbinuous operation, and method of establlshlng and _malntalnlng & frozen wall. Fluorlnatlon of the fertlle salt 1ntroduces probleme 81mllar to those encountered for, the fuel stream. However, fertile-stream processing rates are 8 to 10 times higher, and the flSSlOprrOduct activity is several orders of magnitude . less.. On the other hand, a lower fluorlnatlon eff1c1eney can be tolerated in blanket processing. Another method for continuous fluorination is the gas- phase continuous operation in which Tluoride microspheres are generated and fluorinated as they fall through a tower. This process should be recognized as a potential alternative to the - continuous method of fluorination studied here, but its development should be subordxnate to that of the frozen-wall concept DISTITIATION DEVELOPMENT. The vacuum-distillation concept for separating the LiF -BeFo (fuel-carrier salt) from fission products ~is feasible_from an engineering viewpoint. The theoretical net discard of 'Li in the still waste is low epnough that its cost is insignificant compared with other fuel cycle costs. Thermal ”7problems require that sufficient volume be malntalned for wetting ca 1arge heat ‘transfer surface, and the bulldup of flSSlon products in this volume will almost surely have an adverse effect on the deeontamlnatlon factor of the distillate. Relative volatility data are needed for the multlcomponent mlxture, LiF -BeFo ~figssion products, in which compositions are in the range 99 5~O 50 mol % to 84.5-0.5-15 mol % LiF ~BeFs-rare earth fluorides. It is | strongly recommended that & conbinuous vacuum Sbill be built and . operated to demonstrate a workable de81gn and o dbtaln rate and J-entralnment data 10. i1, - 65 - REDUCTION OF UF, TO UF,. This reaction is quantitative when UF6 is fed into an Fo flame contalning excess Hy, producing a powdery EFM product. The solid product cakesg and adheres to vessel walls, which should be avoided if possible in a remotely operated system. The liquid-phase reactlion proposed here ig more suitable for remote operation and should be developed. The operating conditions that need study are: temperature, U, con- centration, reaction rate, nozzle design, circulation rate, contactor design, and gas-liquid separation. The reducing conditions that exist in this operation are consistent with those required for purging nickel, chromium, and iron corrosion products from the fuel. Therefore the potentiality of using the reduction unit for simultanecously giving the fuel a final cleanup should be investigated. SOLID -LIQUID SEPARATION. The general area of high-temperature, solid-liguid separation in remote operations needs development. Filtration techniques should surely be investigated to determine operability and reliability in molten-fluoride systems. FISSICN PRODUCT BEHAVIOR. A bpetter understanding of fission product behavior throughout the processing plant is needed. In particular, date are needed on the ways in which the variocus nuclides partition in the several processing steps and on the efficiency of removal. A more basic study is concerned with the behavior of the fission products in the reactor environment to determine whether or not certain nuclides remain in the reactor system. URANTUM HEXAFLUORIDE PURIFICATION. The NaF and MgFo sorption units provide adequate decontamination for UFg. The batchwise units can be operated satisfactorily for both fuel and fertile streams of the MSBR; however, a coatinuous, temperature-zoned system would reduce the frequency of cell entry. Probably the largest uncertainties in UFg purification are in the removal of tellurium and ruthenium; nmeans of removing them should be developed. The portion of the fertile stream used as fuel makeup need not be passed through a sorption system because of the very small amount of fission-product contamination. However, the UF6 product that would eventually be handled by contact would need purification. PROTACTINIUM REMOVAL. The most significant advancement in fertile-stream processing can be made in the development of a process that removes protactinium. To be effective, the process must remove protactiniue five to ten times as fast as its decay rate; that 1s, the blanket volume would have to e processed every four to eight days. Simplicity and ease of operation are obvious requirements. Thus, a process based on forming protac- tinium oxide by ilon exchange appears promising and should be - 66 - Effective removal would obviate the need of sending any portion of the fertile stream to waste and possibly reduce makeup cost to that represented by thorium burnup alone. Such a process 18 not required for economic operation of the MSBR since fluorination and fractional discard can adequately control the fission product concentration; however, the potential advantage of reduced waste cost and improved breeding performance argues for basic development of a process. 12. FUEL-STREAM WASTE SYSTEM. The most economical method of waste mapnagement consists in bulk storage in a large heat exchanger tank. The heat generation problem is so severe that the plant waste must be diluted with an inexpensive, inert material, which complicates future processing for reeoverg of any contained values. The calculated loss of Ll and. U’in the fluorination- distillation process is small, being only 1.5 to 2% of the fuel- cycle cost. The possible use of fission-product decay heat should not be overlooked in the evalustion of an MSBR. The accumulated waste generates about 4.5 Mw (thermal) throughout most of the filling period of the waste tank. Thus salt-storage tempersture “can be maintained high enough to make the waste tank a source of high -temperature energy. 13. FERTILE -STREAM WASTE SYSTEM. The fertile-stream waste contains a significant inventory of valuable materials whose recovery is probably warranted. At the end of the 30-year filling period, the waste tank contains about 141,000 kg of Th, 10,400 kg of 114, and 116 kg of 233U; the “33U is isotopically pure material, having been formed almost entirely out of the Tission zone. Although not considered in this study, an in-cell decay period of about six months followed by refluorlnatlon appears to be advisable for greatly reducing the amount of “33Pa that enters the waste tank. 14. PROCESS CONTROL. This aspect of plant operation was given only a cursory review in this study, and no areas of unusual control difficulty were observed. A flow-control device for the molten- salt stream is needed, and the dynamic freeze-valve conceph should be developed. Analytical and sampling regquirements require a more thorough study than was given here. : ACKNOWLEDGEMENT W,IG. Stockdale assisted the authors in the development of the cost of the plant and equipment, and his help is gratefully acknowledged. 10. 11. 12. 13. 1h. 15. - 67 - REFERENCES . S. Bettis, ORNL, personal communication. H. F. Bauman and E. 5. Bettils, ORNL, personal communication on unpublished work still in progress. M. J. Kelly, ORNL, personal communication on unpublished work still in progress. E. D. Arnocld, ORNL, personal communication on unpublished work. R. E. Thome, Rare Barth Halides, ORNL-3804 (Mey 1965), p. L2. C. B. Jackson (ed.), Liquid Metals Handbook, Sodium (NaK) Supplement, 3d ed., U.S.A.E.C., Washington, D.C., 1955. K. A. Sense, M. J. Snyder, and J. W. Clegg, "The Vapor Pressure of Beryllium Fluoride,"” J. Phys. Chem. 58, 223-k (1954). O. Ruff, G. Schmidt, and S. Mugdan, "Arbeiten aus dem Gebeit Hoher Temperatur. XV. Die Dampfdrucke der Alkalifluoride,"” Z. Anorg. Allgem. Chem. 123, 83-8 (1922). H. von Wartenberg and H. Schulz, "Der Dampfdrucke Einiger Salze. II, Z. fur Elektrochemie 27, 568-73 E1921). K. A. Sense and R. W. Stone, '"Vapor Pressures and Molecular Compogitions of Vapors of the RUF-ZrF) and LiF@Zth Systems, " J. Phys. Chem. 62, 1k11-8 (1958). K. A. Sense and R. W. Stone, "Vapor Pressures and Molecular Compositions of Vapors of the Sodium Fluoride -Beryllium ¥Fluoride System," J. Phys. Chem. 62, 453-7 (1958). M. M. Popov, F. A. Kostylev, and N. V. Zubova, '"Vapor Pressures of Uranium Tetrafluoride," Zhur. Neorg. Khim. 4, 1708-9 (1959). W. R. Grimes, Reactor Chemisgtry Division Annual Progress Report for Period Ending January 31, 1960, ORNL-2931L (Apr: 29, 1960), p. 50-1. W. L. Carter and J. B. Ruch, A Cost Estimate for Bulk Disposal of Radioactive Waste Salt from Processing Zirconium-Uranium Fuel by the ORNL Fluoride Volatility Process, ORNL-TM-9L8 (Sept. 25, 196k) (Confidential). D. 0. Campbell and G. I. Cathers, ORNL, personal communication of unpublished data. 16. 17. 18. 19. 20, 22, s 2l 25, 26, | '(9E6'G?S to UF) in s Molten Salt, ORNL-TM-1051 (1965). o7. 28. 29. 30, - 31. 2. 33 ;.68_-. G. Burrows, Molecular Dlstlllatlon, p. 18~31 Oxford University Press, Great Britain, 1960 - W. H. Carr, S. Mann, and E. C. Moncrief, "Uranium -Zirconium Alloy Fuel Processing in the OBNL Volatility Pilot Plant," paper presented at the A.I.Ch.E. Symposium;_"Volatlllty Reproceselng of Nuclear Reactor Fuels, New York; N. Y. (Dec 1961) | e l R. P. Mllford S. Mann, J. B. Ruch, and W. H. Carr, Jr. ; ”Recovefing eUranlum Submarlne Reactor Fuels,” Ind. Eng Chem. 53, 357 (1961) J. C. Mallen, Volatlllzatlon of Uranlum as the Hexafluorlde from Drops of Molten Fluorlde Salt " - paper presented at the A C. S Natlonal Meeting in Chicago, I1l. (Sept 2, 196L4). G. I. Cathers, M. R. Bennett and R. L. Jolly, The Fused Salt-Fluoride Volatlllty Process for Reeoverlflg Uranium, ORNL—2661 (1959). ZM E. Whatley et al ,-Unlt Operations Sectlon Monthly Progress Report, --September 1963, ORNL—TM~785 (1964). "E. C. Moncrlef Results of Volatlllty Pillot Plant Cold Uranium :'Flowsheet Demonstratlon Runs TU-6 and TU-7, ORNL-CF -61-9 37 (1961). CW. WL Pitt, unpublished data, ORNL (1965). R. W. Kessie et al., Process Vessel Design for Frozen Wall Containment of Fused Salt, ANL-6377 (1961). M. J. Kelley, ORNL, personal communication (196L). L. E. Mcleese and C. D. Scott, Reconstitution of MSR Fuel by Reducing C. D. Scott, unpublished data (1965). W. 'L"Carter, R. P. Milford, and W. G. Stockdale, Design Studies and Cost Estlmates of Two Fluoride Volatility Plants, ORNL-TM-522 (1962). _ John Happel, Chemlcal Process Eeonomlcs, Wiley, New York, 1958. jR S. Arles and R. D. Newton, Chemlcal Englneerlng Cost Estlmatlon, TMcGrawafllll New York, 1955. eC H. Chilton (ed. ), Cost Englneerlng in the Process Industrles, '-Mcerawefilll New York,_l960 | | | jW G‘ Stockdale, ORNL, personal communleatlon (1965) .Sargent and Lundy Engineers, Capital Investment for lOOOAMwe Molten "Salt Converter Reference De51gn Power Reactor, SL. 199M (1962) 34, 35- 36. 37 38. 39- - 69 - Octave Levenspiel, "Chemical Reaction Engineering," Wiley, New York, 1962, W. R. Gembill, "Fused Salt Thermal Conductivity," Chem. Eng. 66 (16), 129-30 (1959). B. C. Blanke et al., Density and Viscosity of Fused Mixtures of Lithium, Beryllium, and Urénium Fluorides, Mound Laboratory Report MIM-1086 (Mar. 23, 1959). S. I. Cohen, W. D. Powers, and N. D. Greene, A Physical Property Summary for ANP Fluoride Mixtures, ORNL-2150 (Aug. 23, 1956). W. R. Grimes, Reactor Chemistry Annual Progress Report for Periocd Ending January 31, 1963, ORNL-3417, p. L7. ‘ " M. E. Lackey, Internally Cooled Molteg-Salt Reactors, ORNL-CF-59-6-89 (1959)- APPENDICES _73.. APPENDIX A. DESIGN CALCULATIONS FOR FUEL SALT FLUORINATOR AND COOLING TANK The unit operation of fluorinstion requires temperature contrcl at about 55000 for the salt being fluorinated. The fact that this salt is highly radicactive introduces a problem because 1t is necesgsary to extract this decay heat through the walls of a flucorinator whose size is fixed by other process requirements, such as throughput and residence time. If there is insufficient heat transfer surface available for thlis purpose, then the salt must be allowed to "cool" bvefore entering the fluorinator. The solution to this problem is to insert a cooling tank immediately upstream of the luorinator. The following calculations pertain to the thermal design and size of the fuel -salt fluorinator and the size of the cooling tank. It was determined that the maximum permissible heat fiux for the fluorinator is 5.31 x J_olL Btu hr ftf3 and that the size is L.75 in. in diameter by 10.3 £t high. The cooling tank reguires a volume of 22.5 ftg, A Turther result of this calculation is a graph of heatl generation rate at the cooling tank exit as a function of elapsed time since discharge from the reactor. - The fluorinator design criteria are: Fuel -salt flow rate 15 £ /aay Uranium rate T7.3 g-moles/day Fluorine utilization 33-1/3% Maximum mass flow rate of gas 0.277 slpm/in. Maximum heat flux through the frozen wall 1.5 kw/f{ Heat of reaction (UF) + Fp = UFg) -162 kcal/g-mole Residence time of salt 2 hr The Fg flow rate through the column is: F=77.-3x 3= 231.9 g-moles, = L.83 slpm . Fluorinator cross-sectional area is: .83 .2 A = 57T - i7.5 in. , or the column diameter is: D =/ .__?_‘-':%_Til;'é_: 4. 75 in. The total column volume needed is: 15 %2 3 : V= = 125 067 . Therefore, the column helght is: 1.25 x lhh 175 The area of the frozen wall per linear fodt of column is: - 10.3 Tt . h = The open volume per foot of column 1s | VO'_ 0. 120& ft3 .Therefore, the maxXimum heat removal rate 1n the frozenawall column is: . 1.25 x 1.5 x 3&13 | u a4 -3 "_;_But, the heat of reactlon contrlbutes at the follow1ng rate: 77.3 x 162 -1 ft—g.' BT¥ 1.5 X 0858 - 1620 Bfi“ ar _Thds, allowable fission product heating is: 5.15 x :LolP Btu hr“l ft“3 . The surge tank must have sufficient capacity to allow the fuel salt to L £e73. If the tank is assumed to be well -mixed, and assuming that the heat generation rate of the salt cool to a heat rate of 5.15 x lOLP Btu hr can be expressed as an exponential function of time, then,Bu H(t) = Keukt (heat generation rate), (A-1) CE(t) = % ewt/T (age distribution function), (A-2) where B T = V/F, average residence time, V= fluid volume in tank, F = volumetric Tlow rate, t = time since exit from reactor core, SIK,k = constants. | 'The heat generation rate of molten salt from the exit of the well-mixed surge tank can be expressed as: Hy = wfm%(t) H(t) dtlg | : (A-3) O _75_ The data on heat generation rate with respect to time elapsed since the salt has been removed from the reactor can be expressed as a series of equations of the form of Eq. (A-l). B8ix intervals were chosen from the data in Figs. 2a and b (see text). In each time interval, the constants of the exponential equation approximation were determined by coupling the equation at the two ends of the time interval. This resulted in a repre- sentation of the heat generation data, which was always equal to or some- what greater than the calculated heat generation rate. The values of the constants in the approximate equations are: Time Interval No. (hr) K k 1 0-0.0167 5.98 x 10° 24,2 2 0.0167-0.167 4.28 x 107 3.08 3 0.167 -1 .46 x 107 0.651 L 1-10 1.41 x 107 0.098 5 10 -100 5.85 x 10lL 0.0104 6 100 — 2.43 x 10LL 0.0015k4 After substitution of the constants, integration of Eq. (A-3) by time segments gives: .{ 5.98 x 10” [1 ) e-0.01667 (%-+ 2&.2)} (= + 2h.2) s HE: , 428 x 107 [§~0.01667 (% +3.98) __-0.1667 (3 + 3-98)] T (= + 3.98) 1 N ziu6 x 107 e-0.1667 (¥-+ 0.651) _ e—(%-+ 0.651)] 5 R 1 . 1iu1 x 10 _(z+0.098) _ _-l0(z+ 0.098)] (?-+ 0.098) = 5.85 x 10“ [e-lo(%-+ 0.010k4) e~100(%-+ 0'01ou)] (%-+ 0.010k) 2.43 x 10” [ -1oo(£-+ 0.0015A)]:} + il e T » (F + 0.00154) - 76 - This equation was used to determine heat generation rate of the molten salt at the cooling-tank exit in terms of average residence time (Fig. A-1). At the design heat generation time, the average residence time was 36 hr. ‘Therefore, the volume of the cooling tank must be: _ 36 x 15 _ 3 V = —= = 22.5 1. HE' HEAT GENERATING RATE AT COOLING TANK EXIT (Btu hr—1ff3) 6,0 x 10 ORNL DWG 65-7779 R1 @ 2 T /DESIGN RESIDENCE TIME | - e 10 20 30 40 50 60 70 80 90 100 T, AVERAGE RESIDENCE TIME (hr) Fig. A-1. Heat Generation Rate of Molten Salt at the Cooling Tank Exit. The salt is given 36 hours of decay time to facilitate temperature control in the fluorinator. - .LL? ;_78.; o APPENDIX B. FISSION-PRODUCT HEAT CENERATION RATES IN THE MOVABLE -BED SORBERS AND NaF WASTE TANKS Heat generation rates for the following components are presented here: MovableéBed Sorber | L Sodium Fluoride Waste Tanks ShortJIerm Coollng-Statlon - for Waste Sodium Fluoride Interlm-Storage of Waste FlSSlon products, which are volatlllzed in the fluorlnator, accumulate B in the movable4bed sorbers and create a heat source that must be considered in the de81gn of these units. Excess heat must be removed so that it does not 1nterfere with control of bed temperature Actual removal of the heat 1S:not_the problem_here.slnce_Sorbers,:whleh_accomplish this end, have been designed_and used in the ORNL Velatilifiy Pilot Plant. : p_. The essential problem is to estimate the heat generation rate due to radioactive decay of the fission pro&uets present in the system prejected_ here. This was done by first assuming that half the fissionwprOQUCtfiheet__ generating capacity which reached the fluorinator would exit to the sorber. Further, all the fission-product heat is concentrated in the lower zone in the sorber, and it was removed in the NaF waste stream. As noted before, one -fifth of the lower zone is exhausted per day, and there are two parallel sorbers alternately operating for 12 hr each. This results in an accumula - tion of decaying fission products or of decreasing heat sources in the sorber, in the NaF waste tanks, and finally in the interim waste-storage’ Tacility. i | The heat generation rate for various process components was approximated by determlnlng the average heat generatlon rate during a specific time perlod and by assumlng that this rate decayed as the total flSSlonmproduct ‘heat rate decayed as shown in Figs. 2a and b. The accumulated heat generatlon rate could then be expressed as a rate characterlstle of tfission -f*produets having an -average age_lntermedlate_between phe_oldest and the 'moet_reeently sorbed. - 79 - Movable -Bed Sorber In the movable-bed sorber, the heat load can be approximated by assuming that the volatile fission products from the fluorinator accumulate for 5 days and that those first accumulated will decay or cool as additional accumulation occurs. A table can be prepared for accumulated heat genera - tion rate, derived from the residence time of fission products in the sorber: Reegidence Time of Fission Products in Sorber Heat Generation Rate ~ (days) (Btu/hr) 1 144,000 2 120, 000 3 106, 000 L 96, 000 5 86, 000 Steady-state heat generation rate in movable -bed sorber 552,000 Btu/hr Sodium Fluoride Wagte Containers The Nal' waste tanks accumulate NaF and fission products from the sorbers. FEach tank holds two complete bottom sorber zones from each of two sorbers {10 days' accumulation of fission producte in one-day incre - ments ). These zones exhaust to the NaF waste tank each day for 10 days. According to the slope of the fission-product decay heat curve (Figs. 2a and b), the average residence time of the fission products, and the average heat generation rate of the Na¥ bed material as it leaves the sorber, the following heat generation rate exists in the NaF waste tank at the end of 10 days, at which time it is full. Residence Time of NaF 3 . Average Heat in Container Generation Rate (days) (Btu/hr) 1 110, 300 2 183,400 3 1,500 - 4 62, 600 5 56,600 6 .53,600 I 50,700 8 k7,700 9 L1, 700 10 38, 700 - Heat generation rate in NaF . waste tank when filled 616, 800 Btu/hr Short-Term Cocoling Station for Waste Sodium Fluoride The NaF waste containers are to be cooled for 80 days, within the processing area. This calls for a cooling station with the capablility of cooling eight NaF waste containers whose average age varies from about 7 days to 87 days. The following heat generation rates apply to the cooling station: Average Heat Tdentification Number Generation Rate of Containers (Btu/hr) 1 (about 7 days old) 616,800 2 | 380, 900 3 _ 261, 200 L | | | o - 210, 400 5 | | 163,600 5 145,100 I 119,700 8 (about 87 days old) 104, 000 Total heat generation at : ‘cooling station | 2,001, 700 Btu/hr - 81 - Interim-Storage of Waste As mentioned above, the waste containers are to be cooled for a minimum of 80 days by forced air convection prior to their transfer to the interim waste-storage facility. The average heat generation rate of one container at the end of this time (based on the slope of the fission- product decay heat for this period) will be 104,000 Btu/hr. An average of 28 containers are to be sent to this waste-storage facility per year at one-month intervals — an average of 2.33 containers per month. When the cell 1s opened for transfer, the most recently filled container has 80-day-old material in it, whereas, the first filled is about 110 days cld. On the average, the heat generating rate of the transferred containers is characteristic of one that is 95 days old whose rate is 87,500 Btu/hrg The average heat generation rate of containers sent to the interim storage facility is: 2.33 x 87,500 = 203,900 Btu/hr . The heat generation rate of these containers decays with storage time, and the average rate in the interim facility for one year's acceunulation is: Residence Time of NaF Waste Average Heat Container in Interim Facility Generation Rate (months ) (Btu/hr) 1 203, 900 2 150,200 3 102, 000 Iy 83, 700 5 69, 800 6 62,200 Il 23,'(00 8 42,900 9 37,600 10 33, 300 11 29,000 12 26, 800 Average heat generation rate of a one -year accumulation of containers 895,100 Btu/ar lee - For five years of wasfie.accumulation, the totai“héat géneration rate_:: in the interim facility is: e - B '. | L Average Héat_ Age of NaF Waste Tanks ~Generation Rate (years) - (Btu/hr) -?895;100. © 381, 300 198, 900 | 132,600 5 116,000 o N Heat generation rate from NaF _ o S e waste in interim facility . 1,723,900 Btu/nr - 83 - APPENDIX C. ESTIMATION OF DISTILIATION RATE IN VACUUM STILL 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 by Burrows.l6 For our étill, it is estimated that at lOOOOC and at a pressure of 1 mm of Hg, the distillation rates for LiF and BeF 3 and 2.02 x 107 2, For molecular distillation, a still is designed so that its condensing o 2T, respectively, 3.32 x 10~ g Sec"l cm surface 1s located quite close to its evaporating surface, thereby minimizing the transport distance for the vapor. I the separation distance is small enough and 1f the pressure i1s low enough, a molecule leaving the liguid surface has a very high probability of reaching the condenser without ccolliding with another molecule. This is the essence of molecular distillation. Our still cannot be described precisely as a molecular still because 1ts operating pressure is too high; however, the pressure is low enough that conditions for molecular distillation are approached. TFor this reason, the calculational methods 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; -2 -1 -2 5.83 x 10 P ,M?T g sec cm (Cc-1) where pmm is the equilibrium vapor pressure in mm of Hg, M is the molecular welght in grams, and T is the absolute temperature in °K. The resulting distillation rate is expressed as g sec™ cm ™. The actual rate of distillation at low pressure will be less than the theoretical rate because there will be collisions in the vapor space. Burrows develcoped an expression for the factor by which the theoretical distillation rate should be multiplied to get the actual rate. Summarizing, -"8_4'-' his treatment considers three events that can occur to the moleeule in the vapor space: Wtdy : 1. Some molecules can reaeh the condenser without a collision; this fraction is e4K, where K 1s a dimensionless .factor depending upon the distance between evaporator and condenser, the equilibrium . mean free path of the molecules, and the shape of the evaporating suriace. _ ' . T 2. The fraction of molecules that collide is (1 - e ™ }, and the % XK o fraction of these molecules that reach the condenser is (l -e e approx1mately ' ' 3. From purely geometrical considerations, the probablllty of a molecule's striking the condenser after many collisions, which result in random motion, is a factor ¥, the ratio of condenser area to condenser area plus evaporator area. The Tfraction of these molecules that reach the condenser after meny colllslons is F(1 - eX)(1 - eX). - The total fraction L of wvaporized molecules that reach the condenser - is given by the sum of the three fractions above: f=e™ 4 (1 - e—K)e—K + F(1 - e—K)(l - e_K) , =7 o+ (1 - 1) (e Ty | =1 -(1L -F)(1 - e‘K)2 - - (c-2) When the theoretical rate is multiplied by f, defined to be the evaporation coefficient, the rate of distillation for a single component becomes : - - 2 5.83 x 1070 20, [T 7T g osec s oem™ . (¢-3) The evaporation coefficient I is not readily calculated because of the difficulty in determining the proper value of K, which is defined to be: K = k—%E—— (C-4) where gap distance between evaporator and condenser,. mean free path of molecules-in-eqflilibrifim vapef, suitable consfiant, which is 1 or 1arger,_used tb relate actual .conditions to average eonditions in the gap. _Burrows16 reporfie ezperimental values of k in the range 3.5_£o 30. gL However, as K increases, the value of f approaches the value of F, a number that is known from the configuration of the still. For the still shown in Fig. 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 that KE _ 2.3 x-1022 T , em (¢-5) B © where o = molecular diameter, cm, T = temperature, OK, P = equilibrium pressure, mm Hg. Substituting into Eq. (C-4), d pmm 02 K = : (c-6) For LiF at T 1273%K, c X 3.26 x 10'8 cm, I pm = l.O mi Hg) and for this still, d ¥ 10 cm. Using these values, one finds that K= 12.1 for k = 30 , and K= 103.7 for k = 3.5 . In either case, K is sufficiently large that Eq. (C-2) reduces to f~FZ0.58. i In processing the fuel stream of the MSER, distillation must treat a multicomponent mixture of LiFéBeFE-UFu—fission products. However, LiF and BeF2 constitute more than 99 mole % of the mixture, as shown in Table C-1, making it possible to treat the mixture as a binsry solution for purposes of calculating the distillation rate. Burrows has shown that for a binary mixture, Eq. (C-3), which gives the distillation rate, becomes: 5.83 x 107 fPleAqffii7T g/sec , (c-7) 86 - Whefe _ _ X = mole fraction of component 1 in the liquid mixture, ST | : - 2 A = area of evaporating surface, cm_, _ P, = equilibrium vapor pressure of component 1 at temperature T, mm Hg, . M, = gram-molecular weight of-componént 1. A more rigorous representation of Eq. (¢-7) would include the act1v1ty | -coeff1c1ent 71 of component 1 1nstead of' the evaporatlon coefflclent T to account for_deVLat;ons from_ldeallty. _However,_the activity chfflclentS afé fiot knofin.fbr these.éalt SOlutions, 80 fdr.this Stfidy it will Dbe. aséfimed .that Eq (C-?) gives a reasonably'valld estimate of the distillation rate. '~ Since the still operates at constant volume, a material balsnce fequlres that _BeF2 and LiF:d;Stlll ‘at the same rate at which they enter the still. The vapor composition is therefore about 69-31 mole % LiF -BeF,, _the same ag it 1s in, the stream enterlng the stlll If the total pressure :1n the vapor space of the stlll is kept at 1 mm Hg, the partial pressures of LlF and BdF are O 69 and O 31 mm Hg, respectively, assuming that there 2 -Table.C—l.- Reactor-Data and Approximate Composition of Fuel Stream at Equilibrium Reactor Data Fuel volume - 671 T3 Cycle time = 58 days Power = 2160 Mw (thermal) Approximate Composition (mole fraction) '233U 0.0029 -235U - | .Négligible Other U __' - | 0.0002 S e S SR O BeF, - | 0.318 Fission products | ~ 0.0008% 8. ' . ' ~Gaseous fission products purged in the reactor circulating locp, and noble fisslon products that are removed on a very short cycle by attaching them- 'selves_tq_the:INOReS walls, do not contribute to this value. 87 are no other volatile compounds in the still. This agsumption is wvalid because the quantities of volatile fission products are quite small. If we assume that Raoult's law applies, the partial pressure of BeF2 is: p = P BeF at 1000°C is about T4 mm Heg; therefore the mole BeF2 XBeFE The vapor pressure of BeF2 fractions in the still liquid are approximately: xBng = 0.31/Thk = 0.00k42 , and Lar T 7 *per,, T 0.9958 . By Eq. (C-7), the rates at which LiF and BeF,, evaporate from a surface having area A can be determined; Wi = 5.83 x 1072 A [ xP [M7T:]L1F g/sec , (c-8) and 2 Voen = 5.83 x 107° fa[xp JW/T ]LiF g/sec . (c-9) 2 The vapor pressure of LiF at lOOOOC is 0.61 mm Hg. When this wvalue and the values of the other quantities are inserted in Egs. (C-8) and (C-9), the specific evaporation rates are found to be: 3.32 x 10”3 g sec ™ cm_2 for LiF, and 0.02 x 107 g sec™ an™ for BeF, . Adding these last two equations and solving for the area of the evaporating surface: "LaF " VBer, A = s) ’ (c-10) 5.83x107 £[(xP [M/T) . + (@ /M7T)BQF2] (436,600 + 359,000) g A = (254 )(3600) > sec N 2 (5.83xlo'2)(o.58)[(0.9958)(0.69)/2671273~P(o.ooue)(7u{[E77j§7§g)Secgcme = 1720 cm2 5 ~ 1.85 £6° . 88 - Now, the area of the pOOl surface lh-the efiill (see. Fig.-9 inetexfi)'is _h 9 ft However, 1t is not safe-to say that the stlll is overde51gned _by a“factor of about 2. 5. Better data on, vapor pressures and partlcularly .3-act1V1ty coefflclents mlght make a con51derable dlfference 1n the calculated ' _dlstlllatlon rate. It is Strongly suggested that the stlll con81dered for e'thls study be Vlewed only as an approx1mate de51gn that w1ll probably change _as mere . is learned about the dlstlllatlon process. : - 89 - APPENDIX D. FISSION-PRODUCT ACCUMULATION AND HEAT GENERATION RATE IN LITHIUM FLUORIDE POOL IN VACUUM STILL Nonvolatile fission products accumulate in the vacuum still as the LiF--BeF2 fuel carrier and velatile fission products are distilled away. The still is initially charged with 4 Tt° of makeup 'LiF, evacuated to & pressure of 1 mm Hg or less and adjusted to a temperature of about lOOOOC. Fluorinated fuel salt, contalning fission products, is allowed to flow continuously into the pool of LiF, and the still is operated so that the rate of distillation is exactly equal to the feed rate. There is no bottom discharge, and the volume remains constant. Kelly's work25 showed that the initial Li¥F and BeF2 1000 from rare earths. Operation 1s continued at the above pressure and distillate is decontaminated by a factor of 100 to temperature until one of two phenomena forces termination: (1) Either the 3 solubility of fission products in the 4 ft” of LiF is exceeded and trouble- some 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 LiF~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 out-of -reactor holdup, and since the solubility of the rare earths in LiF at 1000°C is about 50 mole %, it is apparent that condition (2) above will be controlling. Decay-heat removal will be a serious problem long before solubility limits are approached. It has been determined that the h»ft3 still can operate continuously for about 6T.h days at a distilla- tion rate of 15 fts/day, processing fuel that has the heat generation characteristics shown in Figs. 2a and b of the texi. The significance of 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 106 Btu/nr, the maximum that can be removed by the cooling system. The calculations below are believed to give a conservative estimate of the still performance. The basic data on fission-product heat generation - 90 - versus decay time represent gross values and hence do not exclude the contributions of those fission products removed by the reacltor gas sparge, by deposition on metal surfaces in the reactor system, or by the fluorina- tion step in chemical processing. Furthermore, the 15-ft3/day design rate is probably excessive for a 1000 -Mw (electrical) MSER; the economic optimum rate is perhaps nearer 12 ft3/day. Also, no credit was taken for periodic interruptions in processing due to the reactor's operating at less than lOO% planfi'faétor. All these factors tended to shorten the still operating cyéle and increase the discard cost. | The calculations are arranged as follows: TFirst, an analytical expression is derived for the heat generation rate of fission products in the still as a function of elapsed time since discharge from the reactor. Second, this equation 1s then used to evaluate the still design. Analytical Expression for Heatl Generation Rate When irradiated fuel salt is discharged from the fluorinator, it has been out of the reactor about 38 hr. A cooling tank downstream from the fluorinator adds another 24 hr of holdup, so that the salt is about 62 hr old when it reaches the still. The heat generation characteristics of the salt durling this periocd were calculated by assuming an infinitely mixed system; these calculations are described in Appendix A, and the results are shown graphically in Fig. A-1. Continuous operation is assumed. throughout the system, and, by the time the fuel reaches the still, its heat generation (Fig. A-1) is aboub 4.45 x 10° Btu hr~t ££73. The still is a sink for fission products that exhibit decay behavior like that shown in Fig. 2a and b. Initially at t = 62 hr = 2.58 days, the 4 £t of LiF in the still is fresh material having a zero heat genera- tion rate, but this condition changes rapidly when the still is put on- stream. Flow into the still is continuous, and the rate of healt generation Wil1 rise until flow is stopped. The magnitude of thié rate at any future time T is therefore an integrated quantity over the accumulation period: to'z 2.58 days to T. To describe the behavior of the still during this ~period, define the quantities o - o} = gpecific heat generation rate of fuel entering still, - b5 x 107 Bt hrt £, q(t) = specific heat generstion rate at time t, Btu hr — £t75, ¥ = flow rate of fuel into still, - 15 £t3/day, v(t) = fuel volume processed at time t, T = time that still has been operating, days, <5 i volume fuel pool in still, = L £g3, The heat generation in the still pool at any time t is V, qs(t), expressed as Bubt/hr, which is the difference between what the rate would have been if there had been no decay and the amount the rate has decreased because of decay. That is, v, ay(6) = [ o) av(s) - [ a(s) avls) - (D-1) v v Noting that dv = F at , (D-2) there results v, oa(t) = qufdt -qu_(t) dt . (D-3) t T The limits of integration extend from to = 2.58 days to t = T days, where T denotes the time after discharge from the reactor at which the heat generation rate is desired. To treat BEg. (D-3) analytically, the function q(t) is obtained from Figs. 2a and b by representing the curve on this graph by four straight- line segments in the range 2 to 400 days. The procedure is diagrammed in Fig. D-1. The general form of the equations is: a(t) = kt" . The slope n of each segment is determined from values of q(t) read directly from Figs. 2a and b. The initial conditions of the fuel entering the still, namely, t = 2.58 days, q = Lk.U5 x :{_olL Btu hr the equation for the first segment to determine the constant k.. The 1 constant k2 was determined similarly by using the end condition of the ft"3, were introduced into HEAT GENERATION RATE (Btu hr"]fi"3) ORNL DWG 65-3023R] i | | l - | | | | ! | | % L L 10 20 100 400 TIME AFTER DISCHARGE FROM REACTOR - (days) _ Fig. D-1. Schqmatic Logarithmic Curve Showing Approximations to Fission~-Product Decay. In the mathematical model for computing the heat generation rate, 1t is convenlent to divide the time scale into four parts. -6 - _93_ first segment at t = 10 days. This stepwise procedure was followed for segments two and three to evaluate k., and Kh' The four equations so 3 determined are: -0.43 q; = 6.69 x 10” ty 2 <%, =10 days , (D-L) a, = 9.686 x 10" £, 70" 27 10 < %, < 20 days , (D-5) ay = 2L.2 x 10" t3“0'85” 20 < ¢, < 100 days (D-6) q, = 194.8 x 10" ), 7" 992 100 < ) = 400 days . (D-7) In these equations, g 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 of Egs. (D-4 to (D-7), and the integration is carried out to give: v_q () = 15 x 10% [4.h5 (T - 2.58) - 11.7k (th.57 - 2.589-27) ) 23.62(t20‘”l _ loo.ul) _ 1&5.2(t30'1”6 ) 200.1&6) + 581.5(tu-0°335 - 100'0'335)] . (D-8) The units of T, © t.,, and t, are days; the units of V q are by 58 1’ o 3 Btu's/hr. The restrictions on the several t's are those specified for Egs. (D-4) to (D-7). 1In solving Eq. (D-8) for v, qS(T), a value is chosen t for T, and this value is assigned to either t or th in the 1 B by fellowing way: For * 2.58 < T < 10 days, let b= T 10 < T < 20 days, let t2 = T ; 20 < T = 100 days, let t3 = T ; 100 < T < 400 days, let th =T . All ti < T assume their maximum values; all ti > T, of course, are not A considered. Filgure D-2 is a plot of Eq. (D-8); it gives the integrated heat generation rate for times between 2.58 and LO0O days after discharge from the reactor. Note that the accumulation time for fission products in the still is (T - 2.58) days. - oL - ORNL DWG 65-1821 R2 108 [ e o o o~ 0w : ‘\ S f— o} ~! - - el o [2.0) b o L I HEAT GENERATION RATE (Btu/hr) | { (=] o o~ D X 10-1" 511 | ! - 1 2 ; 4‘ 5'878910 2 3 45678L|02 2 3 456785103 0 at 2.55 d TIME AFTER DISCHARGE FROM REACTOR (days) 10 Fig. D-2. Heat Generation Rate in the LiF Pool Resulting from Pission-Product Accumulation in the Still. The still is charged with L £t3 of fresh, molten LiF at the beginning of the distillation cycle. Barren fuel-carrier (2.58 days old) flows into this LiF at a rate of 15 £t3/day, and LiF -BeF, distills at the same rate, keeping the volume constant. Accumulatlng fisslon products cause a rapid increase in the - heat generation rate. - 05 - Evaluation of Vacuum-Still Design Two principal conditions had to be satisfied in the still design: (1) the evaporating surface had to be sufficient for obtaining the required 15—ft3/day distillation rate, and (2) maximum heat transfer surface had to be provided to minimize the frequency with which the still is drained to the waste tank. The choice of dimensions was somewhat arbitrary; a tube- sheet diameter of 2.5 ft was chosen because 1t gave an evaporating surface that was about 2.5 times the calculated areca, and a closely spaced arrange=- ment of 2.5-ft-long by l/2-in.udiam tubes was used for high cooling-surface- to-volume ratio. The primary unknown operating condition is the length of time that fission products can be accumulated before the integrated heat generation taxes the capacity of the cooling system; an estimate of this time is determined in the following calculations. The schematic diagram in Fig. D-3 depicts the still operation and summarizes calculated performance and physical data. The calculations which are described %below indicate that the still can accumulate fission products for about 67.4 days in a h—ft3 LiF pool when the gtill 1s operated continuously at & feed rate of 15 ft3/day. The discard rate for LiF is therefore a very small fraction of the processing rate, being about O.h%. As mentioned above, the heat generation rate predicted by Eq. (D-8) is probably excessive because gross instead of net fission-product data were used; therefore this accumulation time can be treasted as & lower limit. Physical Data for Still Applicable data are given on Fig. D-3. Heat Transfer Characteristics The 1000°C temperature of the IiF pool in the still was chosen to achieve adequate distillation rate and fission-product solubility. The pool transfers heat to the tube walls by natural convection, and this heat is picked up on the outside of the tubes by NaK (22.3-77.7 mole % Na-K eutectic) under forced convection. The NaK coolant enters the still at . STILL BISTILLATE 15 ff3/dc|y 0 . VACUUM SYSTEM FLUORINATOR ~550°C NaK 750°C LiF-BeFz-FP STILL FEED ..96.,. CONDENSER COOLANT (AIR OR NaK) b e e | LiF-BeF3 DISTILLATE 15§32 FEED TO UFg—eUFy4 REDUCTION COLUMN Fig. D-3. (5§13 /day NaK 750°C WASTE RECEIVER 4§13 67 DAY HOLD-UP (900-1000°C)___ ORNL DWG 65-3040 LiF FROM OUTSIDE MAKE-UP LiF MAKE-UP 413 AIR OR Nak NaK (550° C) COOLANT i153 gpm (max) NaK (550°C) COOLANT 1153 ngrn {max) DRAIN TO UNDERGROUND STORAGE LiIF&FP STILL AND WASTE RECEIVER DATA 817-Ypin. x 16ga. INOR-8 TUBES TUBE SURFACE =267 §12 TUBE SHEET =2.5ft Dio.x2.5¢f High . STILL CONDENSER =69 fi2 HEAT LOAD {max):31x108 piyshr EVAPORATING SURFACE = 4.9 §12 TUBE SPACING = %5 in. A CENTERS Schematic Diasgram of Vacuum Still Operation. - 97 - SSOOC and exits at 75000, giving a logarithmic mean temperature difference of 340°C (612°F). Heat transfer characteristics of LiF were calculated by using the Nusselt eguation for natural convection, in conjunction with the physical-property data given in Appendix F. A Nusselt-type equation for flow normal to banks of tubes was used for calculating the heat transfer properties of the coolant. Since the total heat dissipation requirement of the still was unknown (because fission-product accumulation time was not known), the coolant heat transfer coefficient was expressed as a function of the heat generation rate. It was determined that the still could dissipate about 31 x 106 Btu/hr for the NaK flow conditions shown in Fig. D=3. The overall heat transfer coefficient.for maximum heat flux is about 190 Btu hr ™~ ££™= °%F L. From Fig. D-2, the time after discharge from the reactor corresponding to the above integrated heal generation is 70 days. The fission-product accumulation time is 2.58 days less or 67.4 days because of the time lag before fuel reaches the distillation step. Li¥F, BeF,, UF,, and Fission-Product Discard Rates. — At the end of the 67.4-day cycle, the still contents are primarily LiF and fission products. The equilibrium BeF2 concentration was estimated above to be only 0.4 mole % because of its relatively high vapor pressure atl 1000°C. If it is assumed that the fluorination step is 99.7% efficient for uranium removal, then in 67.4 days about 3.64 kg of U will have entered the still. Uranium tetrafluoride has a vapor pressure at 1000°C about 2.3 times that of LiF, so a portion of the UF)Jr will be recovered in the distillate. The amount recovered cannot be calculated until more is known about the vapor- liguid edquilibria of multicomponent melten salt mixtures. Filssion-product accumulation during this period is aspproximstely 190.1 kg; most of the RbF and CsF of this inventory will distill becsuse of relatively high vapor pressures. * An estimate of the inventory relegated to waste every 67.4 days is given in Teble D-1. A reactor plant-factor of 100% was assumed in the calculation. These values are based on ideal-solution behavior, particularly with regard to LiF and BeFE; this is almost surely not the case for this mixture. As more is learned about the activities of the components, it can be expécted that the compositions of Table D-1 will be different from those shown. =98 - Table D-l. Estimated Comp081tlon of the Vacuum Still After 67.k Days' Operation &t 15 £t 3 /Day Still volume = H £ Weight Mole (kg) Fraction @3By 33 . <0.002 235U RO o <05 S ey | T g wr 19509 0.850 BeF, 2.1 0.005 Fission products®™ - 190.1 0.143 _aMolecular weight.= l50,~a$sumed. Heat Removed by Stlll Condenser. m'The load on the still condenser s relatlvely small, con81st1ng only of the latent heat in 15 ftB/day of "a 69-31 mole % LanBéFg mlxture plus some radiation from the pool surface. The latent heat of vaporlzatlon of LiF was estimated from vapor pressure ‘data to be MM,OOO cal/g-mole; an experimental valuell of 50,100 cal/g-mole was used for_BeFe. Radiative heat transfer amounted to about 62, 700 Btu/hr, giving a total condenser duty of 198,300 Btu/hr. ' _For smooth, nonflashing disfiillation, the condensate is the LiF-Bng eutectic having a melting point_about 5OOOC, and. a condenser temperature | slightly_higher than this is éati#factory. - IT there are deviations from .ideal-opération so that a higher@melting composition distills, the con- -_défiéer tempera£ure could be adjusted TO temperatures slightly above the end 845°C for LiF. | melting poi@ts of the pure components, 803°C for BeF, 2 99 - APPENDiX E. DESIGN CALCULATIONS FOR WASTE-STORAGE SYSTEM Separate storage is provided for fuel and fertile stream wastes as fTluorides in underground tanks designed for bulk accumulation over a 30- year period. Waste management in the post-~-30-year period was not con- sidered in this study. However, as mentioned in the text (see Table 2), it will probably be desirable to reprocess the fertile-stream waste, which is only mildly radiocactive, at some future time for recovery of thorium, 233 lithium, and uranium values. About 116 kg of U will be present. Fission products, separated in this recovery, could be stored longer, depending on the activity. - In the fuel-stream waste, the only significant value, other than the possible future value of individual fission products, is 7Li. However, as explained below, it is necessary to add a mixture of NaF -KF to this waste to facillitate heat transfer. Since these compounds are chemically similar to LiF, recovery of the lithium is difficult. In any event, at the end of the 30-year period, the desirability of recovery would have to be analyzed in light of the prevailling costs. Two basic problems must be solved in designing the waste storage system: (1) the integrated rate of heat generation by fission ~-product decay must be determined, and (2), using the results of (1), the most economlc deslgn for the prevailing conditions must be found. The heat generation rate is computed Trom the fission-product decay behavior exhibited in Figs. 2a and b (see text), and the results are shown in Figs. E-1L and E~-2. A previous study by Carter and Ruc:hll!L examined a similar waste-storage problem, and, in accord with their recommendation for economic waste management, bulk storage in large, heat-exchanger tanks is adopted for our wastes. Conceptual designs of the waste facilities are shown on Dwgs. 58080 D and 58081 ¢ in Appendix F. Over the 30-year period, 784 £ This volume includes 520 ft3 of LiF—fission-product mixture drained from the processing cell, plus 284 ft3 of NaF -KF' diluent. The storage tank is of fuel -stream waste are collected. 16 f+ in diameter and 6.33 4 high. The corresponding volume of fertile- stream waste is 1783 ft3; this is stored in a tank 13.5 ft in diameter by 13.5 £% high. ‘ . -100 - Most of the following calculations are concernéd with the design of t"tne waste tank for fuelustream effluent because tne most dlffloult deszgn: - problems ooour for thls waste After the . deoay eharaoterlstlcs of the j flSSlon produots and 1ntegrated heat generatlon rate are determlned the -lcaloulatlons examlne, 1n the follow1ng order, the ba51c features of tank _,]fl ”de51gn, ‘maximum allowable heat generatlon rate, volume of. dlluent waste- 'tank deelgn, and the. underground storage fac111ty : Once these calculatlons- are. made, oomputatlons for the fertlle-stream waste are almost 1no1dental _.:Fuel¢StreandWaete Systen 'Decay'Characteristicsdof'the Fission.Products The 1n1t1al problem 1n deszgnlng the waste tank is to determlne the 'z'deoay in terms of the tlme-related oharacterlstlcs of the salt belng added "e'ito the tank That lS to say} tne stlll bottoms, which represent an accumula~ | '--tlon of flSSlon products hav1ng ages in the range 2.58 to 70 days, generate d,neat at a rate onaracterlstlc of flSSlon produots hav1ng an average 'age =somewnere between these values Tne tlme-related benav1or of tne 1ntegrated .f.speclflc heat generatlon rate is shown in Flgf L1l of the text for the case of'no inert salt dilution. The graph covers a 5-year collection period of h-ft3 batches every 67.4 days " The initial point on the graph begins at -.about lhO days, this 1is the accumulated time in the processing plant with a -reference time of zero taken as the day the fuel is dlsoharged from the _nreactor An 1n-cell delay perlod of 67.54 days after removal from the still 1s 1ncluded in this time to permlt some initial coollng before dralnlng to E 'the underground tank The decay curves of Flgs Eadand b oan_be.used to'determine tne.decay :__behav1or when.thls average age'has been fOund.d'Tne volume of fuel salt 'ffrom Whlch flSSlon produots have accumulated 1s 15 ft3/day X 67 u days f 1011 ft3 '~”uIt was determlned in Appendlx D that the heat generatlon rate at age equal o 6 - -:to 10 daye out of the reactor is 31 x-10 Btu/hr Therefore on the ba51s d_of unlt volume of core salt the average heat generatlon rate 15 o 6 7f9$@;1£f%%L.w31fi6x;&%%uhr f@3,-~ - 101 - Define the average heat generation, /\q dv Qg = (2-1) where v is the volume. The processing rate F, in ftB/hr, is steady, so in the time interval dt, dv = F dt . | (B-2) It was shown in Appendix D that the instantanecus heat generation rate is represented by an equation of the form: -1 1 , Btuhr™ £t (E-3) Combining Egs. (E-2) and (E-3) in (E-1): tavg k /\ £+ gt O qavg B t ? avg [% O . [tu-n) o) | B avg 0 Yavg T (1 - n)(t, -t,) q =kt : (B-4) g where k and n are characteristic constants for the decay curve; to (= 2.58 days ) ig the age of the fission products at the beginning of distillation. The desired average age is tavg’ the time corresponding to qavg' It will be recalled from the discussion of Appendix D (in particular Fig. D-1) that the q-versus-t curve is best represented by four segments over the range of interest. The four equations are: 4 t~o.u3 q; = 6.69 x 10 1 2 < tl < 10 days, (E-5) a, = 9.686 x 1ou t50'59 10 < t, < 20 days, (E-6) a3 = 21.2 x ].Ol‘L t§O.85h 20 < t3 < 100 days, (BE-7) q, = 194.8 x 10* t£1'335 100 < t, < 40O days. (E-8) - 102 - If t ' lles 1n the rauge 2 to lO days, then 1ts value ean be found dlrectly .L_from Eq (E &) by utilizing k = 6 69 x th and n O 43 from Eq. (E~5) | . However,_lf % fg'ls in the range lO to 20 days, Eq. (E-M) coitalns a second - term in the numerator 1nvolv1ng the constants k 9. 686 x 10" and n = 0. 59 | of Eq. (E~6) | Slmllarly for t ve in the range 20 to 100 daye, the. numerator . of Eq. (E-h) contalns a thlrd term 1nvolv1ng the constants of Eq; (E T) "t'The solutlon is ea51ly determlned by trlal, and for thls case the proper :;Vform of Eq_ (E—h) contalns the constants k and n from Eqs (E—B) end (E-6) T' ne o L ___ "'k". : . . o "1, 1-n 4n 2 1-n " ~n SR 1 - l 2 2 o avg e 6 69x10“ | 0. 57 | o 57 9 686xlolL .. 041 o;ul - *w-————-(10 2 58 ) (2 71070 3 066 lou 0. 57 .41 ave avg | .The solutlon To: thls equatlon is: _tévg = ll 6 days An average age near “the 1ower end of the time ecale would be expected because of the 1arger ..t'contrlbutlou by‘the younger flSSlon products to heat generatlon The quantity tlo':_lo days used in Eq. (E-9) denotes the upper limit of 'Eq. (E-5) and the lower limit of Eq. (E-6). .:_Eutegrated Heat Generation in Waste Tank 3 ~When the vacuum stlll is taken of T stream and dralned its L-ft volume is generatlng heat at 31 X 106 Btu/hr, but this rate is decreasing 1n 8 manner characterlstlc of ll 6~day~old fission products, as shown on - Fig. Eb At this point, the change in rate is rather rapid, 1nd1cat1ng "that a, short delay tlme 1n the cell before dralnlng to underground storage w1ll apprec1ably alleviate de51gn requlrements for the 1arge tank. 3 'Installatlon of a second T vessel qulte 51mllar +0 the vacuum stlll :-fllS 8, eouvenleut way of pTOVldlflg a 67 Mnday holdup, durlng this tlme the 't;_7heat generatlon rate decreaees by'a factor of about h 5. A decay curve was. calculated for a typlcal batch of stlll re31due by 'toeiaesumlng that the curve was parallel to the decay curve of Flg. 2b for MSER flSSlon producte after ll 6 days. The 67 h-day in cell delay makes “the & accumulated tlme from reactor dlecharge to underground storage equal to - - 103 - 137.4 days; at this time the heat generation rate of the h-rt3 batch is 6.8 x 106 Btu/hr. The curves of Fig. E-L are plotted to show the thermal history of the underground waste tank over a 5-year collection period during which 27 batches are added; alsc the decay in the post-b-year collection period is shown. The curve with positive slope represents the buildup of heat generation rate from all batches accumulated up to the indicated time. This curve is actually a stepped curve, but for convenience it has been drawn smooth through the maximum point of each step. The waste tank is designed for 30 years' collection, and the accumulated heat genera - tion at this time can be obtained with little error by extrapolating the positive-sloped curve of Fig. E-l. Figure E-2 has been included to show the integrated heat generation rate in the underground waste tank when the still residue is drained into the tank immediately upon completion of the distillation cycle. Figures E-1 and E-2 represent an upper limit for fission-product heat generation. As pointed out earlier in this report, the values are for gross fission-product decay, which includes nuclides that have been removed prior to wvacuum distillation. Also the curves were calculated for a reactor plant factor of 100%. Basic Features of Tank Design A previous studylLL showed that bulk storage in large tanks is the most economical management for fluoride wastes. A 30-year periocd was chosen because it coincided with the amortization pericd of the reactor plant, and g single tank is sufficient because the overall waste volume is small. For economy and reliability, cooling by forced air draft was adopted, and an upper limit of 60 fi/sec was assigned to the velocity. Waste temperature in the tank is not to exceed 750°C. Meximum Allowable Heat Generation Rate Under the above ground rules it is apparent that there is an upper limit that can be tolerated for the volumetric heat production rate, and that the waste will have to be diluted to wet sufficient cooling surface. In 30 years at 80% reactor plant factor there will be 520 ft3 of radiocactive o - : _ ORNL DWG 65-1817 Rl A S E I B B N l T T T I — ' _ e ACCUMULATED SRR ' ' ' : e HEAT GENERATION : .27 BATCHES ACCUMULATED ~OVER 5-YEAR PERIOD AT . 67.4-DAY INTERVALS =~ 4 BY/BATCH HEAT GENERATION RATE (Bt/hr) . _ 1 YEAR 2 5.0 a0 4 ! ) Il Pl !]| | | 'i ! ll } |. 1|[ | 10 | L _ . S 102 B 103 - | RN 104 S 10 TIME AFTER DISCHARGE FROM REACTOR {days) ' . Fig. E-1. Decay Curves for Fuel-3tream Fission Products Cooled 67.4 Days Before Draining to Waste Tank. The total heat generation rate in the underground waste tank is significantly decreased by the 67.4-day ~ holdup in the processing cell. Compare with Fig. E-2. Waste is from an ~ MSBR operating on a 58-day cycle at 2160 Mw (thermal). T - 105 ORNL DWG 65-1808 RN F T 71Tl 0000004 | I I B | ACCUMULATED —%AT GENERATION i i ii 0000 XK ! T TTL T T TT 108 T HEAT GENERATION RATE (Btu/hr) q P T R T N BATCH {4 #1% cach) 27 T 11T ¥ YEAR 2 £ !tfi%%i!l? E J G 5 24 Figa E—Eu Tank Without Prior Cooling. 103 TIME AFTER DISCHARGE FROM REACTOR {days} Decay Curves for Fuel-Stream Fission Products Sent to Waste When the still bottoms are drained directly to the underground waste tank, the total heat generation rate is about 2.8 times that shown on Fig. E-1, in which a 67.4-day cooling is allowed. cycle time and power of the reactor is the same as for Fig. E-l. The n_lo6_q waste salt accumulated whose heat generation rate 1s about L. 76 X 107 Btu/hr, obtained by extrapolating the upper curve of Fig. E-L. Allow1ng the coolant to undergo a temperature rlse of 125 C, the ailr regquirement is: Btu 1o F X. 1 This 1is equlvalent to 85,190 ftB/mln at an average temperature of 87 C. 1.76 x 10/ = 3.13 x 107 1o/hr . | The requlred Cross sectlon for alr flow is: 85 100 £63 sec oo e e - 23.7 £t . ThlS area can be ebtalned w1th 89M tubes 2.5 in. in diameter, with a 9-gage wall, in the flnal &e51gn, Draw1ng 58080 D, Appendix ¥, 937 tubes were used. These tubes, arranged in a Uatube eonflguratlon, can be accommodated in a 'l6wftmd1am tank | | Waste salt is stored on the shell side in the tank. The available - volume per foot of tank lS: 2 LE . (No. tubes)dt e tank H C@0e)® [ (e =325 | ar p a3 o [l (16 x 12)° ] e/ The heat transfer surface per foot of tank height is: £° > = (2)(937)(0.577), F—= 108L.3, £t7/ft . - = S 2. Therefore, the surface avallable to each cubic foot of waste 1s 7.88 £t /ftB. “The'mean temperature difference between coolant and salt is 65100, and the overall heat transfer coefflelent is estimated to be 6.2 Btu hr -1 ftmg O, =1 TFT The maximum allowable heat generatlon is then: 6.2 gti 7 38 gfi (651)(1 8) °F = 5.75 x 1oh Btu hr“l;ft‘3 . hr ft F ffi ' D _'rAt all tlmes during waste aocumulatlon, there must be suff1c1ent volume o present so that the integrated heat generation (Btu/hr) divided by the total volume (ft3) does not exceed this figure. | - 107 - Volume of Diluent The total volume of fluid in the waste tank at any time is the summation of individual batch volumes from the process plus the required diluent. Expressed mathematically, N v(t) = % (h+7V ) n=1 where Vdn ig the reguired diluent volume for the nth batch, and the summation is carried out over N batches. The above paragraph shows that the total volume at any time is the quotient of': integrated heat generation rate (Btu/hr) allowable volumetric heat generation (Btu nr ft—S) a(t) = I 5.75 x 10 The required diluent volume for each batch can now be determined by solving alt) - 3 (h+v, ) . 5.75 x 10 n=1 The value of q(t) corresponding to the nth batch is read from Fig. E-1. It is apparent that the largest diluent volume is required when the first batch is drained to the waste tank. Eventually the tank contains sufficient volume so that no further inert diluent is required. This behavior is shown on Fig. E-3. The total volume of diluent is 264 ft3, meking the 30-year volume of waste plus diluent equal to 78L fts. In actual practice, the total diluent volume would probably be added at the beginning of waste collection rather than in discrete steps, as shown in the figure. Waste Tank Design The 937 U-tubes are installed in the tank with one end open to the interior of the vault, and the other end welded into an exhaust duct leading to the stack. Air forced into the vault passes over the outside of the tank before entering the U~tubes. This design provides about 6100 ft2 of tube cocoling surface, which is about 2.5 times the calculated 108 - "ORNL DWG 65-3042 400 = " ACCUMULATED 3 TOTAL VOLUME 300 i £ ACCUMULATED 5 DILUENT VOLUME 200 VOLUME (") 100 e o ' TIME AFTER DISCHARGE FROM REACTOR (days) Fig. E-3. Proportions of Diluent (NaF -KF') Required to Tnsure Suitable H Heat Transfer from Waste Tank. First batch of fuel -stream waste requlres *'largest proportion; no dilution requlred after Thk days . . | - 109 - requirement. The relatively large number of tubes and the small waste volume lead to a tank that has a low height ~to-diameter ratio. The tank is 16 £t in diameter and 6.33 ft high and has a storage volume of about 86O(ft3. Monel was chosen asg the structural material. Underground Storage Facility The underground storage area is shown on Drawing 58080-D, Appendix F. In addition to the vault for the fuel -stream waste tank, the area contains a storage vault for solld Na¥ and Mg’F2 wastes from the UF6 sorption step. The design of this portion of the waste system is discussed in Appendix B. Fertile Stream Waste System Design bases used for fertile-stream waste-storage were: Thirty -year accumulation in a single waste tank. Fertile stream power of 62 Mw (thermal). Only one blanket volume (1783 £t3) discarded in 30 years. Fission-product heat generation as shown on Figs. 3a and b. Cooling by natural air convection. V1w O It was estimated that the integrated heat generation rate at the end of the 30 -year filling period would be only 5.9 x th Btu/hr, and, if the storage temperature is allowed to be as high as 900°F, this heat can be dissipated by about L0 ft2 of cooling surface. Therefore 1t 1s only necessary to place cooling surfaces over the tank cross section in locations that shorten the path for heat conduction thrcugh the salt. Twelve L-in.-diam pipes equally spaced over the cross section are provided to remove internal heat. Regions of salt most distant from a cooling surface might be molten during some period in tank lifetime, but this will not present a corrosion problem because cooling surfaces will always be covered with a Ifrozen salt layer. The storage tank is 13.5 ft in diameter and 13.5 It high, providing about 1900 ft3 of storage volume. ©Stainless steel can be used in the construction because only occasionally will molten LiFJThFhufission product mixture contact a metal surface. The tank is contalned in an underground concrete vault as shown on Drawing 58081-C, Appendix F. - 110 - APPENDIX F. PHYSICAL-PROPERTY DATA AND DRAWINGS This Appendix contains the following information: Table F-1 Thernal Date for LiF, BeF,, Na, K, and NeK Figure F -1 Calculated Density of MSBR Fuel Salt and LiF Figure F -2 Calculated Density of MSBR Fertile Salt Figure F-3a and b =~ Vapor Pressure-Temperature Curves for Several Metal | ' Fluorides ' Figure Tk Vapor Pressure of Nak Figure F-S. 'ViSéésity.and'Thermal Conductivity of LiF Figure F-6 ViscosityTemperature Curve for NeK (22.3-77.7 wt %) : Alloy : : Figure F -7 Properties of NeK (22.3-77.7 wt %) Alloy Figure F -8 | Procesé_Flowsheet for Fuel and Fertile Streams Figure F-9 Uhdergrbfind Storage System for Fuel -Stream Waste Figure F-10 'Undergrdund Storage System for Fertile Stream Waste Figure F-11 Arrangement of Processing Equipment for Fuel and Fertile Streams Table I'-1. Thermal Data for LiF, Bng, Na, K, and Nak Latent Heat of Veporization (cal/g) LiF 1690 (ref 38; calculated from vapor pressure data) BeF', 1070 (ref 11) Na. 1038 (ref 6) K 496.5 (ref 6) NaK (22.3-77-7 wt %): LiF 617 (calculated) Heat Capacity (cal/g °¢) 0.598 (ref 37) DENSITY (gm/c:mg) ORNL DWG 65-3045 Ri TIT - 1.96 \ “a, o mE 2,00 ~ < NFUEL SALY 1.92 ~ § 1.98 \ : \ 17800 500 500 1.88 o < ~ t (°C) S _ N i \ 3084' \‘ N N _ N | N MP = 8450 1,80 AY \\ . %, _ < - . xf 1.76 ~N 1,72 N < FUEL SALT COMPOSITION \ i LiF ~ 68,5 mol % 7 BeFy - 31.2 mol % 1.68 UF, - 0.31 mol % i MELTING POINT = 500°C 7 1.64 < 1.60 400 500 600 700 800 900 1000 1100 1200 Fig. F-1. TEMPERATURE (°C) Calculated Density of MSBR Fuel Salt and LiF. 1300 DENSITY (gm/em") 410 4.60 4,55 » | _4_.;45. ‘4.-_45 ;-4;35 4,30 4,25 4.20 4,15 4,05 4,00 500 -2 - = . o . /MP 565°C - (1049°F) ORNL DWG 65-3046 \ it = = 4,993 - 0000775 + _- °C . Lif ThF4 233y 0,012 233pq 0,022 " 71 mol % {17.11 wi %) (82,79 {0,035 (0,064 29 i 600 700 800 TEMPERATURE (°C) 200 1100 1200 - Fig. F-2. Calculated Density of MSBR Fertile Salt. 1300 - 113 - ORNL DWG 65-2992R2 104 S 9 8 7 8 5 3 2 MELTING BOILING POINT(°C) POINT(°C) AT 760 mm 3 CsF 684 1251 10"~ BeF, 803 1170 . RbF 798 1410 , SEE= S LiF 845 1676 P RE=n = UFy 1035 1450 _4_?) ‘-‘.‘ AN CSF - - g8 b \E.a A % S 3 % N (%) % n 3 = (W) & X SN o 3 % Q N Nertrbel2 AY % A $ X, \ - \ A ~LiF * c % 3 S UF 4% %\ 102 : . 2 3 % “ R‘.)F 3 8, % s £ -l 7 = 3 =i 6. m = 5 % 5 A3 Y % “ 2 4 a R X 2 3 n > - “ \ LY h:Y 2 Py A A \ ) '8} [§) @) p— = N 4 § 9 TS, b & 51 iy ) & & — VIS N STy 2 5.5 6,0 6.5 7.0 7.5 8,0 8.5 9,0 104/1 €10 Fig. F-3a. Vapor Pressure — Temperature Curves for Several Metal Fiuvorides. - 11h - &4 L'e} fiD.\ra\n ; .\. n\..-U1 o (D06/9)— & | - o ¥ ) T Yiin St il AN © M I =T - ; ] - G _F.M% - lfmlm!.f.r 4 = . . L o 0 y =M r > oy o Ll E » oLzl m F i 3 i Lt | L, X 2 O { I i Afb_ 4 ot 5 2 Ly i ] i o - < i -l i = ML (D008~ ” P i il i & b 4 [a S S y; , T (20868} o o e I i Lis RS h\ Y | ] L5 : 2 % _.FA;M. nl\ T " - “a i al 17 Y e ST (et06) R il - 3 Lot 5 I L L - - i [ oK : L i 1] .aml kl g e |- Bl ot i 5 ) ‘ & | %" 1. St (DolL6)® o i : o4 = 3er i | EMM. i % .” % L Ea \“E. > T » Hil S (D090 L) ! P Hits, I Jof% pr : Qo1 )} : _ - o © =¥ P Py W o~ W i} & o Ly o® % e b o flw tlo nUo oo g <3 ) [an] (B ww) JYNSSI¥! YOIVA 104/1“ e Vapor Pressure - Temperature Curves for Several Metal Fig. F-3b. Fluorides. VAPOR PRESSURE (mm Hg) Composition: 77.7 wt % K, SE; 22,3 wt % Na. ? i " Dgta from Reference 6. i 4 ! I ] v i) j . 109 i 8 7 [ 5. 2 9 —] N : P - = 45 6 7 8 9 10 11 12 13 14 15 104/1 (°K) Fig. F-4. -Vapor Pressure of NaK Butectic. THEMAL CONDUCTIVITY (Btu/hr t °F) 396 20— VESCOSETY (cehfi:po_iséé) o - 116 - : '4.04 . ORNL DWG 65-3049 402 G CALCULATED " BY METHOD ~ OF REF. 35 39| ' \\ w0 w9 S : ]‘EM?ERAT_URE_';(O 0. 950 __moo' Q) 1050 - 1100 K \ uicp) = 0.0586 3936/T T = (Ref. 36). __ L6} AN N 1.4 DN : 102: : - i.?’o_. 800 850 _.900 e 950 ~ TEMPERATURE (°C) 1000 1050 _'1 100 Fig. F-5. Viscosity and Thermal Conductivity of LiF. - 117 - ORNL DWG 65-3051 0e54 1 0.50 0.46 0.42 0.38 0.333 743 o/T \ plcp p0-333 743 o/ 0.34 p = g/cm ] T = VISCOSITY (centipoises) 0.30 0.26 ' \ 02 N ~_ 0.18 - \ 0.14 100 200 300 400 500 600 700 800 TEMPERATURE (°C) ‘ Fig. F-6. Viscosity-Temperature Curve for NaK (22.3-77.7 wt %) Alloy. (Data from Reference 6.) - 118 - ORNL DWG 65-3050 16,5 DENSITY 15._.5. ; | THERMA L CONDUCTIVITY - 15.0 < 14.51 THERMAL CONDUCTIVITY (Btu/hr ft °F) o B N 14.0— 300 400 . TEMPERA - 500 TURE (°C) 600 700 800 90 0.23— D > 0.1 HEAT CAPACITY (cal/g °C) -0.80 10.78 0.76 0.74 0.72 0 - TFig. F-T. -Referegce_6,) 200 400 3000 400 - TEMPERATURE (°C) 500 600 700 800 Properties of NeK (22.3-77.7 wt %) Alloy. (Data from Jo.g2 DENSITY (g/emd) S900 - 119 - VENT TO OFF GAS e _Fa RECYCLE Fe RECYCLE VENT TO 1 z OFF GAS l l\ | | r { 1 Tiguio N | R ' § ! & NET FiSSION POWER MET FISSiON POWER | Q 2 ud 28 Mwe 972 Mwe | by Q 9 LS e S Q) % 06’ ? % | u ’g’"‘ 5? | 5 %————!powm PLANT o292 12008 [power PLANT l—i > 3 l[ Z By o | FERTILE STREAM FUEL STREAM 157 FF¥d I 5 Volume =1783 Fi® pavg - 1.93 Uzs5 - %ifgzsbfg%d i Y% " I uU-233 =627Kg Yolume =671 F3 OtherU = 1.002 Kg/d | Fs UF, | Pa-233 =lIOKg u-233 =735.9 Kg LiF =33.,8 Kg/d -6 o | Th =141, 290Ky U-238% =Hi Kg BeF, =277.6 Kg/d L P, | LiF = 38,765 Kg Other U = 58.1 Kg 7 ; . £4,49 ; 218538 K : E Fos Hoo'E 4 !é;;‘g = 52099 K? E G@;’;fig& | ' £ Fuel wh =38700 Kg . | indl | T100°F. |O00°F. 00468 %4 | _CONTINUOUS _ STV /j\ | LiF- 2.26 Kg/d i w U=2.85 Kg/d _FLUORINATION BeFr 0.017 Kg/d | ; \358% u Rzmovai) Th® G422 Kg/d oGO F SURGE : LIF=1762 Kg/d TANK r U-233+1.24 g/ o/ | : Ps:23372,179g/) U233+ 38,715 | ! h = ] u-235= | T 27%8g/t 564 | Other - 3.057 - 42,560 A S iR LIF = 1028 FLUORINE SUPPLY BeR, = 847 suppLY [ st U233 *2.822 %g Al | o e o163 R¥d | ¥ LiF-3.54Kq/4 gre— 4 n Th*=12.8 Kg/d OFF-BAS LiF = 336.8Kg/d DISTILLATION LB2Fn057 o/ MAKE UP DISPOSAL BefF,® 277 .8 Kg/d 1666°¢C 1E2233= 10040/, Th* I575 Kg/d 238+ 0.02% Ke/d £.mm Hg WASTE LiF= 3,54 Kg/d__ &m;?fiwi?fi"? ‘;3 | [DISTILLATE St [COLLESTION [U-238 = 178 g/d | 0.0468 Fi¥d D‘f;&g;&%’_ uF v 2.2¥ o Befe 0.0IF Kgld | U-233 = (:018 Kgrd U-2s5= o.0002 g/ Ofherii=0.000 Kg/d [ WASTE COLLECTION _UFg SALES | i U-233 +12. 650 Kg/d U-235= 0.190 Kg/d U-233: 2.644 Kg/d| Otherl= &-999 Kyg/d REFERENCE DRAWINGS NO. { LFg Recycle %o fuel stream DAK RIDGE NATIONAL LABORATORY OPERATED BY UNION CARBIDE NUCLEAR COMPANY DIVISION OF UNION CARBIDE CORPORATION CAK RIDGE, TENNESSEE LTS OK OIMENSioNs uness | MSBR. INTEGRATED PROCESSING PLANTSOC 4500 P PROCESS FLOWSHEET DECMALS & FUEL & FERTILE STREAMS AMGLES & o CHEMTIED TEPDIED ! APPROVED sonsNons oD T Tseom (01 Fig. F-8. Process Flowsheet for Fuel and Fertile Streams. - 120 - COQLING AR INLET COOQLING AIR INLET { {8'x9" DUCY) y . {za"x 30" oucy JffT 7 ol AlR SUPPLY DIFFUSER (tye.} ) 3 ¥ 124 14 ARRAY 10"DIA. CYLINDER { SEE CYLIHDER BETAIL) ; 465 0" DIA. % 6'-4" HIGH I5-6" WF BEAM ON U'-6" SR 18°-0"LG. : COOLING - U-TUBES ————ToT i {932 TUBEg-~2:1/4"0.0. —5‘% b = : Of. 3-3/4 " CENTERS ) STD. RING JOINT (METAL) GE 342710 1 4.3/47 0.0, FoR E.VZHS%A;NG JOINT LMETAL ‘ééoé';ém'"e doi foa' 10" FLANGE L : " $ ROD K ' l - vy 1 . i Lyt TA({:'KpV;’ELD 143 | / KEH:TL% RA Dsp LED) H - (171 /i3 K | *—*=4‘\~fi--:f~—3“ SCH. 40 PIPE 2 : WT:-— : , : S T R\ : 10*-150™ FLANGE . “ - : 10" SCH, 40 PIPE 2.0 . 8 %%g')f’ERS ~COOLING AIR COOLING AIR EXHAUST . i o i : EXHAUST IO RIO"DUC (28" 30" DUST) ; | 2l.0“ !8'-0“ 2';051 22!_ Ou 43"9“ 1 L T i : = L ag.o" HW ] PLAN 8-0 TOP DETAIL r-4" REMOVABLE CONCRETE - e e S AT T T O s PR ; . SLABS — : _ T e~ ! A N ] N % L4 Kl /8" § 7 N { 4 i 1 S . 180" T e =3 / ‘\2 it ; ; BB . 2T IS e s E fi# - _u.}__w_. /4" THK ST.S. ) SUPPLY DUCT &" WF BEAMS AR i LINER 3.0 ML (typ.) t2'-G"CENMTERS) ' . : ELEVATION REFERENCE DRAWINGS - N, o . OAK RIDGE NATIONAL LABORATORY - OPERATED BY _ Union QarsiDeE NUCLEAR ComPANY - DIVISION OF UNION CARBIDE CORPORATION CYLINDER DETAIL OAK RIDGE, TENMESSEE LIMIYS ON DIMENSIONS UNLESS ; 4 LBG. | L SN Quesions MSBR INTEGRATED PROCESSING PLANTIS4500 %0, REVISIONS _ : : ' FUEL STEANM WASTE SYSTEM - . DRAR ) 3] DECIMALS % o ' L ' foird,G.K. {3-26-65 o o o : W ANGLES = SUBMITIED APFRENED AFPROVED i 5-CDS -B5i —— oH T GATE PP " APHRVED REV] ___Eégmfia_. a0 & 1 [ Tseosd[p[™ Fig. F-G. Undergrouna;Stprage $ystem_for Fuel -3tream Waste. -~ 121 - DUCT OPENING TOBE 13'-8" DIA. x STAINLESS . e =9 ol 2o ____E‘- 16" -0 " em—z"g" E . W 2-o" 10"x 10™ COOLANT INLET LOCATE CENTER 1-C" BELOW TOP PLUG. 3 ".o" : FROM FLOOR-—— 136" W, x 3/8"TH - STEEL TARNK ) 19-0 15-Q" 12-COOLING TUBES {4” SCH.40 PIPE} EQUALLY SPACED OVER TAMK DIA, -y 4 -WASTE INLETS (t"SCH. 40 PIPE)? ; ¥ 2"0" H k] i74" S8T. LINER 3-0" HIGH AND OVER BOTTOM —f- . AN 10" 10" COOLING EXIT DUCT CENTER 1-0" FROM BOTTOM OF PLUG. s o0 " " e . CERC LA MIAMY 5. R ° . LA 138" > L i o 3 o 8o o o s ©a? © o o 0 o e BT S 3 ©o oo 3 | LA oo Tl 6n=,:_‘,C' H [ A PRI S | B i va o H P UL W - [ LT 20-6 0" g osas o . 1A s 4 16-6 %", . P ; '\q°° i FIN, o 94 RN 2%, H RIS RS H e o H Tt e i ey i cre® i o ® o ' as® * ": ey «%0 ) 2-0 o 6" WF BEAMS {5-REQ'D) FRACTIONS £ NO. REVISIONS | oare [areo] ape0 FERTILE STEAM WASTE STORAGE SYSTEM DRAWN DATE | SUBMITTED 1 DATE | APPROVED | DATE DECIMALS + Ford, G, &, ;4-5-656 : DESIGNED DATE APPROVED DATE APPROVED | DATE ANGLES * SUBMITTED APFROVED CHECKED DATE | APPROVED | DATE | APPROVED | DATE scaLE: 174"= 10" APFROVED 1 |5808| |C IREV‘ ELEVATION REFERENCE DRAWINGS NO. 0ak RIDGE NATIONAL LABORATORY OPERATED BY UNioN CARBIDE NUCLEAR COMPANY DIVISION OF UNION CARBIDE CORPORATION OAK RIDGE, TENNESSEE LIMITS ON DIMENSIONS UNLESS OTHERWISE SPECIFIED: MSBR INTEGRATED PROCESSING PLANT BLDG. No, 4500 Fig. F-10. Underground Storage System for Fertile Stream Waste. 122 - 00 olg - : 2379 prie .70 P 828 ul i . : i i e AR A AT et ol R : : i ¥ v I . I [ 7 & i3 ¥ -, : ¥ 4 o Ry a;’.,‘ ‘: 2 Hoa o e YR B oo By i » L - 4 N V ? Vo' B LILAEN o i, "‘%‘ ¥ ¥ o AR g ‘cg’ f WEF * Fen, i = g PROCESSING CELLS— . : S REACTOR CELL / 4 5 e R, u ; v o W AR-NaK ¥ -9 HEAT i = 3 G EXCHANGER o 4 3 . i -2 4 f‘; / “P‘ i b 0 Pl VI e i Ry e Ve By TeF gk TS FE a’ !’" " Lo P ' o - H— - -0 "1 REACTOR HEAT ¥ EXCHANGER a o B G v, B D B i P e TR 1%k B ”J C Nak T i 7 SUPPLY | 7 it A - % i * ) b . v ! : " : PUMPS L] 4 5 1 > = 3§ REACTOR 3 El B ; i 1] ' - 2 2 e B o o0 o0 Ak b Rl ) A R > [ A £ 4 : {5 : SUPPLY AND E =3 SO @ \;) 3 : > % 3By - i 7 o SURGE 1w 2 3 MOAKEUP AREA 14 o / TaNg 17 A S — ; > 4 { O b - b 2 7 i FLUORINATOR ; i 2 i R - . I o NOEEN o avee ) f\} B PRI i N O 7 O TNTANKGT Y b - ; r ot & Y \o ot A B K,/ y TE| 151 ClurgTrae ~ soRBERs = 1% 4 g a . A A e T S S E PN S A Y TP e ey L Al B} L3 Ry F t—i ’ i P! . - g e P e o e el e n R e e ¥ %) b . CT} \) F : {i_t B B & g 15:é,-_ & v ooE | 2O OO0 % % o My O O B “ KoY KOH COLD TRAPS A e © Dol Z1TANK SCRUBBER STILL ; ; sPPLYE @ [F 3 L s DD W WG4 _CONDENSER == ook o wiSTe " i i O S ! OGN K Fp E UFg=LiF g a0 5 5 HYDRATOR REDUCTEON R ; LR oF SORB & & He CH® 5 ? 5 1 i - SUPPLYD O I, Iy % Pels £ ©OF COLD TRAPSZ o =\, P P 2 S Q¢ F | FERTILE ,’ W, s i - > . ‘ STEAM | VI e e o a S LI UFg fl: ! MAKE-UP N 7 " s - P PRODUCT (y [ 7 PRI gTER 2 | & 3 P B ThF 4~ LiF; > r N - S 3 * MAKE-UP A 2 o . ; z o, e e 3 REFERENCE DRAWINGS Ho. 9 [ k4 Lisa N P N W e S B o R e S T Lo S L B P LS - | n 2 ,.”ZE, OAK RIDGE NATIONAL LABORATORY : © g E Ry, = Bl s o o g B L) o -9 e SRRV 5229 _ _ . & Unton CARBIDE NUCLEAR COMPANY 20 12-9 Ny : RS DIVISION OF UMION CARBIDE CORPORATION ¥ I Bde Baiosu pinop viry poars grd B o Bt w0 Fiws s £ e el te S . OAM RIDGE, TENNESSEE - S LNITS ON DINENSIONS UMLESS | MSER INTEGRATED PROCESSING PLANT - - BilGasog L SR ‘ OTHERWISE SPECIF e _Ne. NC. REVSIONS | patE !APPO!A?PD i AR § FRACTIONS ' PRELIMINARY EQUIPMENT W1 DAIE | SUBMITIED | BATE ] AT ' S MALS + ‘ FARMER §3-15-65] ! ! . S e LAYOUT ~TERGRES | DRE | AHOID | DR | RROVED | OATE _ R ST T P — SUBHTES REFRONED APPRONVED 14 ! 1 . L CHECKED | DATE | APPROVED | DATE | APPRGVED | DATE S _ : R FOPROVED ; y TV, : ! ! , = SOME3/16" =170 [ Tseos0]p [™] Fig. F-11. Arrangement of Proce531ng Equlpment for Fuel and Fertile Streams. The highly radioactive operations in fuel-stream processing are carried out in the smaller cell (upper left). The other cell houses equipment for the fertlle stream and the cooler fleld stream operatlons. - 123 - ORNL-3791 UC-80 — Reactor Technology TID-4500 (46th ed.) INTERNAT, DISTRIBUTION 1. Biology Library 65. C. E. Larson 2-4. Central Research Library 66. R. B. Lindauer 5. MSRP Director's Office 67. H. G. MacPherson 6-7. 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