X822 OAK RIDGE NATIONAL LABORATORY Operated by UNION CARBIDE NUCL EAR COMPANY Division of Union Carbide Corporation uee Post Office Box X QOak Ridge, Tennessee DATE: January 13, 1959 SUBJECT: A Preliminary Study of a Graphite Moderated. Molten Salt Power Reactor TO: Liszted Distribution FROM: H. L. D. E. M. Lo Jo G. Jo Ga G. B“ w‘ E. A, W. D, MacPherson Alexander Grimes Kinyon Lackey Mann Miller Whitman Zasler Abstract g ¥ » : f ;i ~& . . ) 4 . STER COPY MAL - External Distribution Limited ORNL CENTRAL FILES NUMBER 59-|-=2¢ COPY NO. -{é A preliminary desjgn and cost study has been made on a one region unclad graphite moderated molten salt power reactor. Included are conceptual plant laysuts, basic information on the major fuel circuit components; and a discussion of the nuclear characteristics of the core. For a plant electricsl output of 315,000 kw and & plant factor of 80 percent, the energy cost was approximately T.k mills/kwh. NOTICE This document contains information of a preliminary nature and was prepared primarily for internal use at the Oak Ridge National Laboratory. It is subject to revision or correction and therefore does not represent a final report. The information is not to be abstracted, reprinted or otherwise given public dissemination without the approval of the ORNL patent branch, l.egal and Information Control Department. 1. General Features of the Reactor A power reactor of the molten selt type using a graphite moderator achieves a high breeding ratio with a low fuel reprocessing rate. The graphite can be in contact with the salt without causing embrittlement of the nickel alloy container. The salt selected consists of a mixture of LiF, BeFE, and UFh (70, 1o, 20 mol %), melting at 9%32°F. The uranium is 1.30% enriched. The core is 12.25 feet diameter by 12.25 high, with 3.6" dismeter holes on 8" centers. 16% of the core volume is fuel. The choice of the power level of this design study was arbitrary, as the core is cagpable of operation at 1500 Mw(t) without exceeding safe power densi- ties. An electrical generator of 333 Mw(e) was chosen, with 315 Mw(e) as the station output, which requires 760 Mw(t). The heat transfer system includes a fluoride salt to transfer heat from the fuel to either primary or reheat steam. The salt selected has 65 mol % LiF and 35% BeF,, which is completely compatible with the fuel. The Loeffler steam system, at 2000 PSI, 1000°F, with 1000°F reheat avoids the problems associated with a high température fluid supplying heat to boiling water. The fuel flow from the core is divided among four circuits, so that there are four primary heat exchangers to teske care of the core heat genera- tion. Two superheaters, one reheater, three steam generators are required for each circuit. This arrangement is based on the practical or economic size of the respective components. While it would have been possible to design this graphite moderated molten salt reactor plant identical to the homogeneous plant described in ORNL 2634, Molten Salt Reactor Program Status Report, an effort was made to include new designs evolved since then for a number of festures and components. These include the meintenance concept; heat exchanger design, fuel transfer and drain tank system, gas preheating, barren salt inter- mediate coolant and the Loeffler steam system. In most of these, the actual design chosen for a plant will not greatly affect the overall economy and operation. It is highly probable that the -3- Table 1 REACTOR PLANT CHARACTERISTICS Fuel 1.30% U°>°F, (initially) Fuel carrier 70 mole % LiF, 10 mole % BeF,, 20 mole % UF), Neutron energy near thermal Moderator carbon Reflector iron Primary coolant fuel solution circulating at 35,470 gpm Pocwer Electric (net) 315 Mw Heat 760 Mw Regeneration ratio Clean (initial) 0.79 Estimated costs Total 279,250,000 Capital g252/kw Electric 7.4 mills/kwhr Refueling cycle at full power semicontinuous Shielding concrete room wall, 9 't thick Control temperature and fuel concentration Plant efficiency 41.5% Fuel conditions, pump discharge 12250F at ~105 psia Steam o o Temperature 1000°F with 1000 F reheat Pressure 2000 psia Second loop fluid 65 mole % IiF, 35 mole % BeF, Structursl materials Fuel cirecuit INOR-8 Secondary loop INOR-8 Steam generator 2.5% Cr, 1% Mo steel Steam superheater and reheater INOR-8B Active-core dimensions Fuel equivalent dia 14 ¢ Reflector and thermal shield 12-in. iron Temperature coefficient (&k/k)/oF negative Specific power 1770 kw/kg Power density 117 kw/liter Fuel inventory 35 Initial (clean) 700 kg of U2 Critical mass clean 178 kg of 0255 Burnup unlimited el features of the actual plant built would consist of a mixture of those de- scribed in this report; in the previdfia'reporta and evolved as _a result of future design and development work. A plan view of the reactor plant layout is presented in Figure 1, and an elevation view is shown ingFigure éo The reg&to:@and the primary‘heat exchangers are contained in a large rectangular reactor cell, which .is sealed to provide double containment- for any ieakage of fission gases. The rectangular configuration of the plant permits the grouping of similar equipment with a minimum of floor space and piping. The superheaters and reheaters are thus located in one bay, under a crane. Adjacent to it are the turbogeneratcr, steam pumps, and feed water;heaterpand pumps. The plant includes; in addition to the rgactor and heat exchanger systems and electrizal generation systems, the control room and fill-and-drain tanks for the liquid systems. 2o Fuel Circuits " The primary reactor cell which encloses the fuel circuit is a .con- crete structure 22 ft wide, 22 ft long, and 32 ft high. The walls are made of 9 ft thick concrete to provide the biological shield. Double steel line%sflform 8 buffer zone to ensfire that no fission gas that may leak into the cell can escape to the atmosphere and that no air can enter ‘the cell. An ifieft atmosphere is maintained in the cell at all ‘times. The fiumps, heat exchangers, and instrumentstion are so arranged that the equipment may be removed through plugs at the top of the cell leaving the fuel-containment shell behind in the reactor cell. In the reactor cell are located the reactor, four fuel pumps,; and four heat exchangers. _The fuel system, gas heating, and cell cooling equipment as well as the fission gas hold-up tanks are in connected side passages. The reactor core consists of a graphite moderator, 12.25 ft in dia—‘ meter and 12.25 £t high. Vertical holes 3.6 inches in diasmeter on an UNCLASSIFIED ORNL-LR-DWG, 35086 COOQLANT SALT rRUMR COOLANT SALT CELL SUPERMEATERS EQUIP. PEMOVAL HATCH PREHEATER EVAPORLATORS LOOFR MHEAT - ————————— - —_— — < ;;?’%EA p EPENEAT LINE NEICHE CELL —— = TURBOGENELATOR CooL/ING ACCESS LOC K & PLUM Pz HGH PRESSURE LINE VEL SALT QOOLING LINES O O AYEL T LXQMANGER (4) FUEL DRAIN TRMNKS o O | FI@GURE | - PLAN VIEW - 760 MW (t) GRAPHITE MODECRATED MOLTEN SALT POWERL REARCTOR PLANT tmxgm et 00 v UNCLASSIFIED ORNL-LR-DWG. 35087 COOLANT PUMF STERM PUMPR — B — — mr————— i e mmmnm— - — - e e — e — - - | S——— — - \ N rbmm— - r—— SUPERMHERTER - FOEL PUOMPR — HERT EXCHANGCER REACTOR EIGURE 2 - ELEVATION VIEW - 760 MW (t) CRAPHITE MODERATELD MOLTEN SRLT PLOWER BEMRCTORL PLONT™ | be—r20 " eight inch square pitch form the fuel passages. The core is mounted in a 1-1/2" thick INOR-8 cantainer. Fuszl azrters at the bottom afid passes through the fuel passages and a two inch annulus between the core and shell which coonls the shell. At the top of the reactor is the fission gas holdup dome described elsewhere. This is shown in Figures 3 and L. From the reactor; the fuel gees to centrifugel pumps of which there are four in parailel. The lower bearing is sali-lubricated, submerged in the fuel above the impeller. Above the fuel surface is a shielding section, to protect the upper bearing lubricant and motor. The beering includes a radial bearing, a thrust bearing, and a face sesl. The motor rotor is on the shaft above the bearing. The foter is canned, so the field windings may be replaced without breaking the reactor seal. KCmoling is profiided for the shielding section and the shaft. The entire pump may be removed and replaced as a unit. The coclant salt pump is of a similar design, with modifications permitted by the lower radiaticn level. The primary exchangers are of the bayonet bundle type, to permit semi- direct replacement. The incoming coolant passes through the center of the exchanger to bottom, then upward on the shell side through the exchanger, leaving inlan annulus surrounding the incoming coclant. Helical tubes are between flat tube sheets. 3. Off-Gas System ‘ An efficient process for the zontinuous removél of fission-product gases is provided that serves several purposes. The safety in the event of a fuel - 8pill is considerably enhanced if the radiocactive gas concentration in the fuel is reduced by stripping the gas as it is formed. Further, the nuclear stability of the reactor igger changes of power level is improved by keeping the high cross secticn Xe continuously at a low level. Finally, many of the fission-product poisons are, in their decay chains, either noble gases for a period cf time or end their decay chains as stable noble gases, and therefore the buildup of poisons is considerably‘reduced by gas removal. UNCLASSIFIED ORNL-LR-DWG. 35088 C;/?,Apd TE C oRE G 3 REQCTOR & PUMPS ~ FLEVEBT IOV £AL WELD ) Db b ‘E'U’ i 5Hin.p i | —~O-RING . | 1 u?‘ . ? LR " ’4‘ ' AA i Exvansidry Tank e SO N\ i 1 T~ b ’ Cc)er&;y:gQ /SALT /’Gfi ¥ Hffir/N[, S‘H:&&' ——f UL FhssAacE C7pr) {} ] Foee in UNCL.ASSIFIED ORNL-LR-DWG. 350839 F/6G 4 REACTOR & FUNTEPS ~ FLAN The solubilities of noble gases in some molten ssiits are given in Table 2, and it is indicated that sclupliiities of similar corders of magnitude are likely to be found in the LiFmBng cbeys Henry's law, so that the equilibrium soiubility is proporticnal to the par- gsalt of this study. It was found that the solubility tial pressure of the gas in contact with the salt. In principle; the method of fission=gas removal consists of providing a quiet free-surface from which the gases can be libersated. In the system chosen; approximately 50% of the fuel flow is allowed to flow intc the reactor expansion tank. The tank provides a large fuel-tc-gas interface; whicfi promotes the establishment of equilibrium fission gas concen- trations in the fuel. The expansion tank provides a iliquid surface area of epproximately 52 ft2 for removal of the entrained fission gases. The gas re- moval is effected by the balance between the difference in the density of the fuel and the gases and the drag of the oprosing fuel velozity. The surface velocity downward 1n the expansion tank is epproximately C.75 ft/sec, which should screen cut all bubbles larger than 0.06 in. in radius. The probability that bubbles ¢f this size will enter the remctor is reduced bty the depth «f the expansion tank being sufficient to aliow time for smail hubb1$s toy conlesce and be removed. Table 2 SOLUBILITIES AT 600°C AND HEATS OF SOLUTION FOR NOBLE GASES IN MOLTEN FLUORIDE MIXTURES In NaeF-ZrF In LiF-NaF-KF In LiF-BeFpn (53-47 mole %) (L6.5-11.5-42 mole %) (6Ua’6 mole %) Gas k* k¥ k#* x 1070 x 107° % 167° Helium 21.6 + 1 11.3 * 0.7 11.55 + 0.07 Neon 11.3 + 0.3 holb + 0.2 L7 +0.02 Argon 5.1 + 0.15 0.90 + 0.05 0.98 *+ 0.02 Xenon 1.94 + 0.2 - .28 festimated} Henry's law constant in moles of gas per cubic centimeter of sclvent per atmosphere. -} - > The liquid volume of the fuel expansion tank is approximately 50 ft” and the gas volume is appraximaéély 2k0 ftfi. With none of the fission gases purged, approximstely 3300 kw of beta heating from thekdeéay of the fission gases and their daughters is deposited in the fuelfand on metal surfaces of the fuel ex- pansion tank. This 3300 kw ¢f heat is partly removed by the byfiass fuel circuit ‘and. the valiancs is traneferred thréugh the expansion tank walls to the secondary | loop coclant. The fission product gases will cause the gas pressure in the reactor to rise approximately 5 psi per m@fith. Tfiifl pressure is relieved by bleeding the tank cnce a month at a controlled rate of approximately 5 psi per day to a hold tank. ({See Figure 5.) The gases in the hold tank are held until they have decayed sufficiently tc be disposed of either through a steck or a noble - gas recovery system. A small emcunt of fission gases will collect -above the free surface of each pump. These gases are continususly pfifged”with’helium. The purge gases from the pfimps are delayed in a hold voliume for approximately 5 hours to allow e large ffaétiafi of the shorter lived fission products tc decay before enter- ing the cooled carbon beds. The carbon beds provide a holdup time of approxi- mately 6 days for krypton and mnch‘langer for the xenon. The purge gasss from the carbon beds, essentially freeAfrom activity, are compressed and returned to the reactor o repeat the cycle. L, Molten Salt Transfer Equipment The fuel transfer systems are shown schematically in Figure 6. Fluid is firansferred between the reacter.and drain tanks thrcugh a presgureasiphbn Bys- - tem. Two, mechanical valvea; in series, are placed in this line with a siphon- breaking connection bestween them. ¥Fluid is transferred.frem one system to a- nocther by isolating the siphon«bresker and applying differential gas pressure to establish flow‘and finally complete the %transfer. When the gas equalizing valves in the siphon-~bresker circuit are opened the fluid will drain cut of the transfer line and the valves are then closed. UNCLASSIFIED ORNL-LR-DWG. 35090 MOTOR — MM e ; - ] FueEL DEGASSER puel i l _ ! 4 EXPANSION TAIK PUNPS vew | | [ . ~ | BY PAss ' IC%F};’ .4 = i 17 : | 39,5¢F5 ) 1225°F ]0.553 SCFM | I YR \L r |3'35 FT? l‘l £i~ jo FT3 ‘w-l A__»_N__G—N_“l/ * ' ) ‘ \ ! ! * L‘g' %m .X____w_u_‘ GAS PUMPS L-*-—‘J CORE k * * , , 1 l ; o 3T Laegre : : PUEL BUED °F o= | "‘?‘? .4%006..AUT g ; e —{BS‘QFT.’ |— — FUEL LINE éX * —-—— GAS LINE ALUE —p4—CLOSED U ] L. o To STACK OR —p<— OPE N UALVE INTERMITTENT FLOwW % X NOBWE GAS O.0LTBRLFM @ 1250°F izg{) ;:'1-,3’...1 RECOUERY FIG.S - SCHEMATIC FLOW PIAGRAM FoR REMIVAL OF FISSION PRODUCT GASES, UNCLASSIFIED ORNL-LR-DWG. 35091 INERT GAS INERT (3AS SUPPLY -t PP T - -—-—--—“l—-;M”T - VENT i | . q UE SUPPLY % SIPHON | BREAK X X PoINT | | | SAMP LING - | I ,J . ' A ENRICH\WG ¥ l —(—)——M—T_N c C ( < POINT | ISOLA TION | | | SALT - VALLOES | | l CHARA ING I—L ! . i [ | ' | I | ; | . | I l | | HEAT REACTOR : : : : EXCHANGER | ' | | | t (| oF W) | 0 | | | ' — DRAIN TANKS iy LEQ GE LD FUELL LIME —-— GAS LINE —b Y X Ly UPER WEATER (2 #a. cinceir) C ONDENSATE VM L& }‘\/ C3o, 00 Fuee SALT | | ;\j Clozs® F45° ( £ L Feow HRartées Lwown ARE FOR COMPLETE SVYITEAL Fre. &8 Feen warer - AEstens Steam Fume D — I series Boreer ].._.w — | | DEAERATOR {2, 789 ooD‘{M’“‘& \ M95p Y75 . FEFOwATER M TERS Feepwrrer AN Feow Pracram uétm =20 - as control can be achieved by vaeriations of the steam flow. The primary exchangers are o5f the bayonet bundle typefi to permit semi- direct replacement. The superheaters are of U-tube in U-shell, while the reheaters are of straight tube constructicn. The stgam generators are horizontal drums, 4 foot diameter-by 24 feet long, with 5w1/2 inch walls. These are half-filled with water, into which the siteam nozzles project for direct contact heat tramsfer. Recircuiation of steam provides the heat for genération of stean. Heat exchangef data is summarized in Table 3. 10. Turbine and Electric System Steam at 2000 PSIA and 1000°F, with reheat to 1000°F, iz supplied to the 333-Mw-rated turbine, which has a single shaft, with 4 exhaust ends. The tur- bine heat rate is estimated to be 7670 BTU/kwh, for a cycle efficiency of ' 4i,7%. The generator and station heat rates are, respectively, 7785 and 8225 BTU/kwh. The supply to the bus-ba; is 3195 Mw. These estimates are bhased c¢n Tennessee Valley Authority*heat balances for a similar turbine(l}5 with ad- (2) Justments for the modified steam conditicn and different plant reguire- ments of the molten salt reactor system. 1l. Nuclear Calculstions A number of age theory thermal resctor calceulations were made 0 sur- vey the nuclear characteristics of graphitewmsderafied mélten galt reaatfira(i). In sll cases the salt was of the compcsition LiTFwBeFEwUFk {70+10-20 mole %}, and was located in cylindrical channels spaced on an 8 inch center-to-center square array. The volume percent Qfmfuel in the reactor core was varied and calculafiions vere made for k = ;.OS and 1.10.‘ The calculations yielded the percentage of‘U~235 required in the initial invest@ry of ufaniums the dimen- - sions of the reactor required“forkkeffective to be equal ta one; and the initial conversion ratioc. Table 4 gives the results of the cslcoulaticons. Fuel and Sodium to Sodium Exchangers Number required Fluid Fluid location Type of exchanger Tbggeratures Hot end, °r Cold end, °F Change o AT, hot end;, F AT, cold end; OF AT, log mean, °p Tube Data Material Qutside dia, in. Wall thickness, in. Length, ft Number Pitch; in. Bundle dia, in. Exchanger dimensions Heat transfer capacity, Mw Heat transfer ares, fté Average heat flux, 1000 Btu/hr-fte Thermal stress¥*, psi Flow rate, ft3/sec, 1000 1b/hr Fluid velocity, ft/sec Max Reynolds modulus/1000 Pressure drop, psl * ([ OF X Max AT wall 1 - v 2 L] Table 3 - DATA FOR HFAT EXCHANGERS Primary System Superheater Reheater L 8 b fuel salt coolant salt coolant salt steam coolant salt steam tubes shell shell tubes shell tubes bayonet bundle U=-tube in U-shell straight counterflow counterflow 1225 1125 1125 1000 1125 1000 1075 %5 9%65 650 9%65 620 150 160 160 350 160 380 100 125 125 110 315 335 105 207 21% INOR-8 INOR-8 TNOR-8 0.500 0.750 0.750 0.049 0.083% 0.065 21.8 23 17.5 3173 925 750 ds 0.638 () 1.00 (a) 1.00 (A) = 33 28 50.75 in. dia x 17.5 £t long 190 81.2 27.6 5830 3315 2100 111 84 45 2000 6100 4600 19.8 16.8 7.18 2.46 90l 512 8.80 10.0 L 265 270 L7.6 16.5 23 13 23 12.5 P Table 4 Dimension of Vol fraction % Enrichment Unreflected Initial of fuel ‘of cylindrical Conversion in core uranium reactor ratio Case F Kk e D (ft) H (ft) ICR 1 0.05 . 1.05 1.30 26.3 24,3 0.55 2 0.05 1.10 1.45 17.9 16.4 0.49 3 0.075 1.05 1.25 2k.o 22.2 0.63 L 0.075 1.10 1.39 16.6 15.3 0.58 5 0410 1.05 1.28 22.4 20.8 0.71 6 0.10 1.10 1.6 15.6 14.3 0.65 T 0.15 1.05 1.53 20.5 18.9 0.80 8 0.15 1.10 1.80 14.3 13.1 0.73 ‘Q 0.20 1.05 2. 24 19.4 18.0 0.86 10 0.20 1.10 2.88 13.5 12.4 0.79 11 0.25 1.05 L.36 18.8 17.4 0.90 12 0.25 1.10 7.05 13.1 12.0 0.82 Case 8 of this table is quite similar to the reactor that forms the de- sign basis of this study. The nuclear performance of the actual reactor | design chosen for this study was calculated by the multigroup code Cornpone on the ORACLE. The neutron balance obtained under initial conditions is given in Table 5 below. Also given are the inventories of materials, based ‘on & total fuel balance inside the reactor and in the external circuit of 900 cubic feet. It should be noted that the enrichment of U-235 predicted by the machine calculation is 1.3 vs the 1.8% of case 8, Table 4, and the conversion ratio is 0.79 instead of 0.75. The lower enrichment requirement results from the higher value of eta and some reflection. The long term performance of the reactor was calculated for a case, assuming an initisl inventory of U-235 of 700 kg and an initial breeding ratio of 0.73. ~23. Table 5 NUCLEAR CHARACTERISTICS Element Inventory gkgz Neutron Absorption U-235 717 1.000 U-238 55,000 0.790 Ii 5,760 0.053 Be + C - 0.031 F 37,900 0.026 Teakage - 0.166 eta 2.07 Conversion ratio 0.79 Table 6 CROSS SECTIONS USED FOR REACTOR LIFETIME CAILCULATIONS Effective Cross Section Neutron Yield Capture to Fission Ratio Tlement barns 1 o U-235 605 2.029 0.23% U-23%6 25 U-236 2.5 Pu-239 1903 1.84 0.58 Pu-240 3481 Pu-241 1702 2.23 0.36 Pu-242 491 ~2l The calculation was m&fié by an adaptation of the method described by Spinrad, Carter, and Eggler(h). The buildup of U-236, Pu-239, Pu-2L0, Pu-241, Pu-242, and fissién*products wvas calculated as a function of the integrated flux-time variable. The reactor was kept critical by additions of U-235. The cross sections of the heavy isotcopes were taken as a consistent set from the 1958 Geneva Conference Paper P/lclé(S) and are given in Table 6. For this long burnup reactor, the fission product crecss sections were taken as & Tunction of time based on the celculations of Blomeke and Todd(é)i Xenon=-135 was cqpsifiered tq be removed continuously, but all other fission - products retainéfi. In the early phases of reactor operation this amounted to an initial 1.3% poison plus sbout 48 barns per fission. The rate of fission product poisoning then tapered off té about 18 barns per fission, a value that:was assumed for the refiainder‘of the reasctor lifetime. This (7). latter value is consistent with that proposed by Weinberg Figure 9 shows the accumuelative addition of U«235 required to keep the reactor critical. In the first few months the velue is negative, that _is, U-235 should be removed from the reactor. In actusl practice the extra reactivity would be ccantrolled by adding high cross-secticn burneble poisons, or possibly by contrcl rods. After the first few years the sddition of U-235 is about linear, and at the end of 32.5 years some 4100 kg have been added. Figure 10 shows the inventories of fissionable isctopes as a funztion of time. After seven years of operation, the inventory of U-235 starts to rise above its initial value and is approximately twice its initial value after 32.5 years of operation. It is evident thgtg on the basis of these nuclear calculatiéns, the fuel could be retained in the reactor without re- processing for the life of the reactor. The calculations are open te the criticism that epithermal absorptions by fission products are neglected and they might be import&nt as they build up to high concentrations. The effect will probably be small during the first ten years of operation, however. UNCLASSIFIED ORNL-LR-pwG. 35094 35 MWe Fased Sa/t feoc/sr * Shabtly Lrriched 9000 Grz,o/ e Modors 7o S Q & ! 2000} /000 Cumuletve U5F Addition, Ky & ) } i y 1 { O 5 10 /5" 20 25 20 35 e J 0}9@/’07,(5/7_9 @Zis YEqrs /:/-yz/re S - Cumulalive V5 Adhitor vs Opei’cj?‘mf/ oy -42- UNCLASSIFIED ORNL -LR-DWG. 35095 /500 315 MWe Fused Salf FegTlor 9htly Enriched Graphile Modera?ed 3 2 500 /:/’55//«9 /50%@;)@ /fiV@/yfiJ/‘)/j /fg" /5" TO & 20 35 0/0 erqting 7’7?’)6’) JPars F ;941;/9 /0"'/&755//@ /Sof‘w/oc’-’ //71/&/775/// VS, Opefafifiy Time —98-— -27- 12. Fuel Cycle Costs The fuel