ORNL.-2796 Contract No. W-7405-eng-26 REACTOR PROJECTS DIVISION EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR . C. Alexander L B. W M. E. Lackey H. G . MacPherson DATE ISSUED MAR 8 J. W. Miller F. C. VonderlLage G. D. Whitman J. Zasler 1960 OAK RIDGE NATIONAL LABORATORY Oaok Ridge, Tennessee UNION CARBIDE CORPORATION U.S. ATOMIC ENERGY COMMISSION CONTENTS A D S G, oot eee e e eaeee et e aeaetrr——aataaataee et aatetteaaaaeaae naaaaetataatanaaas arennaaasrarea—aseneeennaaesereasaraaes errees ] Nt OdUCTION GNA CONMCIUSTONS oottt ee et teeeementaraaeseaeesnnnnsaesentesaeseesnnsasassnnssnasssssnnsssssssnssnnnnnnnsennns ] General Description of Reactor and Plant Layout........coriiiiieeeeecee e ] Molten-Salt System AuUxiliQries .....ccoiieiieercecre e et e e e b e e s et sb e et er b s enenenas 10 Enriching and Sampling Sy stem. ...t e e st e e e sebee e ceaeerste e ebe e e sebebre s 10 Fill and Drain Sy stemu. i it ettt sttt et et e e te b s e s asee st araseaseesaessneetesessenseeseesnseseans 13 Preheating and Afterheat Removal Equipment ..ottt 14 OFf=Gas SYSTEM .ueoueniiieiiie et sttt ettt sttt e st et eet et et e st e see et et st e sase et ae e ate b e srerttaueeesaesa e st aesnnennres 16 Molten-Salt Pumps and Heat Transfer EqQuUipment ..ottt s 16 Steam Generating Equipment and Turbine-Electric System ........ccociiiniii 20 R MO MO IM O AN CE oo ieeeeee e ee et eseesaeaaeassesansesse s esaas nnssanasnsnnnssssesssnnnsnsesassssssassnssnessessseesantennnnnnnnseessesessnnsaneeens 20 N UG A P el OrMaNCE ooeieeeeeeieeeee e eeeeeseeeessaereassasseseasessessnstensnssaessessesnsssensssnesssensssassssssseeessssssssssssessssaeeersnnesssane 21 R @A CTOE HOZATAS oo ieeeeeeiieee et et e e e eeeesseeaes s seessseseessnasesnsssesessnsnsessnsnsssssnseesssassnssssssessssnsenrasssnstessnnsssennnssssnsnnsenssnnens 23 ST oot eeeaeeeees e attaaaearareeturann__——___aseseeeseteettuternn—_——..—aesieaebabena—ateetaenraaa aeeneteeaeetereanteaneteetsanasaassanes 26 A EINAEE D STIGNS.ccuiitiitiiitiitieiirte ettt e ettt b bbb e E bR b s R b b s R s h et et eb et et 28 L ist OF DIaWINGS cuviireiririreeieteiiert st st et sees et bt st et s b ee s b sh et b eb s e Re b e bt e b e st b eR e e e s d e bR sa bbb s s 32 EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR L. G. Alexander B. W. Kinyon M. E. Lackey H. G. MacPherson J. W. Miller F. C. Vonderlage G. D. Whitman J. Zasler ABSTRACT A preliminary design study has been made of an experimental molten-salt-fueled power reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler steam-generating cycle. Conceptual plant layouts, basic information on the major fuel circuit components, a process flowsheet, and the nuclear characteristics of the core are presented. The design plant electrical output is 10,000 kw, and the total construction cost is estimated to be approximately $16,000,000. INTRODUCTION AND CONCLUSIONS The molten-salt-fueled reactor system described in this report represents a preliminary design and is to be considered as a reference design for further experimental molten-salt-fueled reactor studies. Designs have been developed for the major components which are sufficiently detailed to permit an initial evaluation of costs and con- struction problems. Information on nuclear per- formance was obtained to give a basis for study of the major problems involved in operating a molten-salt-fueled power reactor. The molten-salt concept has been considered for a variety of reactor types. These may be classified as homogeneous or graphite-moderated systems, in the main, with a variety of modifi- cations depending on particular objectives. Single- and two-region homogeneous reactors have been considered in burner and converter cycles, and, most recently, unclad-graphite-moderated reactors have been emphasized for breeding cycles. The major aim of this study was to design a nuclear power plant which, with a minimum of developmental effort, could be built in the near future and would provide considerable information applicable to larger or more complex molten-salt power plants. With this objective in mind, a single-region homogeneous burner employing a semidirect-maintenance concept, operating on a Loeffler steam cycle, and producing 10 Mw of electricity was chosen as a reference design. The system embodies a simple core from which heat is transferred through a coolant salt to a steam system to produce a useful amount of power by a proved cycle. The semidirect-maintenance philosophy circumvents the more complicated oper- ations involved in complete remote-maintenance schemes, but does not prevent the replacement of the more vulnerable components of the fluid-fuel system. This reactor plant could be used to demonstrate the nuclear performance of a homogeneous core, the reliability of components, the stability of the molten-salt fuels, the handling of gross quantities of molten salts, the chemical processing of fissioned fuel, the containment and processing of fission gases which would be stripped from the fluid fuel, the application of a particular high- pressure, high-temperature steam cycle to the molten-salt system, and the practicality of certain maintenance techniques. In addition to the primarily technical data, cost information would be obtained on the construction and operation of a molten-salt-fueled power reactor. |t has been estimated that this experimental reactor would cost approximately $16,000,000 and that about two and one-half years would be required for construction. The general characteristics of the system are listed in Table 1. GENERAL DESCRIPTION OF REACTOR AND PLANT LAYOUT The molten-salt-fueled reactor system adopted for this study incorporates a complete electric generating station but does not include on-site fuel reprocessing facilities or a major-equipment remote-maintenance repair shop. The successful operation of an experimental reactor would not Table 1. Reactor Plant Characteristics F uel Fuel carrier Neutron energy Moderator Primary coolant Power Electric Heat Estimated cost Refueling cycle Shielding Control Exit fuel temperature Steam Temperature Pressure Intermediate coolant Structural materials Fuel system Coolant system Steam superheater Core diameter Temperature coefficient of reactivity, (Ok/k)/°F Specific power Power density (core) Fuel inventory Initial (clean) After second year Critical mass (clean) depend on continuous fuel reprocessing, and, indeed, long-term exposure of the fuel carrier would be required to achieve fission product concentrations comparable to those expected in prototype molten-salt-fueled power systems. Small quantities of fuel could be withdrawn from the >90% UZ3F, 63 mole % LiF, 37 mole % BeF2 Intermediate Lil"'--Bel"'2 Fuel solution circulating at 1480 gpm 10 Mw 30 Mw $16,000,000 Semicontinuous Concrete Temperature and fuel concentration 1235°F 1000°F 1450 psia 65 mole % LiF, 35 mole % BeF2 INOR-8 INOR-8 INOR-8 6 ft ~6.75 x 10~> 417 kw/kg 9.4 kw/liter 65.8 kg of U235 107 kg of U233 40.6 kg of U239 drain system and transferred to shielded flasks, which could then be transported to the ORNL facilities for experiments in chemical reproc- essing. One of the primary objectives of the reactor experiment would be to determine the metallurgical and structural reliability of components. Major remote repair facilities were not included in the design, but on-site decontamination and de- structive disassembly facilities were provided so that manageable specimens of the fuel components could be obtained and transported to ORNL hot shops for inspection. |t is intended that part of the hot storage cell be used for such work. The fuel, LiF-BeF,-UF,, is circulated in a single-unit primary container by a centrifugal pump, and a single coolant-salt system containing LiF-BeF, is used as the heat transfer coupling between the fuel and the steam superheater. A salt coolant has several advantages over a liquid metal coolant in this application, such as com- patibility with the fuel salt, lower induced radio- activity coupled with a much faster decay rate, ! nonflammability, and elimination of cold-trap cir- cuits. The major disadvantage of the coolant salt is a relatively high liquidus temperature (865°F) which increases the preheating load and increases the danger of freezeup at off-design conditions. A Loeffler steam cycle is used because of its unique adaptability to the molten-salt-fueled re- actor. The relative merits of this system have been described in connection with earlier molten- salt-fueled reactor studies. ? The primary fuel container is made up of a single structure which forms the core, heat ex- changer, expansion tank, and fuel pump wvolute. This assembly, along with the gas heating and cooling jacket, is shown in Fig. 1. The primary fuel container weldment is shown in Fig. 2. The fuel is recirculated through the system by means of a sump-type centrifugal pump. The fluid enters the upper heat exchanger header from the pump discharge and flows down through the tubes. After passing through the heat exchanger, the fluid enters the 6-ft-dia core via the central pipe and flows to the bottom of the sphere. The flow then reverses, washes the reactor vessel wall, and returns to the expansion tank region through the annulus surrounding the heat exchanger assembly. The flow is directed into the pump by means of a horizontal baffle located on the level of the D, J. McGoff et al., Activity of Primary Coolant in Molten-Salt Reactor, ESP-X-400, Oak Ridge, Tenn,, MIT Practice School, UCNC-ORGDP (1958). 24, G.MacPherson et al., Molten-Salt Reactor Program Status Report, ORNL-2634 (Nov. 12, 1958), pump inlet. The entire upper region of the fuel container assembly serves as a fuel expansion tank. Some flow is directed up out of the heat ex- changer inlet header for fission-gas removal and is recirculated in the expansion tank through the pump. The fission gases are purged from the fuel and held temporarily in tanks located in the reactor cell. A dip line is inserted to the bottom of the core for fuel filling and draining. The heat exchanger and pump are removed or inserted from above. These major components can be changed without breaking liquid seals, ‘since their closures are made in a gas volume above the fuel level. The fuel container is surrounded by a jacket of stainless steel, which serves as a container for forced-circulation gas preheating and cooling. The cooling unit was added so that afterheat could be removed, in an emergency, without draining the system. The package comprising the gas blower, heater, and cooler is arranged to be removed vertically from above, as in the case of the heat exchanger and fuel pump. The entire assembly is housed in a two-layer, gas-buffered steel enclosure, which is surrounded on the sides by 8 ft of concrete shielding. The heat exchanger and pump project upward through S5 ft of concrete so that semidirect maintenance operations may be performed in a shielded area. A multilayer organic-cooled boron steel shield surrounds the reactor to reduce the radiation level in the reactor cell. The coolant salt, which forms the intermediate coupling between the fuel and steam systems, is directed into the heat exchanger through a central pipe. This fluid flows on the outside of the tubes and out of the heat exchanger through the annular pipe. An elevation drawing of the plant is shown in Fig. 3, and two plan views at elevations 818.0 and 846 ft are shown in Figs. 4 and 5, respec- tively. A perspective view of the building is shown in Fig. 6. The area immediately above the reactor cell is the base for all fuel system maintenance oper- ations. Contaminated parts may be removed from this maintenance bay to a hot storage cell, which is included in the building. The maintenance bay is shielded and is sealed off during reactor operation. UNCLASSIFIED ORNL-LR—DWG 43505 PUMP EXPANSION TANK é ey | OFF-GAS m BLOWER CHAMBERS —| / I~ N\ AB4 EXCHANGER L HEATER-COOLER v PACKAGE //SHIELD g ¢ . 3 X 0 N * 7 |k J g/ | GAS DUCT REACTOR o R i % P 4.)} AN X X AKX 28 2N X XOSONAAAAKR R K XK XX (WXXXXYYYWYXXXYXXXXXXX —\ Fig. 1. Reactor Assembly. & MAIN SUPPORT FLANGE HEAT EXCHANGER BARREL OFF-GAS CONNECTION\ FILL AND DRAIN CONNECTION— o o o o 0o 0 0 0o O N © 0o 0 0 o 0O/ Qo o o o o n_o © o TOP VIEW THROUGH EXPANSION TANK GAS SEAL WELD JUNCTURE ETAIL MAIN SUPPORT FLANGE DETAIL FILL AND DRAIN LINE\\/ EXPANSION JOINTS T\% 1 UNCLASSIFIED ORNL—-LR—DWG 39229A PUMP BARREL PUMP VOLUTE EXPANSION TANK |_— GAS JACKET Fig. 2. Fuel Container Weldment and Gas Jacket. \\PUMP DISCHARGE || — HEAT EXCHANGER HOUSING //—FUEL RETURN ANNULUS ISOMETRIC OF EXPANSION TANK, BOTTOM VIEW CORE VESSEL UNCLASSIFIED ORNL-LR-DWG 37815A EL. 878.0 ft COOLANT PUMP SUPERHEATER (COOLANT TO STEAM) EL. 849.0 ft FUEL PUMP BLOWER 838.0ft SLAL EL. 834.5 ft OF ~GAS_ “HOLDUP * - TANKS Y HEATER AND COOLER DEAERATOR HOT CELL . 'A.} REACTOR ", NE-GENERA ¢ EL.803.0 ft v FEED-WATER HEATE FEED-WATER CIRCULATING AND CONDENSER PUMPS Fig. 3. Power Plant Elevation. 9 ¥ HOT CELL < A _—REAC .Y v. " TOR q e ) A ol T = o -, -.v‘ " HOLDUP TANKS -+ ELEVATOR— CHARCOAL TRAPS v . uP GAS HEATER, COOLER, AND BLOWER':’ > bl 0] mMEN D D ==l JANITOR'S CLOSET UNCLASSIFIED ORNL—LR-—DWG 37813A L FEVAPORATORS _—TURBINE-GENERATOR Do e A." oo b.'. 2 OFF-GAS v Rl - = - 4 I O ] c \STEAM PUMP I I -] 3 O.Q ' FEED-WATER HEATERS O J1 l FUEL REMOVAL AREA < 1 ...|q' oo . L JLTe e "".". R ‘\ 1 Q DEAERATOR Fig. 4. Power Plant Plan ot Elevation 818 ft. 124 ft—0in. 88 ft—O0 in. STACK LOBBY OFFICES e |°] Ll 1] ,\:DN“/A ELEVATOR ST OFFICES UNCLASSIFIED ORNL-LR-DWG 37812A ACCESS TO REACTOR GAS HEATER, ! COOLER,AND BLOWER CHARCOAL TRAPS (BELOW) S Fooodl [Fang Tl o w FUEL DRAIN VALVES (BELOW) @ CONTROL ROOM L. CELL COOLER |; (BELOW) o fiFu—:ACToR'Q r) % i i SAMPLER, ENRlCHER HEATER (BELOW) '\ f. COOLER (BELOW) st e Eoa VR "l SUPERHEATER UPPER PART OF TURBINE ROOM 4~ COOLANT PUMP 1 COOLANT DRAIN TANK [ e e 1 Caccess To EQUIPMENT H [r REMOVAL TUNNEL CHANGE ROOM e » Figo 5. PLAN—EL. 846.0 Power Plant Plan at Elevation 846 ft. b UNCLASSIFIED ORNL-LR-DWG 37810A R i e R il ""““I' | |"||I|”|I||n i i ot d \ b AR RN AR} ¢ S I EETNRR ‘\'.‘:_\\"'\:{tj\\‘\“{ Y y \\\'\\:{"n ,‘."’\1"'5‘."~~ I “\un.\ o Suvh A L R » H\‘}'I ',€ . SN o AN B VWA, “?\\‘(\-‘}? N\\\!’l\‘&‘;‘.;:}:_" — | P - \ oAy llu}m"‘l)ft‘.“ : pt A s S, o S, IR T Iy " 'Il’//”r, Vit (] n iy, ! " vy ‘/' h » [ Yy Y R ey h """" ’1 Astne u';‘ v o : “":’;” Yoo ’(1',"|‘ , Yy \ ' fy . . MR et e e - Sl Vet N ,.v-" o Fig. 6. Power Plant Perspective Looking Southwest. The intermediate-coolant pump and the steam superheater are housed in a separate shielded area which is accessible from above. Direct maintenance is planned for this portion of the system. As previously mentioned, the induced radioactivity in the coolant salt would decay rapidly, and direct maintenance could be performed after a few minutes’ delay to allow the 1l-sec fluorine activity to decay. The Loeffler system evaporators and steam pump are located in the turbine-generator bay. The bridge crane, in the turbine-generator bay, is located to service the Loeffler components, feed- water heaters, deaerator, and turbine-generator system. The control room is located between the reactor area and the turbine bay. Office and laboratory space is located to one side of the plant and extends over four floors. An extended work area surrounds the reactor cell on three sides at the 834.5-ft elevation. A site was chosen along the Clinch River at Gallaher Bend in the Oak Ridge area. The site plot is shown in Fig. 7. The plant is located on sloping terrain, which would minimize the ex- cavation requirements, since the general arrange- ment placed the reactor cell well above the condenser to prevent the possibility of flooding. This layout minimizes cooling water pumping requirements, but increases the building costs. This relationship has not been optimized, and the arrangement presented was selected as a reasonable first approximation. MOL TEN-SALT SYSTEM AUXILIARIES Enriching and Sampling System Fissionable material is to be added to the circulating fuel on a semicontinuous basis to sustain criticality, within design temperature limits, and fuel samples are to be withdrawn throughout the course of operation for chemical assay. A relatively small volume of fissionable material will be added at any one time, and a comparable, or smaller, volume will be withdrawn as a sample. A single mechanism is provided to accomplish both these operations. The fuel will be added as solid U235F4 at a free liquid surface in the fuel system expansion tank or in a separate vessel in which a free surface is presented. Sampling will be accomplished by reversing this operation. 10 That is, a small portion of the circulating fuel will be removed by ‘‘thief sampling,’’ allowed to solidify, and then placed in a shielded container for transfer out of the area. It is estimated that 70 kg of would have to be added to the system during the first two years of power operation (30 Mw at 100% plant factor) to take care of burnup and fission product over- ride. This quantity is equivalent to a daily addition of 96 g of U?33. Considering the density of UF, to be 6.70 g/cm®, it may be seen that approximately 14.3 cm? of UF, would have to be added daily. The frequency of additions would be contingent on the allowable mean temperature deviation in the fuel system. The quantity of fissionable material, AM, consumed per day was calculated, as mentioned above, to be 96 g. Without fuel additions, this would result in some mean temper- ature decrease of the fuel, AT . These quantities can be related as follows: U235 AM AT = —— m aBM / where a is the temperature coefficient of reactivity per °F, B is the mass reactivity coefficient, and M is the critical inventory in grams. For a temper- ature coefficient of —6.75 x 10~>, an assumed mass reactivity coefficient of 9, and a critical inventory of 90 kg, a mean temperature drop of less than 2°F per day would be experienced. The system could be operated at full power for several days before the decrease in fuel temperature would seriously affect the thermodynamics of the power cycle. Several experiments have been run in which slugs of UF, weighing approximately 40 g have been inserted into 1200°F LiF-BeF, salt to study solution rates and homogeneity of the UF, in the melt.® The UF, slugs were held in per- forated copper tubes, and the tubes were dipped into the salt mixture. The results of these tests indicate that solid UF, dissolves quite rapidly. The chemical assays of the resultant mixture gave excellent material balances. The fuel sampling and enriching device is shown schematically in Fig. 8. The entire Zzgm Quar. Prog. Rep. Apr. 30, 1959, ORNL-2723, p . UNCLASSIFIED ORNL-LR-DWG 37811A MAX. EXPECTED FLOOD LEVEL WITH DAM 745 ft msl NSNS NN LATITUDE I | ©0 0 l o 0 M 16 32 64 128 FEET _ o w > i 3 z b= - m . A B L S m ~ O 2 o 3 ~ M | © I o % © Fig. 7. Site Plot for Molten-Salt-Fueled Power Plant. UNCLASSIFIED ORNL-LR-DWG 35667 MOTORIZED WORM-GEAR DRIVES MOTORIZED WORM-GEAR DRIVES ENRICGHING-CAPSULE ELEVATOR SAMPLING-POT ELEVATOR | I t | / SAMPLE-DEPOSITING ELEVATOR—~ HORIZONTAL CONVEYOR SAMPLING CAPSULES ON CONVEYOR CANNED, REMOTELY CONTROLL: ED BALL VALVES ' - GANNED, REMOTELY CON- ROLLED BALL VALVES IELDED SAMPLE % CARRIER Auermioly ENRICHER-MAGAZINE PRESS N CANNED, REMOTELY CoN- A T s ENRICHING-CAPSULE TROLLED BALLVALVE /£ i~ o) | ~ (- w | =z . - / i ] 'l . ,L"Tm lm““",lf%m P ” // - N s T SAMPLE-CARRIER PRESS — GANNED, REMOTELY CONTROLLED - GATE VALVE 8 ~ REACTOR CELL TOP SHIELD SAMPLER AGCESS AREA Fig. 8. Fuel Sampling and Enriching Mechanism. 12 " v) ) mechanism is enclosed in a vacuum-tight struc- ture. All the material transfers in and out of the - reactor system are made at ‘‘air lock’’ sections which can be purged and isolated from the process and cell side of the equipment. This equipment would be maintained by the semidirect type of techniques employed on the other major com- ponents of the fuel system. The mechanism consists of a sample carrier elevator, enricher elevator, horizontal conveyer, and reactor sample enricher elevator. The ele- vator sections transport the capsules between their respective containers and the horizontal conveyer section which interconnects the three vertical sections. Actuation of these elevator elements is accomplished by motor-driven gear packages. Isolation of the various sections of the mecha- nism is accomplished by remotely operated ball and gate gas valves. Gas connections made between pairs of valves allow contaminated gas to be purged from the connections before dis- assembly and inert gas to be charged into the system after a connection has been made. All operation must be remotely controlled and will be interlocked to prevent improper manipulation of the equipment. Preliminary layout drawings have been made of this mechanism, and details for the gas valve operators and capsule handling have been developed. Fill and Drain System A fuel fill and drain system having a storage capacity of approximately 400 3 is provided. This system is made up of four vertical cylinders 2.5 ft in diameter and 20 ft high. The drain vessel size was selected to give adequate heat transfer surface for afterheat removal in a subcritical geometry. These vessels are manifolded in two pairs so that a spare fuel system inventory of salt can be made available in the plant. This excess salt can be used for system flushing during startup operations or system cleanup during any phase of the operation. Fill and drain lines extend to the bottom of each vessel and are coupled to the fill and drain line in the reactor vessel. Fuel is transferred between the reactor and drain tanks by a pressure-siphon principle. Each pair of drain vessels is mani- folded through two mechanical valves in series. A gas-connected surge chamber is located between each pair of valves. Fluid is transferred from the drain vessels to the reactor or vice versa by applying differential gas pressure to establish flow. During this operation the mechanical valves are open and the gas vent to the surge chamber is closed. When this gas vent is opened and pressures in the reactor and drain gas systems are equalized, flow is stopped and further fluid transfer is impossible. The mechanical valves may then be closed. They would not normally be exposed to liquid. The surge chamber also provides a source of buffer gas between the mechanical valves when the fluid is in the drain system and the reactor vessel must be opened for maintenance. The preheating and afterheat removal schemes are discussed in the next section of this report, and the fuel temperature rise from afterheat production for various values of heat loss are plotted against time after shutdown in Fig. 9. The data presented are based on one year of operation at full power; no credit was taken for fission-gas removal. It was assumed that the fuel temperature rise would be limited to 1500°F and that a heat sink of 300 kw would be adequate to maintain the temperature below this maximum, even with an excursion of 100°F resulting from fuel flow stoppage. The effects of a flow stoppage incident are discussed in the section on ‘‘Reactor Haz- ards.”’ A preliminary check of the criticality problem in the drain vessels showed the system to be safe. One hundred kilograms of U3 was con- sidered to be contained in a single vessel and UNCLASSIFIED ORNL-LR-DWG 43506 — 500 "O% ~ L Q ° > L7 < \,OV w % w2 A 400 X W @ < \O(y & 300 -~ 055 T—— : ABE o - '%J 200 7 ~ i ‘ L = 400 v 300 kw HEAT Loss 0 0 i 2 3 4 5 6 7 8 9 10 TIME AFTER SHUTDOWN (hr) Fige 9. Fuel Temperature Rise from Afterheat Pro- duction. Running time, one year; volume of salt, 175 ft3; heat capacity of salt, 64 Btu-ft_so(oF)-l. 13 was presumed to be settling out of the carrier salt as an oxide compound, so that all the fissionable material would be contained in the lower part of a vessel and the top surface would be reflected by the carrier salt. When the u23° was contained in the lower half of the vessel, a 10-ft-high by 2.5-ft-dia volume, the multiplication constant was 0.596. When the U%3° was con- sidered to be contained in the bottom as a cube 2.5 ft on a side, the multiplication constant in- creased to 0.827. This value was considered uncomfortably high; however, the drain-vessel diameter could be reduced at the bottom to offset this rise in multiplication. Fuel carrier and spent fuel salt would be loaded into and out of the plant via the drain vessels at a fuel transfer area adjacent to the drain vessels. Remote liquid connections, probably flanges, would be provided in the lines used to transfer radioactive material out of the plant. |t was presumed that freeze-flange junctions could be used in this application. These transfers would be made infrequently, and the transfer equipment would not be used for long periods. Estimates have been made of the shielding required for carrier flasks containing 2 f13 of fuel. The radiation level for various shield thicknesses is shown in Fig. 10. This calculation was made on the basis of one year of operation exposure at full power with 2.32 days of cooldown and no self-shielding in the fuel. Preheating and Afterheat Removal Equipment The liquidus temperatures of the fuel and coolant salts are above 800°F; therefore, means are provided for preheating all the process equip- ment containing these materials. Those components which contain the fuel salt are preheated, in the main, by forced gas circu- lation. This scheme eases the remote-maintenance problem in that a single, replaceable gas heating and blower package may be used to heat a large portion of the fuel system. The main preheating system for the reactor is shown in Fig. 1. This equipment includes a blower, a heater, and a cooler section. The cooler section was added to take care of emergency situations when it would be desirable to remove fission product decay heat without transferring the fuel to the drain tanks. UNCLASSIFIED 3 ORNL-LR-DWG 43507 10 — DOSE RATE (rads/hr S, 6 8 10 12 14 16 18 SHIELD THICKNESS (in.) Fige 10. Dose Rate from Carrier Flask. Fuel cyl- inder, 13,5 in. ID and 24 in. high. The heater-cooler package, shown schematically in Fig. 11, is connected to a thermally insulated stainless steel jacket which surrounds the entire reactor assembly. A centrifugal blower capable of delivering 3000 cfm with a pressure differential of 5 in. of water at a gas temperature of 1200°F recirculates gas through the heater-cooler package and then around the reactor assembly. One hundred eighty kilowatts of heater capacity is designed into the heater sections in the form of high-temperature (1500°F) alloy sheath tubular elements. A water-cooled fin-and-tube heat ex- changer capable of handling 250 kw is used as a heat dump. Hydraulically operated baffles are manipulated to direct the gas through either the heater or cooler unit. A preliminary estimate indicates that a cold reactor system can be preheated to 1000°F in less than two days by using forced air circulation. Resistance-type heater elements are provided in the upper section of the fuel heat exchanger. . A graph of the spectral distribution of fission density is presented in Fig. 16. It may be seen that 55% of the fissions are epithermal. Operation at power will require an increase in the fuel loading to override the effects of the accumulation of fission products and nonfis- sionable isotopes of uranium. Inventories and neutron balances for the initial state and after operation for two and five years at 30 Mw(th) are given in Table 4. The sum of the inventory 4/5 T x fissions x 10° per cm® per fission in core UNCLASSIFIED ORNL-LR-DWG 43508 . S T~~~ 0 1 2 3 4 5 6 7 (x109) RADIUS CUBED (cm3) Fig. 15. Relative Distribution in 30-Mw Experimental Molten-Salt-Fueled Reactor, Fission UNCLASSIFIED ORNL-LR-DWG 43509 20 c>5 18 % ' 2 5 " N £ 5 o \E LLSL.L ° y% ) 5 N \\w\‘ £ \\\ “\i ) crlniintnmy 4 0 2 4 6 8 10 12 1 LETHARGY o ® Fig. 16. Spectral Distribution of Fissions in 30-Mw Experimental Molten-Salt-Fueled Reactor, 21 ¢c Table 4. Nuclear Performance of Spherical Molten-Salt-Fueled Reactor Core diameter: 6 ft Power: 30 Mw(th) Plant factor: 1.0 Fuel volume: 171 §#° Fuel geometry: Spherical Fuel processing rate: 0 times per year A. lInventories and Neutron Balances After Cumulative Power Generation of 60 Mw-years After Cumulative Power Generation of 150 Mw-years Initial State During Two Years During Five Years Inventory Absorption of Operation of Operation (kg) Ratio Inventory Absorption Inventory Absorption (kg) Ratio (kg) Ratio Fissionable isotopes u23s 65.8 0.513 107 0.522 226 0.554 Py239 0.50 0.009 1.6 0.011 Fertile isotopes 4.9 0.0 9.0 0.018 19.1 0.029 Fuel carrier Be’ 912 0.009 912 0.008 912 0.007 Li’ 1200 1200 1200 19 0.099 0.069 0.045 F 7124 7124 7124 Fission Products 22.7 0.035 56.7 0.053 Parasitic isotopes (U236, etc.) 6.5 0.012 18.5 0.024 Miscellaneous: core vessel and leakage 0.368 0.327 0.277 B. Performance Data After Cumulative Power After Cumulative Power Generation of 60 Mw-years Generation of 150 Mw-years Initial State . . . During Two Years During Five Years of Operation of Operation Neutron yield, n 1.95 1.88 1.77 Total fuel inventory, kg 65.8 107 228 Cumulative net burnup, kg 0 28.7 73.3 Net fuel requirement of U235, kg 65.8 136 301 U235 amounts to 70 kg increase and burnup of during two years of operation. In the initial state, the neutron leakage from the spherical thermal shield amounts to about 108 neutronsecm™2.sec™! at a power level of 30 Mw(th). The spectral distribution of these neutrons is shown graphically in Fig. 17. UNCLASSIFIED 6 ORNL-LR-DWG 43510 24 22 20 ////V//%Z} Z llrZ4 Ll ALL 0l L Ll d L/ //I/ LA LN ) 77 PER CENT OF NEUTRONS PER UNIT LETHARGY ] 0000 N &\\\ AT T e 0 2 4 6 8 10 12 14 6 LETHARGY N WLLL LA A A A @ 0 Fig. 17. Spectral Distribution of Neutrons Leaking from Thermal Shield of 30-Mw Experimental Molten- Salt-Fueled Reactor, The gamma-ray heating in the core vessel was estimated by means of the Oracle code GHIMSR to be 2.5 w/cm®, and the strength of the gamma- ray current at the surface of the core vessel was estimated to be 5 w/cm?. The estimated spectral distribution is shown graphically in Fig. 18. The attenuation of this gamma-ray current in the thermal shield is substantial. It was estimated that the photon current leaking from the thermal shield would not exceed 0.001 w/ecm?. The estimated spectral distribution is given in Fig. 19. A breakdown of the gamma-ray escape-current energy in terms of gamma-ray sources is given in UNCLASSIFIED ORNL-LR-DWG 4354t FRACTION OF GAMMA ENERGY PER Mev O 1 2 3 4 5 6 7 8 9 10 1" {2 GAMMA -RAY ENERGY (Mev) Fig. 18. Spectral Distribution of Gamma Energy Leaking Through Core Vessel in 30-Mw Molten-Salt- Fueled Reactor. Table 5. |t may be seen that the major source of gamma rays is inelastic scattering of neutrons by fluorine. These gamma rays are, however, relatively low in energy. The fission and fission- product-decay gamma rays will probably provide the major biological shielding problem. REACTOR HAZARDS Exhaustive studies of the effects of nuclear transients induced by off-design operation or malfunctioning and breakdown of equipment of molten-salt-fueled reactors of this general design have not yet been made but must, of course, be completed before a design is approved for con- struction. Exploratory studies were made of a generally similar reactor of 600-Mw(th) cc:pc:cit‘y,4 and it was tentatively concluded that the reactor 4H. G. MacPherson et al., Molten-Salt Reactor Program Status Report, ORNL-2634, p 42-59 (Nov. 12, 1958), 23 is inherently stable and that a minimum of safety control equipment would be needed. The failure which led to the most serious transient was that resulting from a sudden stoppage of fuel flow, and therefore such a failure has been investigated in relation to the design proposed here. UNCLASSIFIED ORNL-LR-DWG 43542 FRACTION OF GAMMA ENERGY PER Mev O 4+ 2 3 4 5 6 7 8 9 10 1 12 13 44 GAMMA-RAY ENERGY (Mev) Fig. 19. Spectral Distribution of Gamma Energy Leaking from Thermal Shield of 30-Mw Experimental Molten-Salt-Fueled Reactor. The flow stoppage incident is defined as the one in which fuel circulation instantaneously ceases at a time prior to which the reactor was operating at design power. Practically, because of inertial and thermal convection effects, this precise incident is impossible to achieve, but, in the limit, it approximates a pump failure. The peak temperatures computed for such an incident are upper bounds of those which would occur as a result of a pump failure. During power operation of a circulating-fuel reactor some delayed-neutron precursors decay from the fuel while it circulates through the regions external to the core. Thus the fuel leaving the core is richer in delayed-neutron precursors than is the entering fuel. The steady- state precursor concentration in the core of the reactor when the fuel is circulating is accordingly less than when the fuel is not circulating. In order to compensate for the partial loss of delayed neutrons arising from the precursor deficiency, the reactor is just critical when k& exceeds 1 by the appropriate amount. When circulation is suddenly stopped, the precursor concentration gradually rises to the steady-state value appropriate to a stationary-fuel reactor (i.e., # = 1), and the corresponding in- crease in delayed-neutron production tends to make the reactor supercritical. Simultaneously, cessation of heat removal by circulation of the fuel tends to cause the core temperature to rise, and the reactor, because of its negative temper- ature coefficient, tends toward subcritical. The result is a complex nuclear power transient and a corresponding temperature transient which, in the present reactor, after a moderate overshoot Table 5. Components of the Gamma-Ray Current from the Core Vessel of a 30-Mw Experimental Molten-Salt-Fueled Reactor Fraction of Energy Mean Energy of Photons Gamma-Ray Source of Gamma-Ray Current (Mev) Fission 0.245 1.1 Fission product decay 0.191 0.9 Inelastic scattering of neutrons 0.487 0.2 in fluorine Captures in beryllium 0.007 5.6 Captures in fluorine 0.031 5.8 Captures in uranium 0.037 0.3 24 > finally settles down to a higher critical temper- ature. The above-described transient was studied for the present reactor with the aid of the ORNL Analog Computer Facility, using methods de- scribed by C. S. Walker.® The pertinent pa- rameters that were used to describe the reactor and the computed effect on the critical temperature are given below: Fuel u233 Spherical core volume 113.2 13 External fuel volume 57.5 £+ Fuel circulation rate 3.18 cfs Volumetric heat capacity of fuel 64 Btu'f'#—a-(oF)*] Design power (initial condition) 30 Mw Nuclear temperature coefficient ~6.75 X 1073 per °F of reactivity Prompt-neutron lifetime 8.8 x 107 sec Rise in critical temperature 19.5°F (computed) As expected, for the case in which no heat escapes from the core, the temperature rises asymptotically to a maximum, where it remains indefinitely, maintaining the reactor in a sub- critical condition. The temperature rises about 35°F in the first 10 sec, and the total rise in temperature is less than 110°F; 90% of the maximum is reached 90 sec after the start of the transient. Nuclear afterheat was not taken into account in the above calculations, and, of course, in any real reactor the temperature would not be asymptotic but would continue to rise because of afterheat. To limit the temperature during a prolonged stoppage of fuel circulation, afterheat must be removed by some alternative means. In general, if the removal rate is a fraction of the initial reactor power and exceeds the afterheat production rate, instantaneous stoppage of circulation will result in an initial temperature rise followed by a fall. After perhaps oscillating a few times, the temperature will settle on the new critical value appropriate to the noncirculating core. This 3. s. Walker, Simulation of the ORSORT Buttermilk Reactor,) Loss of Fuel Flow, ORNL CF-58-7-64 (July 31, 1958 stagnant-mean core-critical temperature is 19.8°F above the circulating-mean core-critical temper- ature. To gain insight into the maximum temperatures which might be attained and the times to reach them, the behavior of the reactor was investigated under a condition of auxiliary cooling wherein the cooling rate less the afterheat rate was held constant and the fuel circulation was instan- taneously stopped. The peak temperatures and times to reach them as a function of the assumed power-removal rate (in excess of that required for afterheat), which is constant throughout the transient, are indicated in Fig. 20. As expected, the peak temperatures and times to reach them decrease with increasing power-removal rates. UNCLASSIFIED ORNL—-LR—DWG 43513 120 \ ‘ l 120 ® TIME TO REACH \ \,/PEAK TEMPERATURE (sec) 100 ‘ 100 3 | Q = \ PEAK MEAN CORE _ @ ° - A °F & 5 50 3 TEMPERATURE RISECF) | g ¢ < n & \ x x = 60 N, 60 2 < = a > '\ = S 40 a0 & O F_. LJ = F_. O 10 15 CONSTANT POWER REMOVAL (Mw) (IN EXCESS OF THAT REQUIRED FOR REMOVAL OF AFTERHEAT) Fig. 20 Effect of Instantaneous Reduction of Heat Flow from Reactor on Temperatures. The initial rate of rise, which is important for thermal shock considerations, is very modest, as shown by the curve giving the rise in the first 10-sec period. Oscillations about the new critical temperature are well damped; the damping in- creases and the times between peaks decrease with increasing power-removal rates. Typically, for 3 Mw heat removal, the first maximum occurs in 82 sec following initiation of the transient, with a 55°F peak (above the new critical temper- ature); the second maximum occurs 9 min after 25 initiation of the transient, with a 5°F peak. It is clear that, if need be, ample time is available to adjust heat-removal rates to near the afterheat production in order to maintain the reactor above the critical temperature and, hence, subcritical without oscillations, so long as afterheat is sufficient to maintain control. COSTS An estimate of construction costs has been prepared and is presented in Table 6. It was concluded that the plant could be built for about $16,000,000. Primary emphasis was placed on obtaining estimates of the cost of the INOR-8 Table 6. Capital Cost Summary FPC Account No. 310 Land and land rights 311 Structures and improvements Site improvement Site facilities Station buildings Reactor building Turbine-generator building Total structures and improvements No cost $ 10,000 160,000 600,000 800,000 $1,570,000 312 Reactor and steam-generating equipment Primary system Fuel container and gas shroud Pumps and pump drive Primary heat exchanger Subtotal Primary system auxiliaries Fuel drain and storage Enriching and sampling system Purge system Off-gas and effluent system Inert-gas system Other auxiliary systems Subtotal Intermediate system Pump and pump drive Superheater Drain system Piping and valves Emergency dump system Other intermediate system equipment Subtotal Reactor cell: shielding and containment 275,000 250,000 150,000 $ 675,000 218,000 100,000 50,000 150,000 50,000 50,000 $ 618,000 150,000 194,000 50,000 100,000 75,000 50,000 $ 619,000 1,150,000 Heating, cooliing, and ventilating systems Reactor primary heating and cooling system Intermediate heating system Cell cooling and ventilating system Stack Subtotal 26 50,000 50,000 100,000 15,000 $ 215,000 312 314 315 316 Table 6 (continued) Reactor system instrumentation and controls Steam generator and feed-water system Loeffler evaporators Loeffler steam pump Feed-water heaters Boiler feed-water pumps and piping Other equipment Subtotal Total reactor and steam-generating equipment Turbine-generator equipment Turbine-generator and accessories Condenser and circulating-water system River intake, weirs, and shore-line structures Other equipment Total turbine-generator equipment Accessory electrical equipment Miscellaneous power plant equipment Reactor maintenance equipment Cranes Manipulators Viewing equipment Cutting and welding Miscellaneous tools Subtotal Spare parts Pumps (molten salt) Heat exchangers Steam pump Miscellaneous parts Subtotal Original inventory of molten salt Total miscellaneous power plant equipment Transmission line Total direct construction Contingency General expense Engineering and design Interest during construction Total cost 750,000 200,000 50,000 50,000 50,000 50,000 400,000 800,000 175,000 200,000 125,000 75,000 150,000 50,000 75,000 100,000 450,000 300,000 150,000 20,000 100,000 570,000 825,000 4,427,000 1,300,000 200,000 1,845,000 50,000 3,000,000 1,400,000 1,500,000 900,000 $ 9,392,000 6,800,000 $16,192,000 components of the molten-salt systems. These items made up the more unique portions of the plant and, in an over-all analysis, were subject to the greatest estimating errors. Most of the design time was concentrated on the major components of the salt systems to facilitate cost estimating. Engineering layouts were prepared of the fuel container, heat ex- changer, and superheaters. These drawings were reviewed by the Y-12 shop personnel for fab- ricability and first estimates of manufacturing time. This group has had considerable experience in nickel-base alloy fabrication, and they have built reactor-grade components. In addition, several outside fabricators were asked to review the drawings from the standpoint of design and fabricability. Many constructive comments were obtained from these reviews, but no serious objection with regard to concept or design was raised which would invalidate the system or seriously change the manufacturing estimates. Some development work would be required in the tube fabrication of the fuel-to-coolant-salt bayonet heat exchanger, and test work would be required to check out some of the welding and brazing on the major INOR-8 components. The finished INOR-8 components in the plant were estimated to weigh approximately 50,000 Ib. Most of the material required to fabricate these components would be in the form of plate and tubing. Facilities for supplying INOR-8 stock are available, and a sizable inventory is currently on hand. The raw-material prices used were obtained from vendors’ estimates on sizes and quantities needed for the experimental reactor and from actual costs of material received for the molten-salt-fueled reactor development pro- gram. The reactor buildings and site were not suffi- ciently specified to get complete cost information. A site was chosen near an area that had been studied for a similar reactor installation, and estimates for site improvement and facilities were available. No cost was applied to land acqui- sition, since the reactor was assumed to be built in the Oak Ridge area. The gross volume of the reactor building, offices, hot cells, laboratories, and control room was estimated at 200,000 3. The turbine- generator building was calculated to have a volume of 400,000 f+3. 28 The cost of most of the auxiliary systems was estimated without detail. These systems, or subsystems, were estimated as gross packages on the basis of general experience. One exception was the reactor heater-cooler unit, which was developed to the point of engineering layout and specification of the major components. ‘The more conventional portions of the steam generator plant were determined from manufac- turers’ data. Included in the first plant cost was the molten- salt inventory. This quantity included a spare fuel volume and 50% overage for the coolant-salt volume, resulting in a total of 550 ft3. This entire quantity was assumed to contain Li’ at the 99.99% assay level, which contributes one- third of the total cost of $1500 per cubic foot. An over-all contingency factor of approximately 30% of the first cost was used, and the other indirect costs were set at values considered applicable to this type of construction. The operating costs for the system were studied. After the completion of shakedown and planned experiments, it was concluded that the plant could be operated for $635,000 a year. The breakdown of this estimate is as follows: Wages (including supervision) $250,000 Supplies 10,000 Maintenance 75,000 Fuel burnup and inventory charge 270,000 Fuel preparation 30,000 $635,000 ALTERNATE DESIGNS A suggested modification of the fuel container assembly and reactor is shown in Fig. 21. The fuel pump is located on the vertical axis of the core and is a removable part of the heat exchanger assembly. The maintenance concept for this system is the same as that previously described, with the exception that the fuel pump must be removed when major heat exchanger maintenance is required. The pump can be removed or replaced independently. The sump-type centrifugal pump has an annular diffuser. The fluid fuel is directed out of the diffuser section into the tubes of the heat ex- changer. After passing through the heat ex- changer, the flow is directed into the annular COOLANT OUT l COOLANT IN UNCLASSIFIED ORNL—LR—DWG 43514 i PUMP MOTOR OFF-GAS LIN | FILL AND DRAIN LINE M _GAS-DUCT OUTLET X XSO KNS 'y = R/ EXPANSION /l SPACE d PUMP —— M EXCHANGER 1 I REACTOR / W)&X XXX XAy GAS-DUCT INLET Fig. 21. ConcentricsPump Molten-Salt-Fueled Reactor Layout. 29 volume surrounding the reactor and then dis- charged up into the core at the bottom opening. The fuel leaves the core at the top and goes vertically up into the pump section. The fuel expansion tank surrounds the pump barrel and is cooled by the intermediate salt circuit. The coolant salt enters the heat exchanger shell at the bottom and flows upward around the tubes. After it leaves the heat exchanger it passes around the pump region and out of the assembly. This arrangement has several important advan- tages. The primary fuel container geometry is simplified, and the support problem is more straightforward. The fuel inventory is reduced, and the expansion tank region is completely surrounded by coolant to ensure more positive temperature control of this region. The major disadvantage of the concept is the more complicated heat exchanger upper structure. This region, as presently conceived, includes the stationary parts of the pump assembly, which would be discarded in the event of a heat ex- changer failure. Even so, this concept does render more of the fuel system components readily replaceable and would simplify the operations required to remove the entire fuel container from the reactor cell. This fuel system is sufficiently attractive to warrant further design study. An alternate steam-generating concept has also been considered for molten-salt-fueled reactor systems., There appears to be little question as to the operability of the Loeffler cycle, but the system does result in a cost penalty because of the massive evaporation drums required in larger power plants. The proposed steam generator, shown in Fig. 22, would be used in place of the superheater, steam pump, and evaporator drums. The molten-salt coolant is circulated on the shell side of the bayonet tubes. Feed water is recirculated through the bayonet-tube assemblies and superheated and/or saturated steam is with- drawn from the steam generator. The feed water is pumped up into the tubes through the inner annulus and spills back down through the central tube to the sump. The water boils as it is pumped up this annulus by virtue of the heat transferred from the molten salt through the steam as it flows down the outer annulus. The system is regenerative in that the steam formed is used as a heat transfer fluid coupling the molten salt and the boiling water. This 30 system, therefore, allows low-temperature water, 600°F or less, to be put into the same vessel with the salt having a liquidus temperature of 865°F. The steam output is regulated by reducing the flow through the steam annulus so that less heat is transferred to the boiling water and the load is reduced. Conversely, when more flow is channeled through the steam annulus, the load is increased. A heat baffle 51 in. long was added at the outlet end of the steam annulus to insulate the steam from the water so that 1000°F superheat could be obtained. Saturated steam would be removed from the region above the sump, and this flow would be blended with the superheated steam to regulate the steam temperature. At lower flows, the outlet steam is at a higher temperature and a larger fraction of saturated steam is used for tempering. The design data for a 30-Mw steam generator are presented in Table 7. This unit is considered to be more easily fabricated than the U-tube, U-shell steam superheater, and experiments are planned to study the control and heat transfer performance of this geometry. Table 7. Design Data for Once-Through Steam Generator Coolant salt In 65 psia, 1160°F Out 37 psia, 1000°F Flow 2.17 cfs Feed-water inlet conditions 96,400 Ib/hr at 475°F, 1600 psia 96,400 Ib/hr at 592°F 192,800 Ib/hr at 535.5°F 96,400 Ib/hr at 1000°F Recirculated feed water Total feed water Steam outlet conditions Configuration Concentric bayonet tubes in cylindrical shell Flow Steam in tubes Tubes Size 1.25-in. OD, 0.109-in.- wall Number 91 Pitch 19/]6 in., triangular Length 27 £+ 9 in. ) UNCLASSIFIED ORNL~-LR—~DWG 39168A N THERMAL BAFFLE \ \\ § WATER \ § = SUPERHEATED N\ | STEAM D \ N\ \ L WATER AND STEAM SECTION THROUGH TUBE ASSEMBLY SUPERHEATED STEAM OUT —=—o = il 2e il ¢ SATURATED WATER OUT Fige 22. Once-Through Steam Generator, 31 It was previously mentioned that unclad graphite was being investigated for use in molten-salt breeder reactors. Graphite test specimens could be inserted in the core of this reactor without undue complication, or the entire core could be modified and a graphite moderator incorporated. Neither of these approaches would alter the basic fuel container concept, and the latter would result in a system with a much smaller liquid inventory and probably a smaller core vessel. LIST OF DRAWINGS A list of ORNL drawings that were prepared for this design study is presented below: Title Plant Section Looking West Plant Section Looking North Plan - Elevation 838 Plan - Elevation 803 Reactor and Cell Fuel Container Weldment Fuel-to-Coolant-Salt Heat Exchanger Steam Superheater Bayonet Tube Boiler-Superheater Reactor Heater-Cooler Assembly Concentric Reactor Assembly Enricher Sampler Assembly Enricher Sampler Assembly Preliminary Site Perspective, S. W, Corner Perspective, S. E. Corner Perspective, N. E. Corner 32 ORNL Drawing No. F-2-02-054-7942 F-2-02-054-7943 F-2-02-054-7938 F-2-02-054-7941 F-2-02-054-7939 F-2-02-054-7940 F-2-02-054-7972 F-2-02-054-7897 F-2-02-054-7928 F-2-02-054-9057 F-2-02-054-9028 F-2-02-054-7657 F-2-02-054-7658 F-2-02-054-7944 D-2-02-054-7945 D-2-02-054-7946 D-2-02-054-7947 ] is PPENO N AN~ . Affel . Alexander . Bettis . Billington . Blankenship . Blizard . Boch . Borkowski . Boudreau . Boyd . Bredig Breeding . Briggs . Browning . Campbell . Carr . |. Cathers . E. Center (K-25) . Charpie . Coobs . Culler . DeVan . Douglas . Emlet (K-25) . Ergen . Estabrook . Ferguson . Fraas . Franco-Ferreira . Frye, Jr, . Gall . Gresky Gregg . Grimes . Guth . S. Harrill . R. Hill . W. Hoffman . Hollaender . S. Householder . H. Jordan . W. Keilholtz . P. Keim . 1. Kelley . Kertesz TOMWP>PMTACT TV OO MEOQEPPIZTOME-PE-MPUSEMD-T-F00EDEAIMZOEOPMIMOME™D AT 4TI >omMmXXxXo>ITrI>» . B. W. Kinyon INTERNAL DISTRIBUTION 49. 50. 51. 52. 53. 54. 55. 56. 57. 58. 59. 60. 61. 62. 63. 64. 63. 66. 67. 68. 69. 70. 71. 72. 73. 74. 75. 76. 77. /8. 79. 80. 81. 82. 83. 84. 85. 86. 87. 88. 89. 90. 91. 92. 93. 94. @EFPOTMOAMPE-OMP>IMIE->IZ-V0OPVEOZF-AAMAL-IETrMERLIAN-F ORNL-2796 UC-81 Reactors -~ Power TID-4500 (15th ed.) . Lackey Lane . Livingston . MacPherson . MacPherson . Manly . Mann . Mann . McDonald . McDuffie . McNally Metz . Milford . Miller . Miller . Morgan . Murray (Y-12) . Nelson . Nessle . Osborn V-r INZTEODCTTMEP>OOMOY>mM . Patriarca . M. Perry . Phillips . M. Reyling T. Roberts T. Robinson . W. Savage . W. Savolainen L. Scott . E. Seagren . D. Shipley . J. Skinner . H. Snell . Storto . D. Susano A. Swartout . Taboada . Taylor . Thoma . Trauger . Yonderlage . Watson . Weinberg . Whatley . White . Whitman oOOmzT=TNwmI 33 34 95. G. C. Williams 102-121. Laboratory Records Department 96. C. E. Winters 122. Laboratory Records, ORNL R.C. 97. J. Zasler 123-124. Central Research Library 98-101. ORNL — Y-12 Technical Library, Document Reference Section EXTERNAL DISTRIBUTION 125. D. H. Groelsema, AEC, Washington 126. Division of Research and Development, AEC, ORO 127-710. Given distribution as shown in TID-4500 (15th ed.) under Reactors — Power category (75 copies — OTYS) | )