; : : ; - LOFT ORNL P 2534 ? . t an Things . : : . is . : .... : i · . f . A $ . . ✓ . . 7. . SOHB LE EEEFEEEE . . . III L25 L6 || LẼ . A MICROCOPY RESOLUTION TEST CHART NATIONAL BUREAU OF STANDARDS - 1963 Onn --28344. NOV 2 9 1966 10. 12:00; 20 CONF-660934-2 MASTER Molten Salt Reactors* by J. A. Lane For presentation at the IAEA International Survey Course on Economic and Technical Aspects of Nuclear Power Vienna, Austria, Sept. 5-16, 1966 RELEASED FOR ANNOUNCEMENT . . ten .. ... . , .. - IN NUCLEAR SCIENCE AESTRACTS LEGAL NOTICE This report was prepared as an account of Goveroment sponsored work. Neither the United States, nor the Commission, nor any person aciing on behalf of the Commission: A. Makes any warranty or representation, expressed or implied, with respect to the accu- racy, completeness, or usefulness of the information contained in this report, or hat the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report, As used in the above, "person acting on behalf of the Commission" includes any em- ployee or contractor of the Commission, or employee of such contractor, to the extent that such employee or contractor of the Commission, or employee of such contractor prepare3, disseminates, or provides access to, any information pursuant to his employment or contract with the Commission, or his employment with such contractor, . AL . ti M . WS. *Research supported by the U.S. Atomic Energy Commission under contract with the Union Carbide Corporation. Hl.. . BLANK PAGE partiti Introduction LT : ... .. . . .. Nuclear power based on current technology converter reactors seems to be an assured commercial success in many countries of the world. The competitive position of these reactors relative to conventional fuels, however, depends on the continued availability of low cost uranium ore. Although no one knows for certain how much cheap uranium exists or when it might become exhausted, most people agree that supplies are not suf- ficient to sustain present low prices for very long if one continues to build poor fuel utilization reactors. This circumstance has placed upon those countries committed to the large scale use of atomic energy, the burden of forestalling any serious rise in the cost of nuclear fuel. It is for this reason that the development of an ecoromical. breeder reactor, at one time viewed as a long range goal, has emerged as the central task of atomic energy developmenit. Moreover, as country after country commits itself to nuclear power, the stake in developing the breeder rises. As a matter of fact, the stakes are so enormous that no one should even admit the possibility that breeder development might fail. All plausible paths to a successful breeder, therefore, should be explored as extensively as possible, The popular approach to breeder development, of course, is via the fast reactor route which because of neutronic and other reasons has been deemed by most to be more promising than the thermal breeder route. The apparent reasons for this choice are (1) the fast breeder uses plutonium which is produced in converter reactors, (2) fast reactor fuels are not logy developed for converters can be applied to fast reactors, and (4) cancellation of the aqueous homogeneous project in 1959 made it appear that fluid fuel reactors might be too difficult to develop. While most of these reasons are still valid and appear to substantiate the choice of fast breeders as the main line approach, in view of developments which have taken place during the last several years it appears worth- while to re-examine the question of fast versus thermal breeders. In this regard two factors have considerably altered the situation that existed formerly. These are (1) the development of high performance economically competitive fast. breeders appears to be considerably more difficult than originally supposed and (2) the successful operation of the molten salt reactor experiment (MSRE) has demonstrated that it is possible to build and operate a fluid fuel reactor. This essentially removes the last and main reason for objecting to thermal breeders. The Incentives for Developing a Fluid Fuel Breeder The fact that it is possible to build and operate a fluid fuel re- actor is not in itself sufficient justification for the development of such a reactor. To be successful a breeder must meet three requirements. First, the breeder must be technically feasible. Second, the cost of power from the breeder must be as low or lower than that from competing systems; and third, the breeder must be able to meet all of the fissile material needs of a full fledged nuclear economy so that mining of high - 2. meet these criteria as well as, and in some respects be other reactor system. Moreover, since the technology of molten salt breeders in no way overlaps the technology of the solid-fueled fast reactor, its development provides the world with a completely inde- pendent alternate pate to long term cheap nuclear energy that is un- affected by any obstacles that may crop up in the developmer. of the fast breeder. The molten salt reactor though seemingly a newcomer to the reactor business, in fact represeats the culmination or more than 16 years of research and development. The incentive to develop a reactor based on fluid fuels has been strong ever since the early days of the Oak Ridge National Laboratory; however, it was not until 1959 that the molten salt reactor emerged as the most promising of the fluid fuel approaches. Up to that time three prominent fluid fuel reactors were being developed. These were the liquid bismuth, the aqueous homogeneous, and the molten salt. In 1959 the USAEC assembled a task force to evaluate these con- cepts. The principal conclusion of their report was that "the molten salt reactor has the highest probability of achieving technical feasi- pility". This verdict, however, was not surprising in view of the ex- pected thermal, cherical, and nuclear advantages of the fluoride salt fuel syytem. The more important of these advantages are (1) complete thermodynamic stability of fuel salt and container system, (2) absence of radiation damage to fuel, (3) low vapor pressure of salt, (4) inert in contact with air, (5) continuous removal of neutron poisons from fuel, tinuous removal of fissile product from blanket, (7) high neutron econoray, (8) low fissile inventory, (9) low doubling time, aná (10) high thermal efficiency. To us who have followed the molten salt project over the years, it is obvious that a reactor concept possessing such advantages should be thoroughly exploited. The Molten Salt Reactor Experiment The present status of molten salt reactor technology is best described in terms of the MSRE ană related information. In the MSRE, the fuel, instead of being the usual uranium dioxide pellets encased in metal tubes, is in the form of UF, dissolved in a mixture of lithium fluoride, beryllium fluoride, and zirconium fluoride. While these salts are solid at room temperature, they melt at 425°C to 510°C and are nearly as fluid as water at their cperating temperature of 650°C to 700°C. This fuel mixture is circulated by a pump through holes in the graphite core and delivers its energy through a heat exchanger located outside of the core. The obvious simplicity that this provides for the reactor core and the greater ease associated with the handling of fluids instead of solids have always had some attraction (particularly to chemical engineers) but, until now, there has been an aura of mystery and fear associated with pumping and circulating the highly radioactive fuel solution. The MSRE was built for the purpose of proving that this problem is amenable to solution by demonstrating the operability of salt 2 "Report of Fluid Fuel Task Force," TID-8507, U.S. Atomic Energy Commis- sion, February 1959. . . . ** - 3 . reactors and at the same time investigating the compatibility of fuels and materials in a high radiation field. The design conditions are shown in the flow diagram in Figure 1, and the general arrangedient of the plant is shown in Figure 2. The fuel for the MSRE 16 65 mole of L?F-29.1% BeF2-50% ZrF:-0.9% UF4. Except for a small amount 01 Zrk and a higher UF4 concentration, it is the fuel for the core of the reference breeder, În the reactor primary system the fuel salt is recirculated by a sump-type centrifugal pump through a shell and U-tube heat exchanger and the reactor vessel. The flow rate is about 80 liters/sec; fuel enters the reactor at 625°C and leaves at 663°C. The base pressure in the system is 1.3 atm in the helium cover gas over the free surface of salt in the pump bowl. The maximum pressure is about 4.7 atm at the outlet of the pump. The heat generated in the fuel salt as it passes through the re- actor vessel is transferred in the heat exchanger to a molten-salt coolant containing 66% LiF and 3406 BeF2. The coolant is circulated by means of a second sump-type pump at a rate of 5? 1/s through the heat exchanger, entering at 552°C and leaving at 593°C, arid thi'ough a radiator where the heat is dissipated to the atmosphere. The base pressure in this system is also 1.3 atm in the pump tank; the maximum pressure, at the discharge of the pump, is 5.7 atm. Drain tanks are provided for storing the fuel and the coolant salts at high temperature when the reactor is not operating. The salts drain from the primary and secondary systems by gravity. They are transferred between tanks or returned to the circulating systems by pressurizing the äraill tanks with helium. The fission product gases krypton and xenon are removed continu- ously from the circulating fuel salt by spraying salt at a rate of 3.2 1/s into the cover gas above the liquid level in the fuel pump tank. There they transfer from the liquid to the gas phase and are swept out of the tank by a small purge of helium. After a about 1-1/2 hr in the piping, this gas passes through water-cooled beds of activated carbon. The krypton and xenon are delayed until to the atmosphere. Fuel and coolant systems are provided with equipment for taking samples of the molten salt through pipes attached to the pump tanks while the reactor is operating at power. The fuel sampler is also used for adding small amounts of fuel to the reactor while at power to compensate for burnup. Finally, the plant is provided with a simple processing facility for treating fu-l batches of fuel salt with hydrogen fluoride and fluo- rine gases. The hydrogen fluoride treatment is for removing oxide con- tamination from the salt as H20. The fluorine treatment is the fluoride volatility process for removing the uranium as UFG. The equipment - 4. ... ORNL-LA-DWG 59158A COOLANT PUMP FUEL PUMP 552°C 1025°F HEAT EXCHANGES 52 liter/sec .830 gpm 663°C 1225°F 625°C 1175°F 1250 | 1100°F gom ī 393°C 80 litert, soci COOLANT SALT I LiF-66% | BeF - 34% REACTOR VESSEL 300°F 150°C REACTOR CELL AIR 167,000 cini 100°F 79 m3/sec 38°C DRAIN TANK CELL | FREEZE T VALVE T 'SPARE FILL AND DRAIN TANK (6811) 1.9 ms FILL AND DRAIN TANK (68 113) 1.9 m3 FLUSH TANK (68 773) 1.9 m3 COOLANT DRAIN TANK (4011) 1.1 m3 Fig. 1. MSRE Flow Diagram. omnitong 3-4201 REMOTE MANTENANCE CONTROL ROOM REACIOR CONTROL ROOM, . سدانش 2 نن .... 1. . I REACTOR VESSEL 2 HEAT EXCHANGER. à FUEL PUMP 4. FREEZE FLANGE 5 THERMAL SHOLD 6. COOLANT PUMP O 7. RADIATOR 8. COOLANT ORAIN TANK 9. FANS 12 ORAN TANKS IL FUSH TAMK 12 CONTARMENT VESSEL 13. FREEZE VALVE Fig. 2 General Arrangement or MSRE. - 6 - approaches the size required for batchwise processing of the blanket of the 1000-Mw(e) reference reactor. All the equipment in the MSRE that contains salt, is made of Hastelloy n. All of it was designed to be able to operate at 700°C. Because the liquidus temperature of fuel and coolant salts is near 430°C, it is desirable to keep the salts molten in the reactor systems and in the drain tanks. To do this the major pieces of equipment are installed in electric furnaces and the piping is covered by electrical heaters and insulation. The reactor primary system, the fuel drain tank system, and some auxiliaries become permanently radioactive during the first few hours of operation at appreciable power. Maintenance of this equipment and associated heaters, insulation, and services must be done remotely or semi-remotely by means of special tools. Tools have been developed for accomplishing this maintenance for the MSRE equipment. The MSRE reactor vessel is shown in Figure 3. It is about 1.5 m diameter by about 2.6 meters high from the drain line at the bottom to the center of the outlet nozzle. The wall thickness of the cylindrical section is 1.4 cm; the top and bottom heads are 2.8 cm thick. The core contains approximately 600 vertical graphite bars 5 cm square x 1.7 m j.ong. Most of the bars have grooves deep machined along the full length of each face and adjacent bars are aligned to form channels 3 cm x 1 cm for the salt to flow through the core. The graphite is a new type with high strength, high density, and because the salt does not wet the graphite it cannot penetrate through the small openings into the pores unless the pressure is raised to 5 to 20 times the normal pressure in the core. The MSRE fuel and coolant systems were first heated for the pre- nuclear testing in the fall of 1964. The reactor was first critical in June 1965 and reached its maximum power of about 8 Mw'(th) in June 1966. The accumulated operating experience through July 10, 1966, is presented in Table 1. Table 1 Accumulated Operating Experience with MSRE 2730 4660 Fuel System Circulating helium above 537°C, hr Circulating salt above 537°C, hr Full thermal cycles, 38°C to 648°C Coolant System Circulating helium above 537°C, hr Circulating salt above 537°C, hr Full thermal cycles, 38°C, tó 648°C Time Critical, hr Integrated Power, Mwhir Thermal 1970 5200 1600 6800 7. "It'. "..... . 7 OM UW G10010 FLEXIBLE SONDUIT TO CONTROL ROD DRIVES SAMPLE ACCESS PORT — COOLING AIR LINES ACCES8 PORT COOLING JACKETS FUEL OUTLET REACTOR ACCESS PORT CONTROL ROD THIMOLES CORE CENTERING ORIDA LOW DIGT.RIBUTOR BV WWWUWW AAVANAVAV GRAPHITE-MODERATOR STRINOER FUEL INLETS REACTOR CORE CAN a ::* REACTOR VESSEL T ANTI-SWIRI, VANESS VESSEL DRAIN LINE MODERATOR SUPPORT GRID Fig. 3. Reactor Vessel. :- :. - 8 . In most respects the reactor has performed exceptionally well. Analyses for corrosion products in the salt indicate that there has been essentially no corrosion of the Hastelloy N by the salt. Inspec- tion of some parts of the fuel system confirmed that the corrosion was negligible after about 1890 hr of circulating salt in prenuclear and critical tests. Analyses of the fuel salt for uranium and reactivity balances indicate that the fuel has been completely stable. Although there have been numerous problems with auxiliaries and electrical sys- tems, few problems have been encountered with the major reactor systems. The time to reach full power was extended several months by problems with the housing of the radiator in the coolant circuit and with plugging of small lines in the off-gas system that handles the helium and gaseous fission products from the pump bowl. The radiator housing is a large pusulated, electrically heated box around the radiator coils and is required so the radiator can be kept hot and the salt in it mo.lten when the reactor is not producing fission heat. The difficulties were in obtaining proper operation of the doors and in controlling leakage of hot air through joints and through ducts for electrical leads to prever.t overheating of equipment outside the housing. Future molten-salt reactors are unlikely to have similar radiators, but the experience will be helpful in designing better fur- naces for other equipment. The off-gas system was designed for a small flow of gas, essentially free of solid or liquid aerosols. Some difficulty was experienced with micron-size particles of salt collecting in the tiny ports of the flow control valves, but much more difficulty was experienced after the re- actor began to operate at I Mw with organic solids and viscous organic liquids collecting in the valves and at the entrance to the carbon beds. Investigation of this problem is incomplete and the observed phenomena cannot be explained in detail, but the general mechanisms seem clear. The bearings on the fuel circulation pump are lubricated and parts of the pump are cooled by oil. The oil is separated from the pump tank by a rotary seal. Provision is made for directing the normal seal leak- age of 1 to 10 cc per day of oil to a waste tank and preventing liquid or vapor from coming in contact with the salt or cover gas in the pump tank. Under special conditions, recently demonstrated in a pump test loop, this oil can leak through a gasketed seal in the pump presently in the MSRE into the pump tank where it vaporizes and the vapors mix with the helium purge stream and flow into the off-gas system. The oil has no effect on the fuel salt, but the organic materieis polymerize in the off-gas system under the intense beta radiation cé the gaseous fission products to form the visccus liquids and solids that plugged the valves. This problem has been reduced to a ninor nuisance in the MSRE by installing absolute filters and a carbon bed for trapping solids and heavy liquids ahead of the control valves. The leakage path has been eliminated in future pumps by substituting a welded seal for the gasketed seal. Small amounts of organic and inorganic vapors or aerosols are - 9- likely to be found in the off-gas from future reactors, but they can be easily controlled by the use of filters, traps, and absorbers. The maximum power reached in the MSRE is 20 to % below the design power. It is limited by the heat transfer performance of the radiator, primarily the air-side coefficient, but the overall heat transfer co- efficient of the primary heat exchanger is also less than had been calculated. Better values for the properties of the salts and better relationships for calculating heat transfer coefficients are needed, but the deficiencies can easily be overcome in the future by making the heat transfer equipment larger. 1000 MWE Molten Salt Thermal Breeder As the next major step in the development of molten salt breeders, it seems reasonable to build an actual two region prototype breeder. To provide the basis for the design of such a prototype, the reference design of a 1000 MWE thermal breeder power plant (MSBR) and of some alternatives as improvements to the reference design have been developed by the Oak Ridge National Laboratory. Details of this study are re- ported in ORNL-3996, the results of which are summarized here. Reactor Design In contrast to light water reactors which have solid fuel elements and liquid moderator', the MSBR uses liquid fuel and solid moderator ele- ments. A cluster of such elements called fuel cells is shown in Figure 4. As seen in this figure, the fuel salt circulates upward through outer holes in the graphite and downward through a larger central passage in each cell. The entire core consists of 534 of these cells stacked together inside a 3 m x 3.8 m reactor vessel shown in Figure 5. Fuel circulates from the outlet plenum through the pumps to the heat ex- changers and then back to the reactor. A 46 cm thick blanket and an 8 cm thick graphite reflector surround the core. The thorium salt circulates through the blanket region, through the passages between fuel cells in the core, and through the heat removal system outside the reactor' vessel. The heat generated in the fuel and blanket salts is transferred from the primary salt system to a secondary coolant salt system shown in Figure 6. This secondary coolant system which generates supercritical steam acts as a barrier between the highly radioactive fuel salt and the power generating equipment. The secondary system also serves to minimize the size and fuel inventory requirement of the primary fuel salt system. A plan view of the MSBR is shown in Figure 7. On two sides of the reactor cell are four shielded cells containing the boiler-superheaters and the reheaters; those cells can be isolated individually for main- tenance. A cell for handling the gaseous fission products from the re- actor and two cells for processing the fuel and blanket salts are adjacent to the reactor cell. Cells are also provided for decontamination and storage and repair of radioactive equipment. - 10 - ORNL-OWO 66: 4796 BL.ANKET PASSAGE MODERATOR (GRAPHITE) FUEL PASSAGE (UP) FUEL PASSAGE (DOWN) -3%-in. OD FUEL TUBE PITCH MODERATOR HOLD DOWN NUT (GRAPHITE) hom Bi gali aan 1 ammy REACTOR CORE Sääväkäntara - SPACER - METAL TO GRAPHITE SLIP-JOINT . METAL TO GRAPHITE BRAZED JOINT - ---BRAZED JOINT ALL FUEL INLET PLENUM LLEC: .Luci FUEL INLET PLENUM - FUEL OUTLET PLENUM Fig. 4. Cross Section of a Fuel Cell. - ll - ORNL DWG. 66-6968 FUEL PUMP MOTOR CONTROL ROD | DRIVE BLANKET PUMP MOTOR CONSTANT SUPPORT HANGERS Z : - - BLANKET HEAT EXCH FUEL DUMP TANK WITH COOLAG COILS FOR AFTER KATI REMOVAL - Asad 42.0 O FT. DIA CORE REACTOR VESSEL, - FUEL SALT DISTRIBUTION PLENUMS REACTOR PEDESTAL. 1 19'6" araniitit PRIMARY HEAT EXCH C!!!;!;!;- L-REACTOR CELL HEATERS Fig. 5. Reactor Primary Equipment. 1000 MWE MSBR - 12 - ORNL DWG. 66-7110 CONTROL ROD DRIVE COOLANT SALT PUMPS FUEL CIRCULATING PUMP BLANKET CIRCULATING PUMP tehnikort SUPERHEATERS GROUND LEVEL REACTUH 1:48'0" FUEL HEAT EXCH. LBLANKET HEAT EXCH. LREHEATERS Fig. 6. Molten Salt Breeder Reactor - Cell Arrangement, Elevation. ORIL DWG. 66-7111 . REHEAT STEAM H.P. STEAM FEEDWATER WASTE GAS CELL FEEDWATER -HP. STEAM -LP. STEAM 28 COOLANT SALT PUMPS P FUEL HEAT EXCH PREACTOR U V2173 . 13 U SA 12" : 8 REHEATERS 28511 min. X DECONTAMINATION ANO STORAGE L-16 SUPERHEATERS : L BLAMET LHEAT Exor AWIT CONTROL AREA" Fig. 7. 'Molten Salt Breeder Reactor - Cell Arrangement, Plan View. RAT .. . .. . . 2 - 14 - The reactor vessel and all other equipment that holds salt is made of Hastelloy N, a nickel-base alloy containing about 17% molybdenum, 7% chromium, and 40 iron. This material is highly resistant to corrosion by fluoride salts and has good strength at high temperature. The high- temperature creep properties of Hastelloy N presently obtainable com- mercially deteriorate under irradiation, but small changes in the alloy offer promise of eliminating this deficiency. The graphite is a high-density grade processed to achieve small pore openings for low permeability to salt. Superior resistance to damage by irradiation is important, but the core is designed to keep the flux gradients small across individual pieces and to permit the graphite to expand or contract with little restraint. Values chosen for some of the MSBR design parameters are listed in Table 2. Table 2 Values of Some MSBR Design Parameters Power, Mw Thermal Electrical 2225 1000 0.45 0.80 Thermal Efficiency Plant Factor 3.81 3.05 Dimensions, meters Core height Core diameter Blanket thickness Radial Axial Reflector thickness 0.46 0.61 0.07 Volumes, liters Core Blanket: 27,800 31,100 Salt Compositions, mole % Fuel LiF BeF2 UF4 (fissile) Fertile LiF BeF2 63.6 36.2 0.22 71.0 2.0 27.0 0.0005 ThF2 UF4 (fissile) Core atom ratios Th/U C/U 41.7 5800 . - - 15 - Table 2 (continued) 769 260 Fissile inventory, kg Fertile inventory, thousands of kilograms Processing by fluoride volatility and vacuum distillation Cycle time, days Rate, ft3/day Fuel Stream Fertile Stream 23 47 14.5 144 Fuel and Blanket Processing The primary objectives of the processing are to separate fission products in low concentration from the other constitué salt and to separate bred fissile material in low concentration from the other constituents of the blanket salt while keeping the losses and the costs low. With the fluoride fuel and blanket salts of the MSBR, these objectives can be fulfilled by a combination of fluoride volatility and vacuum distillation processes. The processing is done continuously or semi-continuously in cell space and adjacent to the reactor; services and some other equipment required for the reactor are shared by the processing plant. Shipping, long storage at the reactor and reprocessing sites, and refabrication of fuel and blanket are eliminated. All these factors lead to reduced inventories, im- proved fuel utilization, and reduced costs. The fuel salt is conveniently processed by fluoride volatility to remove the uranium and by vacuum distillation to separate the carrier salts frrin the fission products. Blanket processing is accomplished by fluoride volatility alone. The effluent UFG is absorbed by purified fuel salt and reduced to UF4 by treatment with hydrogen to reconstitute a mixture of the desired composition. Principal steps in the processes are shown in Figure 8. Small streams of core and blanket fluids are withdrawn continuously from the reactor and circulated through the processing system. After processing, the decontaminated fluids are returned to the reactor at convenient points such as the storage tanks. Inventories in the processing plant are estimated to be about 10% of the reactor fuel system inventory and less than 1% of the blanket inventory MSBR Capital Costs Preliminary estimates of the capital cost of a 1000-Mw(e) MSBR power station indicate a direct construction cost of about $80 million. After applying the indirect, cost factors used in the advanced converter evaluation (ORNL-3686), the estimated total plant cost is $114 million. A summary of plant costs is given in Table 3. The conceptual design was not sufficiently detailed to permit a completely reliable estimate; however, the design and estimates were studied thoroughly enough to make meaningful comparisons with previous converter-reactor plant cost ORAL-DEG es-ergoa UF, RECYCLE TO REACTOR SORDERS 110717 SORDERS NOFT 100-400°C / WASTE WASTE STORAGE Kof/MOF/FP) PE EXCESS 2 PROOUCTION . UFG+ VOLATILE FP WASTE STORAGE 2 Maf/Mgfx/FP MAKE UP LiF/8e/Thr, UF + VOLATILE FP MAKE UP LIF/Dofz - 16 - FERTILE MAKE UP I CONTINUOUS FLORIDE VOLATILITY Z HOLDUP/ FOR YFP DECAY SPENT PUEL LiF/Befrei CONTINUOUS FLORIDE Vaawa DISTILLATE I VOLATILITY DISTILLATION Lif/borz PZI-5500 -1000®CV 500°C UFUFIon / UFU REDUCTION INFILTRATION $50-500°C -550°C Lif/BE TW Lif/BeFz/thf, /FP Lif + RARE EARTH FP -Hz REDUCED METALS C.FN DISCARD FOR FP REMOVAL WASTE STORAGE FEKTILE STREAM RECYCLE Lif / Bofy/UF, RECYCLE Fig. 8. MSBR Core and Blanket Processing Scheme. PM - 17 - studies. The relatively low capital cost results from the small physi- cal size of the MSBR and the simple control requirements. The results of the study encourage the belief that the cost of an MSBR power station will be as low as for stations utilizing other reactor concepts. The operating and maintenance costs of the MSBR were not estimated Based on the ground rules used in ORNL-3686, these costs would be abouc 0.3 mill/Kwhr(e). Table 3 Preliminary Cost-Estimate Summary for a 1000-Mw(e) MSBR Power Station Costs (thousands of dollars) Federal Power Commission Account Land and land rights 21. Structures and improvements Reactor plant equipment Turbine-generator units 24 Accessory electrical 25 Miscellaneous Total direct construction cost (excludes account 20) Total indirect costs (includes account 20) Total plant cost 360 9,340 44,850 22,530 2,900 800 80,420 33,200 113,620 Fuel Recycle Plant The capital costs of the fuel recycle plant were obtained by itemizing and costing the major process equipment and by estimating the costs of site, buildings, and instrumentation, waste disposal, and building services associated with fuel recycle. Including a 25% contingency, total costs are estimated at $5.3 million. The operating and maintenance costs for the fuel recycle facility include labor, labor overhead, chemicals, utilities, and maintenance materials. The total annual cost for a capacity of 425 d of fuel salt per day and 3000 e of blanket salt per day is estimated to be about $720,000, which is equivalent to about 0.1 mill/kwhr(e). Nuclear Performance and Power Costs The fuel cycle cost and the fuel yield are closely related, yet independent in the sense that two nuclear designs can but significantly different yields. The objective of the nuclear design calculations was primarily to find the conditions that gave the lowest fuel cycle cost, and then, without appreciably increasing this cost, the highest fuel yield. • 18 - The calculations were performed with OPTIMERC, a combination of an optimization code with the MERC multigroup, diffusion, equilibrium reactor code. The program MERC calculates the nuclear performance, the equilibrium concentrations of the various nuclides, including fission products, and the fuel cycle cost for a given set of conditions. OPTIME.IO permits up to twenty reactor parameters to be varied, within limits, in order to determine an optimum, by the method of steepest ascent. The designs were optimized essentially for minimum fuel cycle cost, with lesser weight given to maximizing the annual fuel yield. Typical para- meters varied were the reactor dimensions, blanket thickness, fractions of fuel and fertile salts in the core, and fuel and fertile stream proc- essing rates. The basic economic assumptions used in the calculations are given in Table 4, and the corresponding performance of the MSBR is given in Table 5. The total power cost of the reference design based on these data is calculated to be 2.7 mills/Kwhr. The distribution of costs is summarized in Table 6. Table 4 Basic Economic Assumptions 1000 45 0.80 Reactor power, Mw(e) Thermal efficiency, % Load Factor Cost assumptions Value of 233U and 233Pa, $/8 Value of 235U, $/8 Value of thorium, $/kg Value of carrier salt, $/kg Capital charge, annual rate, % Plant Nondepreciating cap..tal, including fissile inventory Processing cost, dollars per cubic foot of salt Fuel (at 283 /day) Blanket (at 2830 2/day.) Processing cost scale factor (exponent) 228 8.5 0.4 Table 5 MSBR Performance Fuel yield, myr 4.9 Breeding ratio 1.05 Doubling time, yr 20 Fissile losses in processing, atoms/fission absorption 0.006 Neutron production per fissile absorption, ne 2.22 Specific inventory, kilograms of fissile material 0.77 per megawatt (electrical) - 19 - Table 5 (continued) . . -* . *. . . . . . Specific power, megawatts (thermal) per kilogram of fissile material Power density, core average, kw/liter Gross In fuel salt . . ..... . . . . . " Table 6 Power Cost, mills/Kwhr(e) Capital costa 1.95 Operating and maintenance cost 0.30 Fuel cycle costº 0.45 Total 2.70 412% Fixed charge rate, 80% load factor, 1000-Mw(e) plant. Nominal' value used in advanced converter evaluation, ORNL-3686. Costs of on-site integrated processing plant are included in this value. Improvements and Alternatives to the Reference Design The MSBR reference design involves primarily a scaling up of MSRE design features and as such gives only a partial indication of the ultimate potential of molten salt breeders. A brief mention of possible alternatives to the reference design and the resulting im- provement in performance is therefore worthwhile. continuous Protactinium Removal - The most straightforward way to improve the fuel cycle performance of the MSBR is to incorporate in the blanket processing a method for removing 233 Pa continuously on a cycle of 10 to 60 hr. This would result in a higher breeding gain, lower inventories of fissile and fertile material, shorter doubling time, and lower fuel cycle and total power cost. A comparison of the performance of MSBR's with and without 233 Pa removal is shown in Table 7. Two methods for separating protactinium from molten fluoride salts have been tried in laboratory experiments with some success. In one, PaO2 was found to coprecipitate with ZrO2 from a molten fluoride salt to which solid ZrO2 was added. In a second, protac- tinium was extracted from a fluoride melt by contacting the melt with molten bismuth or lead in which thorium metal was dissolved to act as a reducing agent. - 20 - Further work seems certain to lead to an inorganic ion exchange or a liquid-metal extraction process that would remove protactiniun from the blanket salt of a breeder reactor continuously and inexpen- sively. A protactinium removal process should be developed and in- cluded in the first power breeder station. Table 7 Comparison of Performance of MSBR with and Without Continuous 233 Pa Removal Without Protactinium Removal With Protactinium Removal Breeding ratio Specific inventory, kg/MW(e) Specific power, Mw(th)/kg Fuel yield, % per annum Fuel cycle cost, mills/Kwhr Total power cost, mills/Kwhr 1.05 0.77 2.9 4.9 0.45 2.7 1.07 0.68 3.3 8.0 0.33 2.6 Modular Design - The reference design has four fuel circuits and four blanket circuits operating off one reactor vessel in order to produce 1000 Mw(e). One coolant circuit is provided for each fuel and blanket circuit. If one pump in the primary system were to stop or one tube in a primary heat exchanger were to fail, the entire plant would have to be shut down until the fault was repaired. We believe the components can be made reliable enough so that such shutdowns will be infrequent, but they will happen. As an alternative, a modular design was evolved. Each primary circuit and its secondary circuit was connected to a separate reactor vessel to provide four 555-Mw(th) reactor modules. The modules were installed in separate cells so that one could be repaired while the others were operating. The layout is shown in plan view in Figure 9. Although the modular design has four reactor vessels, they are about 80% as large as the reference vessel. The average power density in the fuel salt and the blanket thickness are the same as in the reference design. Al.. the rest of the equipment in the two types of plants is the same and the plants are of very nearly the same size. The total plant costs are found to be the same within the accuracy of the estimates; there is no significant difference in breeding performance. These results are preliminary; however, if they continue to hold true, it appears that because of its potentially higher plant avail- ability factor, the modular design would offer an advantage over a single core MSBR. E ORAL DHG. 66-7114 EEEE Ceee COM Eutelsa3). EEESB0300 FFFF39) STEAM GENERATORS CECHEF Boon porno REHEATERS COOLANT PUMPS T7.0" . ya PRIMARY HEAT- EXCHANGER AND FUEL PUMP - 21 - . ELANKET HEAT EXCHANGER AND PUMP THERMAL SHIELD more Y INSULATION E REACTOR . 11:0—- Fig. 9. Plan of Modular Units. - 22 - Direct-Contact Cooling with Molten Lead The reference-design MSBR has three volumes of fuel outside the core in heat exchangers, piping, plenum chambers, etc., for each volume of fuel in the core. Studies indicate that the fuel volume could be reduced to about one volume outside the core for each volume in the core if the fuel salt were circulated and cooled by direct contact with molten lead. The lead would be pumped into a jet at the lower end of each fuel tube. Salt and lead would mix in the jet and be separated at the outlet. The salt would return directly through the graphite cells to the core and the lead would be pumped either through intermediate heat exchangers or directly to the steam generators. This system has several advantages. Ideally, the specific inven- tory could be reduced to 0.3 to 0.6 kg of 235U per megawatt (electrical) and the doubling time to 5 or 6 years. Relatively inexpensive lead would be substituted for some of the lithium and beryllium fluorides. The lead pumps and heat exchangers could be arranged for maintenance of individual units with the remainder of the plant operating. Some parts of the plant should be considerably simplified. There are some uncertainties also. Thermodynamics data indicate that lead, fuel and blanket salts, graphite, and refractory metals such as niobium and molybdenum alloys should be compatible. Prelimi- nary tests indicate that this is true and that the much less expensive iron-chromium alloys might be used in the main lead systems. However, the materials problems are almost unexplored; little is known of the effects of radiation or fission products or of the ease of separating lead and salt. The lead-coo.led reactor represents an almost completely new technology that cannot presently be given a good evaluation. Work on the basic chemical, engineering, and materials problems of the system should be pursured aggressively to make a good evaluation possible within three or four years. If direct-contact cooling proves to be practical, its adoption could produce impressive improvements in the performance of the thermal breeders and could point the way to the use of molten-Salt fuels in fast breeders, .. Comparison of the MSBR with Other Advanced Reactor Concepts It is evident from the foregoing that the reference design MSBR still leaves a lot of room for improvement, however, even without such improvements, the MSBR compares favorably with all other advanced re- actor concepts, not only in terms of economics but also in terms of fuel utilization. No large molten-salt reactors or fast breeders and few Large advanced converters have been designed in complete detail, so most of the costs must be educated estimates based on comparisons of the reactor systems and judicious use of information from reactors that are being built. Such a comparison was made of several advanced converter reactors and reported in ORNL-3686. The results are summarized in Table 8. A comparable estimate of costs for a large molten-salt Table 8. A Comparison .of Estimated costs for Breeder and Advanced Converter Peastors Based on Lvestor-Owed Utilities Charges PAR SBR Advanced Converter Reactors I GA SGR U Tb WR HOCR Moltes sult Thermal Breeder Reactor WAP 3251-1 Fast Breeder Reactors CEND ACNP GEAP 200 64503 418 GA 5537 83 Cest icr 1000-Müle) power plant, $ millior. Direct costs Indirect costs Total capital Special fluids Fuel processing plant dhe po 133(e) 13 ngono AYOO 849 go mm 2.3 2.0 0.3 2.2 0.3 2.1 0.3 2.3 2.1 0.3 2.0 0.3 2.0 (2.3) (2.3) (2.3) 0.3 (2.3) (2.3) 0.3 0.16 0.26 0.de 0.346 0.39 0.40 0.02 0.150 0.31 0.02 0.13 0.0% 0.69 0.19 Forer costs (nills;kwhr) Fabrication Sinus and losses Frocessing Shirfire Inventory I3terest on vcrtine capital Subtotal Pu or 2330 credit Het cost Special fluids invertory and Capital cost Cperating cost Fiel cycie cost replacemert Total power cost 0.0 0.99 0.20 0.03 0.24 0.07 1.87 0.26 0.20 0.19 0.05 0.52 0.14 1.35 0 1.4 0.20 0.9? 0.19 0.02 0.27 0.02 1.67 0.14 1.5 0.05 2.30 0.24 2.1 0.31 0.81 0.23 0.03 0.20 0.03 1.51 0.35 1.2 0.670 0.22 0.41 0.22 0.39 c.0 1.3 0 1.3 - 23 - 0.52 0.51 0.25 0.39 0.17 0.14 0.02 0.04 0.09 0.30 0.06 0.07 1.1 1.48 0.250 0.9 1.5 0.16 0.01 1.18 0.05 0.330 0.11 0.89 0.42 0.5 0.33 0.17 0.66d 0.04 1.26 0.52 0.7 0.400 0.03 0.96 0.19 0.8 1.76 0.32 1.0 12.9) 2.8) (8.3 .1) (2.0) (ej Included because plant is similar to sodium-cooled fast breeder plants. (b) Although these numbers may be higher than present bid prices for large buclear power plants, the basis is the same as for the other converter reactors and for the MSBR and provides a fair comparisata, (c) Includes capital charge on processing piant. . (d) Adjusted to 10% charge for investor-owed utilities to be consistent with other studies. (e) Capital costs taken to be the same as PWR. (f) Fuel cycle cost is 30-year averaged cost. Ruel cycle cost for equilibrium breeder cycle is 2.4 mills/kwhr. - 24 - thermal breeder reactor, made by the same people and reported in ORNL- 3996, is also included in the table, along with the fuel cycle costs from several studies of fast breeder reactors. Capital costs were not estimated in the fast breeder studies; however, if one accepts, in the absence of estimates, that the costs for building and operating large power plants containing fast breeder reactors should not differ greatly from the costs for the other plants in Table 8, then differences in power costs depend primarily on differences in fuel cycle costs. Ac- cording to the numbers in the table, the fuel cycle costs and the total power costs for the fast breeder plants are mostly lower than for the converter plants but higher than for the molten-salt thermal breeder plant. It is seen from Table 8 that the overall cost of power from a privately owned, 1000-MW(e) molten-salt breeder reactor should come to around 2.6 mills Kwhr(e). In contrast to the fast breeder, the extremely low cost of the MSBR fuel cycle hardly depends upon sale of by-product fissile materiai. Rather, it depends upon certain ad- vances in the chemical processing of molten fluoride salts that have been demonstrated either in pilot plants or laboratories; fluoride volatility to recover uranium, vacuum distillation to rid the salt of fission products, and, with somewhat less assurance, removal of pro- tactinium by liquid-liquid extraction. The molten-Salt breeder, opera ting in the thermal Th-2350 cycle, is characterized by a low breeding ratio: the maximum breeding ratio consistent with low fuel cycle costs is estimated to be about 1.07. This low breeding ratio is compensated by the low specific inventory* of the MSBR. Whereas the specific inventory of the fast reactor ranges between 2.5 to 5 kg/Mw(e), the specific inventory of the molten-salt breeder ranges between. 0.4 to 1.5 kg/Mwle). The estimated doubling time for the MSBR therefore falls in the range of 8 to 50 years. This is comparable to estimates of doubling times of 7 to 30 years given in recent fast breeder reactor design studies. From the point of view of long-term conservation of resources, low specific inventory in itself confers an advantage upon the thermal breeder. If the amount of nuclear power grows linearly, the doubling time and the specific inventory enter symmetrically in determining the maximum amount of raw material that must be mined in order to inventory the whole nuclear system. Thus, low specific inventory is an essential criterion of merit for a breeder, and the detailed comparisons in a later paper show that a good thermal breeder with low specific inventory could, in spite of its low breeding gain, make better use of our nuclear resources than a good fast breeder with high specific inventory and high breeding gain. Total kilograms of fissionable material in the reactor, in storage, and in fuel reprocessing and refabrication plants per megawatt of electric generating capacity. R *.:. - 25. The molten salt approach to a breeder promises to satisfy the three criteria of technical feasibility, very low power cost, and good fuel utilization. Its development as a uniquely promising competitor to the fast breeder 16, we believe, in the national in- terest of every country. References . 1. Report of the Fluid Fuel Reactors Task Force, US-AEC Report TID- 8507 (February, 1959), . .. 2. M. W. Rosenthal et al., A Comparative Evaluation of Advanced con- verters, ORNL-3686 TJanuary 1965). 3. Paul R. Kasten, E. S. Bettis, R. C. Robertson, Design Studies of 1000-Mw(e) Molten-Salt Breeder Reactors, ORNL-3996 (in press). C. D. Scott and W. L. Carter, Preliminary Design Study of a con- tinuous Fluorination-Vacuum Distillation System for Regenerating Fuel and Fertile Streams in a Molten Salt Breeder Reactor, ORNL- 3791 (January 1966). . de i END DATE FILMED 12/28/ 66 . . he . ."