2 Ao bl e I.EGA}. N'oflcz' . nmor the Comminion, rior uny porson ucting on beha!f bf the Cummusion. HA. Makes - ‘any .warranty or repreunfoflon, Bxpress “or - implmf with rnpeci 4o tha accorccy, : R e : S L ST compleuness, or usefulnoss of. tlu Informaflon comulned hr-thll uporf o cht ,the use of" S T T : corrlrcctor of -the Commiuion to fhe a:fent thm s cH amployoe or conlracfor propares, l\andln i op distribufos, or provldes access io, cny infcrmafion pursuant to. hls ""‘PIOYmem o Comracg - Wm" fhe Commlssion. SR Contract No. W-Th05-eng-26 REACTOR PROJECTS DIVISION MOLTEN-SALIT REACTOR FPROGRAM STATUS REPORT H. G. MacFPherson, Program Director Date Issued DEC1 1308 Previously Issued May 1, 1958, as CF-58-5-3 OAK RIDGE NATIONAL LABORATORY Operated by UNION CARBIDE CORPORATTON - for the U.S. ATOMIC ENERGY COMMISSION ORNL-263L O ——— R S e by e ACKNOWLEDGMENTS The following people wrote portions of Part 1, "Interim Design of & ' Molten~-Salt Power Reactor": L. G. Alexender, B. W. Kinyon, M. E. lackey, H. G. MacPherson, L. A. Mann, J. T. Roberts, F. C. Vonderlage, G. D. Whitman, J. Zasler. In addition, the following people made major contributions to the reactor design and to the organization of Part 1: E. J. Breeding, W. G, Cobb, J. Y. Estabrook, F. E. Romie, C. F. Sales, D. S. Smith, J. J. Tudor. ‘The authors of Part 2, "Properties of Molten Fluorides as Reactor Fuels," vere W. R. Grimes, D. R. Cuneo, F. F. Blankenship, G. W. Keilholtz, H. F. Poppen- | diek M. T. Robinson. Substantial contributions were made by C. J. Barton, C. C. Beusman, W. E. Browning, S. Cantor, B. H. Clampitt, H. A. Friedman, H. W. Hoffman, H. Ifisley, S. Ianger, W. D. Manly, R. E. Moore, G. J. Nessle, R. F. Newton, J. H. Shaffer, G. P. Smith, N. V. Smith, R. A. Strehlow, C. D. Susano, R. E. Thoma, W. T. Ward, G. M. Watson, J. C. White. Part 3, "Construction Materials for Molten-Salt Reactors,” was contributed by W. D. Manly, J. W. Allen, W. H. Cook, J. H. DeVan, D. A. Douglas, H. Tnouye, D. H. Jansen, P. Patriarca, T. K. Roche, G. M. Slaughter, A. Taboads, G. M. Tolson. Part 4, "Nuclear Aspects of Molten-Salt Reactors," was written by L. G. Alexander, the work reported there was done in collsboration with J. T. Roberts. - Part 5, "Equipment for Molten-Salt Heat Transfer Systems," was vritten by H. W. Savage, W. F. Boudreau, E. J. Breeding, W. G. Cobb, W. B. MeDonald, H. J. Metz, E. Storto. Other major contributors to the work reported were R. G. Affel, J. C. Amos, J. A. Conlin, M. H. Cooper, J. L. Crowley, P. A. Gnadt, A. G. Grindell, R. E. MacFherson, W. R. Osborn, P. G. Smith, W. I. Snapp, | W. K. Stair, D. B. Trauger, H. C. Young. Part 6, "Buildup of Nuclear Poisons and Methods of Chemical Processing,"” "was vritten by J. T. Roberts, and contains contributions from G. I. Cathers, D. 0. Campbell, and the guthors of Part 2. The over-all editor of the report was A. W. Savolainen. The choice of material included was the responsibility of H. G. MacFherson. The authors wish to express appreciation to W. H. Jordan and A. M. Weinberg " for their helpful guidance of the molten salt work. gk wrt T O 0 - O . CONTENTS ACKNOWIEDGEMENTS 1. 2. » 9. 9 9 9 9. 9. 0. @ © 10.1 Fuel Salt Reprocessing PART 1. INTERIM DESIGN OF A POWER REACTOR Introduction and Conclusions General Features of the Reactor Molten-Salt Systems 3.1 Fuel Blanket Circults 3.2 Off-Gas System 3.3 Molten Salt Transfer Equipment 3.4 Heating Equipment 3.5 Auxiliary Cooling 3.6 Remote Maintenance 5.7 TFuel Fill-and-Drain Tank Beat Transfer Syeteme Turbine and Electric System Nuclear Performance Procedure for Plant Startup Reactor Control and Refueling Accidents: Consequences, Detection, and Reqnired’Action 9.1 An Instantaneous loss of load from a Secondary Sodium Circuit 9.2 An Instantaneous St0ppage of Sodium Flow 1in One of the Primary Heat Exchangers- - 9.3 An Instantaneous Reduction of the Heat Flow Rate from the Reactor Core 4 Cold Fuel Slugging .~ = | 5 Removel of Afterheat by Thermal Convection 6 loss of Fuel Pump 7 Ioss of Electric Transmission Line Connection . to the Plant 8'1’Ieak Between Fuel and Blanket Salts 9< 10 1 e leak Between Fuel and Sodium - . 1eak of Fuel or. Blanket Salt to Reactor Cell Ieaks of Water or Steam to Sodium'” 'Chemical Processing and Fuel Cycle Econcmics 10, e}fBlanket Selt Reprocessing _ -10.3 " Cost Bases . 10.4 Chemical Plant Capital COSts 10.5 Chemical Plant Operating Costs 10.6 Net Fuel Cycle Cost -iii_ 11 s VAN NO OV 18 22 2k 27 3k 3l 42 45 50 52 25 53 53 22 56 29 29 60 61 62 62 62 11, 12. . 12.1 ' Alternate Heat Trensfer Systens 1. o. 1. - Construction and Power Costs 11.1 Capital Costs 11.2 Power Costs fSome Alternates of the Proposed Design .12 2 Alternate Fuels 63 68 69 T2 PART 2. CHEMIGAL ASPECTS OF MOLTEN~FLUORIDE-SALT REACTOR FUELS Choice of Fuel ‘Composition 1 1 Choice of Active Fluoride Uranium Fluoride ' ' Thorium Fluoride 1.2 Choice of Fuel Diluents Systems -Containing UFL ' Systems Containing ThF), Systems Containing ThF) and UF) Systems Containing PuFz Purification of Fluoride Mixtures 2.1 Purification Equipment 2.2. Purification Processing Physical and Thermal Properties Radiation Stability Behavior of Fission Products 5.1 F1351on Products of Well-Defined Valence - The Noble Gases , 'Elements of Groups I-A, II-A, III-B, and IV-B .2 Fission Products of Uncertain Valence «3 Oxidizing Nature of the Fission Processes i\ PART 5. CONSTRUCTION MATERIALS FOR MOLTEN-SALT REACTORS Survey of Suitable Materials Corrosion of Nickel-Base Alloys by Molten Salts 2.1 Apparatus Used for Corrosion Tests 2.2 Mechanism of Corrosion Febrication of INOR-B 1 -Casting 2 Hot Forging 3 Cold Forming 4 Welding 5 3Brazing | 6 Nondestructive Testing - iv - s Th Th 75 75 75 82 90 % % 91 91 93 100 10k 104 104 106 108 108 109 112 112 125 123 125 - 125 125 127 132 'l o ~3 O W\ 10, 2. * Mechanical and Thermal Properties of INOR-8 h.1 Elasticity Plasticity - k.2 | t Z Aging Characteristics Thermal Conductivity and Coefficient of Iinear Thermal Expansion Oxidation Resistance Fabrication of a Duplex Tubing Heat Exchanger Ayailability of INOR-8 -Compatibility of Graphite with Molten Salts and Nickel-Base Alloys - - Materials for Valve Seats and Bearing Surfaces Sunma.xry ofematerial Problems PART 4. NUCIEAR ASPECTS OF MOLTEN-SALT REACTORS - Homogeneous Reactors Fueled with P 1.1 Initial States Critical Concentration, Mass, Inventory, and " Regeneration: Ratio 'Neutron Balances and Miscellaneous Details ' Effect of Substitution of Sodium for Ii7T Reactivity Coefficients Heat Release in Core Vessel and Blanket 1.2 Intermediate States = ‘Withoiit Reprocessing of Fuel Salt With Reprocessins of Fuel Salt Homogeneous Reactors Fueled with U233 2.1 Initial States 2. 2 Intermediate States Homogeneous Reactors Fueled with Plutonium 5 1 Initial States j?"" S ' » ,,jcritical Concentration, Mass, Inventory, and folRegeneration Ratio C , :Neutron Balance and Miscellaneous Details f 3 2 Intermediate States 7f3¢terogeneous Graphite-Mbderated Reactors b1 Inttiel States S fPART 5- EQUIPMENT FOR MOLTEN-SALT REACTOR I-IEAT TRANSFER SYBTEMS . '—Pmnps for Molten Salts 1.1 Improvements Desired for Power Reactor Fuel Pump 1.2 A Proposed Fuel Pump 13k 134 134 - 144 1l 1b7 153 153 159 159 . 162 163 168 179 179 179 18k 185 185 185 191 192 - 192 192 199 199 199 199 201 . 202 206 208 209 ON 1 B W - Heat Exchangers, fixpansion Tenks, and Drain Tanks 211 Valves 211 VSystem Heating | 213 Joints - 21k ' Instruments | 217 6.1 Flow Measurements . 217 6.2 Pregsure Measurements 217 6.3 - Temperature Measurements 219 6.4 Iiquid-Ievel Measurements 219 6.5 Nuclear Sensors ' 219 PART 6 BUILDUP OF NUCIEAR POISONS AND METHODS OF CHEMICAL PROCESSING JBui1dup of Even-Mass-NUMber Uranium TIsotopes 226 Protactinium and Neptunium Poisoning 023 QFission.Pr5duct Poisoning 22k . Corrosion Product Poisoning 226 * Methods for Chemical Processing 206 5.1 The Fluoride Volatility Process 228 - 5.2 K-25 Process for Reduction of UFg to UF) \ 251 5.3 Salt Recovery by Dissolution in Concentrated HF 2351 5.4 Rare Earth Removal by Exchange with Cerium 232 5.5 Radiocactive Waste Disposal - Vvi-- 236 STATUS OF MOLTEN-SALT REACTOR PROGRAM PART 1 INTERIM DESIGN OF A POWER REACTOR 1. INTRODUCTION AND CONCLUSIONS Thé general usefulness of & fluid fueled reactor that can operate at high temperatures with low pressures has teefi recognized for a long time. 'Therapplication'of the'molten salts to such a reactor system has been discussed,l and the operation of the Aircraft Reactor Ex.periment2 demonstrated the basic feasibility of the system. Preliminary design studies indicated that power reactors based on such systems would be economicaiiy attractive. This study gives.a more detailed conceptual ~design and outlines operational procedures so that the problems of handling a moiten-salt power reactor can be better visualized. Particular attention has been given to the circulating-fuel system, since this system and its associated equipment will be the heart of any ,molten-salt reactor plant. of perhaps lesser importance are the particular reactor chosen for study (a_tfidéregion'homogeneous converter) and the par- ticular heat transfer system (tfio sodium circuits in series). Alchough later studies may indicate better ‘choices for the reactor and the heat transfer system, those selected for this study are considered to be sound and- to provide 8 good baS1s for estimating the cost of power from & molten- : salt reactor._ - 13 Co Briant and.A. M4 Weinberg, "Molten Fluorides as Power Reactor | :Fuels,“ Nuclear Bcience and Eng. g 797-803 (1957) 2% 5. Bettis, B. W. ‘Schroeder; Gs A. Cristy, E. W. Savage, R..G. 'fAffel, ‘and L. F. Hemphill, "The Aircraft Reactor Experiment - Design and ‘Construction," Nuclear Science and Eng. 2, 804825 (1957); W. K. Ergen; A+ D, Callihan, C. B, Mills, and Dunlsp ScOtt,'“The Aircraft Reactor. ‘Experiment - Physics," Nuclear Science and Eng. 2, 826-840 (1957); E. ‘s. Bettis, W. B. Cottrell, E. R. Mann, J. L. Meem, &nd G. D. Whitman, “"The Alrcraft Reactor Experiment - Operation,” Nuclear Science and Eng. 2, 811-853 (1957). | = L. - The reactor power station chosen for study has a gross electrical capacity of 275 Mw and a net capacity of 260 Mw. Figure 1.1l shows an isometric drawing of the principal portion of reactor plant, and the most important of the reactor"statistics-arg'p:esented in Table 1l.1l. It is estimated that this molten-salt reactor power station could be built for 70 million dollars. At 4% per year interest and an 80% load factor, the fiXed chargesj'including'fueleinventory rental, would amount 0 5.7 mills/kwh. Fuel and salt replacement costs of 1.7 fiills/kwh and an operation and.malntenance charge of 1.5 mllls/kmh (1ncluding chemical plant operation) lead to a total estimated pover cost of 8.9 mills/kwh. ‘The indicated power ‘cost must be considered-together with the state of the technology of mblten'salts,‘of alloys for containingrthém, and of engineering art for de31gn and construction of & reactor 1n order to determine the emphasis that should be placed on gtudies of the system in the future. Summaries of the current state of the technology of the galts, metals, and components are'given in other parts of this report. The fact of adequate solubility of uranium and thorium in the molten salts and the strong position that is developing with respect to con- tainment of the salts are characteristics that make the molfen_salt 8YyS= tem unique among fluid-fuel systems. Although the materials studies are not complete, the early results are so encoursging that plans should be made now for the continued development of the molten-galt system. The program visualized calls for carrying out the conceptual design of an expefimental reactor during the fiscal year 1959 so that detailed design could be started by July 1, 1959. The experimental reactor would be designed to test typical comstructlon, operation, and maintenance features of a large power reactor and could be completed by July 1, 1962. After a'two-year operationai period, a very sound basis would exist for decidlng whether or not to build large molten-salt power reactors. In this proposed program, it should be noted that a substantlal part of the materials compatibility program.would be complete before the major expendi=- tures for an experimental reactor were made. & Fig. 1.1. Isometric View of Molten Salt Power Reactor Plant. Table 1.1. REACTOR PLANT CHARACTERISTICS Fuel enrichm:zny " Fuel carrier Neutron energy' | Moderator "?rimary'eoolant Power Electric (met) Regeneration ratio Clean”(initial) Average (20 years) Blanket Estimated costs Total Capitel _Electric Refueiing‘cyfile.at full power .Shielding , Contrql_ Plant _effiéiency Exit fuel temperature > 90% U°3F, (initielly) 62 mole % LiF, 37 mole % BeFp, 1 mole % ThF), Intermediate Fuel solution circulating at 23,800 gpm 260 Mw 640 Mw 0.63 ~4 0.53 T1 mole % LiF, 16 mole % BeF,, 13 mole % ThF), §69, 800,000 §269/kw 8.88 mills/kwh Semicontinuous Concrete room walls, 9 £t thick Temperature and fuel concentration %0.6% 1210°F at approiimately 83 psia T&ble 1.1, Steam Temperature Pressure Second loop fluid Third loop fluid Structural materials Fuel eircuit Secondsry loop Tertiaxy loop Steam boiler Steam superhester and reheater Active-core dimensions Fuel eqnivalent diameter Blenket thickness Temperature coefficient , (Ak/k) /°F ‘ Specific power Power density Fuel inventozy | Initial (clean) Awerage (20 years) uf,_fijritical mass clean o up ( Continued) 1000 F with 1000 F reheat 1800 peia Sodium Sodium INOR-8 Type 316 steinless steel 5% Cr, 1% Si steel 2.5% Cr, 1% Mo steel 5% Cr, 1% Si steel 8 ft 2 ft (3.8 £ 0,04) x 10~ ~1000 kw/kg 80 kw/liter 600 kg of P37 ~ 9% kg of o ) 26_7 kg of P27 Unlimited 2, GENERAL FEATURES OF THE REACTOR The ultimate -power reactor of the molten-salt.type will probably have & grephite moderetor, since & high breeding ratio is & mejor aim. In order to obtain & breeding ratio as high as 1.0, it will be necessary for the graphite to be in direct contect with the salt and with the nickel - alloy «:cnrl:aa.:!.ner.3 Although it now seems probeble that graphite will be | satisfactory for use in contact with the molten salt (see Part 3), the ‘technology is not considered to be far enough advanced to propose such a -system for the initial reector. This considera.tion led to the specifica- - tion of a homogeneous molten-salt reactor for this design sttjdy. | A number of molten fluoride salts that are suiteble for & reactor fuel are described in Part 2. The base salt chosen for the fuel solvent ' is 8 mixture of L:lTF and Be]‘2 in the mole ratio 62 to 37, respectively. The Li7 and Be base salts iw.ve the moet desirable nuclesr properties of . any of the possible salt cfitions. Beryllium, in addition to having & low neutron s,bsorption cross section, sadds appreciably to the slowing- . down power of the fluorime in the galt. Lithium-7 has the: lowest cross section of the alkali fluorides. The exact percentages in the mixture were determined &s & compromise of two physicel properties: the melting point and the viscasity. The melting point increases &s the L1 content increases, but the viscosity correspondingly decreases. The fuél mixture is prepared by edding to the base salt small emounts of ThE), ahd‘"EUFu, the ThFu being added to provide some regeneration of fissionable ‘materisl in the core end the UF, being added to made the .reactor criticel. The critical mixture cealculated for initia.l fueling of the proposed reector hes the composition 61.8 mole %:1i Tp-36.9 mole % BeF,~1.0 mole % ThE), - 0. 3 mole % UF,+, with the u:ra.nium 93% enriched with lf?3 5 5 é 3. K. Ergen, A. D. Gallihen, C. B. Mills, eud D. Scott, "The - Aircraft Reactor Experiment,” Fuc. Sci. and Eng. » 2 826-8110 (1957) The selection of an 8-ft-dia core for this study was based primarily ~ on the criterion of critical inventory as indicated by nuclear calcula=- itions covering core diameters of 5 to 10 ft. (Details of the nuclear calculations are given in Part 4). The initial critical inventory for S a U235 fueled reactor could be as low as 100 kg, which corresponds to - & critical mass of about 50 kg. In actual practice, however, thorium (that is, 1 mole % THF)) is added to the fuel to improve the Tegeneration ratio and thus reduce fuel costs, and the resulting initial critical in=- ~ ventory is about 600 kg. With thorium in the fuel, the 8-ft core is a » feasonable choice that yields a good conversion ratio for a given invest- _“menta Further, the 8-ft core provides sufficient volume for the average .nower density in the core to be less than 100 w/cm3, which is well within safe limits.' The gamms heaeting in the thicker parts of the core shell ~was also taken into consideration, and it was estimated that with the 8-ft core the heating in the core shell‘would amount to 12 w/cm.; which is not expected to create significant thermal stresses., . It was decided that it would be worthwhile to include a blanket in this reactor'system, despite the fairly high neutron absorption of the core shell material, since the blanket would add between 0.2 and 0.3 to the regeneration ratio and the increased saving in fuel cosis would amount to about.$l,000,000 per year. Although the blanket adds some complications to the reactor vessel, it offers compensations such as serving as a ther= mal shield and as a convenient coolant for the fuel-expansion-tank dome, which is subject to rather severe’ beta heating by the off-gas. The 2-ft- , 'thick blanket Will allow less than 2% of the neutrons leaking from the 1»core to- pass through it without capture,_ The salt mixture Li7F-BeF —ThFu | li"was chosen for the blanket and its composition vas selected as that which. ‘I_‘would give the highest ThFu content consistent with a melting point at ~ least 100°F below ‘the. lowest temperature expected in the blanket region. fliThis specification led to the composition 71 mole % LiF, 16 mole % BeFp, *13 mole % ThFh, which has a melting point of 980°F More recent chemical data indicate that up to about 16 mole % THF), can be used'without in- creasing this melting point° An alloy with the nominal composition 17 wt % Mo, 7 wt % Cr, 5wt % o Fe, and TL wt % Ni, which is designated INOR-8, wes chosen as the struc- ‘tural material for all components of the reactor that will be in contact s with molten salts. Details of the characteristics and fabricability of .this alloy are presented in Part 3. v} - .The ehoice of the power level of this design study was arbitrary, ;since.the 8ft resctor core is capable of operation at power levels of u§5£0_1900 Mw (thermel) without exceeding safe power densities. An eleefirical generator of 275 Mw capacity was‘chosen, since this is in | theieiie range that & number of power companies have used in recent years; ‘and & plent of this size could be justified in almost any section of the United States. It is estimated that about 6% of the power would be used in the etation, and thus the net power to the system would be about 260 Mw. 'Two sodium circuits in series were chosen as the heat transfer system between the fuel salt and the steamn. Delayed neutrons from the circulating fuel will activate the primary heat exchanger and the sodium passing through it. A secondary heat exchanger system in which the heat will ‘ transfer from the radioactive sodium to nonradiocactive sodium will serve - to prevenfi contamination of the steam generators, superheaters, and re- heaters. A non-fuel-bearing molten fluoride salt is a possible alternate choice for the radicactive intermediate coolant and has some advantage ~in that it is compatible with the fuel. In the design ‘adopted no dsnger is expected to arise from mixing of the fuel salt and sodium, however, and Therefore the cheaper sodium system is preferred. | The fuel flow from the core is divided among four circults, so that there are four primary heat exchengers to take care of the core heat generation. This number of hest exchangers was based on maximum size Vs thermal-strain considerations. Each of the four parallel heat tranefer circuits originating in the fuel system trensfers the heat through two sodium circuits to the steam generators. A similar single circuit is o 2 | providedete'rembve the heet generated in the blanket. | Other 1inkages between the fuel and steam thst have been proposed are a saltato mercury-vapor system and a salt-to-helium ges system. The 1atter system is currently being studied. ' ‘A plan view of the reactor plant 1ayout is presented in Flgure 1.2, and an elevation viev is shown in Figure 1.3. The reactor and the pri- mary heat exchangers are contained in a large rectangular reactor cell, which is sealed to provide double contéinment for any leakage of fission gases'and in which all operations must be cerried'out remotely after the reactor has operated st poser. The primary heat exchangers are laid out -0 provide an ineline heat exchange system. The rectangular configura- tion of the plant permits the grouping of similar equipment with a mini- mm of floor space and piping. The primary sodium circuits are thus located in ome bay under a crane, and in the next bay'are the secondary sodium circuits, the steam generators, superheaters, and reheaters under 'another crane. The plant includes, in addition to the reactor and heat excnanger systems and the electrical generation equipment, the control room,‘chemicsi processing equipment, end the fill-and-drain tanks for " the liquid systems. 3. MOLTEN.SALT SYSTEMS 3.1. Fuel and Blenket Circuits The primary reactor cell, which encloses the fuel and blanket cir- fl.ycuits, is a rectangular concrete structure 2k £t wide, 68 £t long, and -:,}_70 ft high._ The walls are msde of - 9-ft-thick barytes concrete to pro- p;.lvide the biological shield. Steel 1iners on both sides of the concrete __J_fg:_wall form a buffer zone to ensure that no fission gos that may leak into p:iithe cell can escape to the atmosphere and thet no air can enter the cell. An inert ‘atmosphere is. maintained in the cell at all times so that & __5?7sma11 fuel leak will not lead to accelerated corrosion. Penetrations '*a;;for pipes and‘for electrical and instrument lines are sealed on each side ‘of the enclosure, . hB. W. Kinyon and F. E. Romie, Two Power Generation Systems for a Molten Fluoride Reactor, Presented at the Nuclear Engineering And Science Conference of the 1958 Nuclear Gongress (March), Chicago, Illinois. ~9- oL Fig. 1.2, Plan View of Molten Salt Power Reactor Plant. L[l Fig. 1.3. Elevation View of Molten Salt Power Reactor Plant. Once the reactor has been at power, the radiation level of the réactor and the fuel and blanket circuits will be so high that it will not be possible to perform dirgct maintenance operations onlequipment in these circuits. All equipment that might require replacement by re- mote means is instelled in the shielded reactor cell. The alternative of segregating each piece of equipment in a separate enclosure is more costly in terms of space, shielding, piping, and fuel inventory. An air lock is provided through which the crane and maintenance equipment can be brought into the.ceil. The principal items of equipmént in the reactor cell are the reac- tor véssel, the fuel and blanket pufips, the fuel and blanket heat ex- changers, heating and insulation equipment, and the reactor cell cooling system, and, of course, there are many electrical and miscellaneous plumbing lines. The reactor vessel and the fuel and blanket pumps ére a8 closely coupled, integrated unit (Figure 1.4) which is suspended from a flange on the fuel pump barrel. The vessel itself has two regions -~ one for the fuel and one for the blanket salt. The fuel region consists of the reactor core surmounted by an expansion chamber, which contains the single fuel pump. The blanket region surrounds the fuel region and ex- tends above the expansion chamber, and the blfinket salt cools the walls of the expansion chamber gas space and shields the pump motor. Four tangential pipes serve as ducts to return fuel into the lower conical section of the core. The core shell, which should be as thin as possible in order to reducé neutron absorption end to keep thermal strains from gamma heating as low as possible, was chosen to be 5/16 in. thick to provide adequate strength against buckling under conditions of maximum pressure differential between the blanket and the core regions. The floor of the expansion chamber is a flat disk, 3/8 in. thick, which serves as & diaphragm to absorb differential thermal expansion between the core and the outer shells. During reactor operation, the bt A YA 1 1 e i s " FUEL-RETURN UNCLASSIFIED ORNL-LR~DWG 28636A FUEL PUMP MOTOR MOTOR FUEL EXPANSION TANK — BREEDING BLANKET Fig. |.4. Reactor Vessel and Pump Assembly. -13 - temperature difference between the corc and the outer shells will be small, but during thé preheating and during power transients the dif- ferehcé may be higher. The diaphragm will safely allow for differ- entlal expansion corresponding to a temperature difference of 200°F without undergoing appreclable plastlc strain. '-The pipe ducts that enter the reactor_tangentially'will impart a swirling action to the fuel and keep it turbulent near the wall. No fluid flow patterns have(&et been obtained for a core of the shape il- 1ustrated; and flow tests; when made, may dictate changes in the shape} The pump‘for circulating the blanket salt is not supported directly on the reactor vessel, but iS'located.to the side and at a slightly higher elevaticn,tc give full blanket ccverage of thc”reactor at all times. The general requirements of pumps for the;fuel and blanket circuits are-discusscd*in Part 5. This design study is¥based on use of a pump of the type shown in Figure 5.2 of that section, which has a capacity of 24,000 gpm. Based on a fuel volume of 530 £43 in the entire circu- lating'system, the fuel will meske a complete circuit through the reactor, - piping, and heat exchangers in 10 sec. A 1000-hp motor will be needed for the fuel pump, and a shaft speed of 700 rpm will be required. As indicated in Part 5, this pump incorporates advanced features not present in any'mclten-salt pumps operated to date. Some consideration was given to the use of multiple pumps, but since a single fuel pump simplifies the top portion:of the reactor assembly, it was adopted for this study. The blanket pump will be similar to the fuel pump, but of smaller capacity. Sincé the heat generation in the blanket will be no more than 10% of the total heat generated, a pump of about 3200-gpm capacity is requlred. | Four primary heat exchangers are provided for the fuel circuit so that each'heat exchanger will be of reasonable sizepl These heat ex=- changers are designed to have the fuel on the shell side and sodium '__ ‘inside the tubes. This arrangement is contrary to that which might - 1l - fl. o e ol e e 0 1AL ST L W . of power level is improved by keeping the high cross section Xe intuitively be proposed because it might be expected that the fuel volume would be lower if the fuel were inside the tubes. However, the superior properties of sodium es & heat transfer fluid are not realized with the sodium on the shell side, and therefore the over-all system is most compact with the sodium.inside the tubes, The heat exchangers (Figure 1.l) are of semicircular construction, which provides for convenient piping to the top and bottom of the re- actor. The blanket heat exchanger is similar in construction, but it is scaled-down to be consistent with the smeller heat load. A more detailed description of the hest exchanger is given in Teble 1.2 of Section h 3.2, Off-Gas System An efficient process for the continuous removal of fission~product. gases ig provided that serves several purposes. The safety in the event of & fuel spill is considerebly enhanced if the radiocactive gas concentration in the fuel is reduced by stripping the gas as it is formed. Further, the nuclear stability of the resctor under chigges continuously at & low level. Finzlly, many of the fission-product poisons aié; in their decay chains, either noble gases for a period of time or end-their decay chains as steble noble gases, and therefore the buildup of poisons is considerably reduced by ers removalo_ The solubilities of noble gases in some molten salts are given inFart 2, end it is deduced that solubilities of similar orders of *V”’magnituae are likely to be found in the LiFwBeF, salt of this ‘study. - :,i[It vas found that the solubility obeys Henry‘s law, so that the equi- : 7g:librium solubility is proportional to the partial pressure of the gas rd in. contact with thc salt. In principle, the ‘method of fission-gas Tes ,Vffimoval consists of providingAan»efficientumechsnism for contacting the ~ fuel salt vith an inert ambient ‘gas in which the concentra.tion of xenon - snd krypton is kept very 1ow. f{,v'.:-f | | | ’ «l5 In the system .chosen; approximately 3.5% of the fuel flow is mixed with helium purge gas and spreyed into the reactor expansion tank. The mixing and spraying provide a 1arge fuel-to-purge-gas interface, which promotes the estsblishment of equilibrium.fission gas concentrations in' the fuel. The expansion tenk provides a 1iqu1d gurface area of epproxi- mately 26 £4° for removal of the entrained purge and fission gas mixture. The gas removal is effected by the balance between the difference in the density of the fuel and the gases and the drag of the opposing fuel ve=- looity.‘ The surface velocity downward in the expansion tank is approxi- mately 0.07 ft/sec, vhich should screen out all bubbles larger than 0.008 in. in radius. The probability that bubbles of this size will enter the reactor is reduced by the depth of the expansion tank being sufficient to'ellow time for small bubbles to coalesce and be removed, The liquid volume of the fuel expansion tank is approximately Lo ft3 and the gas volume is approximately 35 ft3 With a fuel purge gas,rate of 5 cfm, approximately 350 kw of beta heating from the decay of the fission gases and their daughters is deposited in the fuel and on metal surfaces of the-fuel expansion tank. This 350 kw of heat is partly re- ,moved by the bypass fuel circuit and the balance is transferred through ?the expansion tank walls to the reactor blanket. o | iThe'mixture of fission gases, decay products, and purge helium 1eaies the expansion:tank through the off-gas line, located in the top ef the tank, and Joins with a similar stream from the blanket expansion tank (see Fzgure 1.5). The combined flow is delayed approximately 50 min in a cooled volume to allow & large fraction of the shorter lived fission pflpducts to decay before entering the cooled carbon beds. The carbon beds provide a holdup time of approximately 6 days for krypton and much 1oqger for Xenon, The purge gases, essehtieily free from activity, leave'the carbon 'bed& to'join=the gases from the gas-lubricated bearings of the pumps. The gases are then compressed and returned to the reactor to repeat the cycle. Approximetely every four days one of the carbon beds that hes _ been’ 0perating at minus ho F is warmed to expel the Kr85 end other longu_ '11ved fission: products _ - ",; | - . t - 16 - - '(’ -1 - c - ~ C.C UNCLASSIFIED ORNL-LR-DWG, 28439 R _|,, = wmw Hs R i N K 13SCFM b MOTOR FUEL [~ j Qfimw& /95 scem_ (T \BLANKET pump [ — Q_Jeume | | /955ch HE —] ¥ 0.65 SCFM ] 1 | | [Beamwer exeansion — ; - TANK D.ICFJf-/Z/O"F 5.3 SCEM FUE gxpflfly/0” e 12I0°F —Lflzfl/w([?' BY-A4SS ;gfs'vg"/-" L ’MFJ—E’%-L 1 ez ‘YL— BLAVKET »°F 2109 r el bwms . EeT L8CES L AES0F Le COA’E V saszcem = S /5940%1 L 6776 CFIM o s P50°F | | A cemsem 3 203CEM 1.63CEM | 2I2F. | 62F —4OF 55 FT° 785 FT° '_,_—'|__5/_F'_d /6 FT8 x| % I\ x ¥ | X | 4 =11 ot 5 ' 7 25 0 755 FI° I6FT. BEF 22F 4 62%, 40F,, | - 4~ CLOSED BLOCK VALVE S OPEN BLOCK VALVE ouT T//v l/fl Tozrr loz/r Tw 800°F 105F 105F 00°F OF OF ~40°F ~dO'F COOLANT COOLANT COOLANT COOLANT Fig. 1.5. Schematic Flow Diagram for Continuous Removal of Fission Product Gases. " .3. Molten Salt Transfer Eguipmsnt Ths fuel transfer systems are shovn echematically in Figure 1. 6o Salt freeze valwes, described 1n Part Dy are used to isolate the indi- - t-vidusl components in the fuel transfer linee and to isolate the chemical 'plant from the componente. in-the reactor cell. With the exception of the reaetor dreining operation, which is described below, the 1iqu1d is transferred from one veessel to enother by differential, gss pressure. ' 'By this means, fuel may be added to or withdrswn from ths reactor during tpower operation. " ' - The fuel added to the reactor will have 8, high eoncentrstion of U235Fh, with respect to the process fuel, so that additions to overcome burnup vill rpquire transfer of only a small volume; sindlsrly, thorium- bearing molten salt may be added at any time to the fuel system. The thorium, in addition to being 8 design constituent of the fuel salt, ~ may be added in amounts requirea to serve es & nuclear poison for ad- | Justing the mean core temperature. | When fuel is removed from the reector, it first goes to onhe of the' withdrawal tanks. These tanks will be sized to serve ‘&8 holdup vessels from which material mey be later transferred to the chemical blsnt. The chemical processing plant is considered to be.an integrsl part'offg,' the reactor complex; however, the chemical processing plsnt’is set apart from the reesctor, is conteined in separate cells and has & sepsrste con= trol center. As indicated elsevhere, the fuel-reprocessing cycle assumed for this report requires an average daily withdrewal end addition of gbout 2 ft3 of fuel. If a 30-dsy'holdup of fuel is required for cooling "~ before chemicsl processing, the withdrswal vessels must provide & volume of 60 ft3 . They will require both & heating and a cooling system simi- ummmwmmuwmmmm1mmmmmmmmMu' to. msintsin the temperature within reasonable . 1imits. -"'18- e 61 UNCLASS IFIED ORNL-LE-DWG, 20630 S LEGEND: ' — FUEL LIE ——— —-— GAS LINE e ‘ _______.__..-—--—r«E-.r —g—.rozmow VALVE 1 L REMOTE OPERITED —r'E"" |—£—-y ~42r—m/.x/£ {EVY-FREEZE WaivE S —sueoly V — VENT | i | —— N FLOW t DIRECTION 4 BY-A45S REACTOR HERT EXCHANGER WITHDRAWAL Lzas ENRICHER TANK e ——_—_—— e e 4 ~ > m—gq P T T T T T m et e e o - - A | [l ' I I i Ha * Y : Y : I I ' | _ t . : ! - [ | } — : ] | : MF : [ l . ADSORBER ! MAN FILL € DRAN SYSTEM | I I — *— i ~ALUORNATOR % - - | \ wrkaenm, | TERYR cuemmca processve ALANT . ] Fig. 1.6 Schematic Diagram of Fuel Salt Transfer System. For the main fuel drein circuii, mechanical valves will be placed ~ in series with the freeze valves to establish a stagnant liguid suitable for freezing. Normally th:ese‘mecha,niéal valves will be left open. . Draining of the fuel will be accomplished by melting the plug in the freeze ~ line a,nd,_aliowing‘ the fuel to drain by gra.vi'by._ By opening gas _pressgre | equallization velves, the “liguid in the reactor will flow to. the dré.in tank, _and the gas in the drain tank will be transferred to the reactor system. Thus gas will not have to 'be added to, or vented from, the primary system Two valves are 1ocated. in parallel in the fuel drain line g0 that a spa.re path will be available in the event of failure or need for repa.irs. All the liguid transfer lines will be equipped with heaters and covered with insulatibn so that the system may be held at temperatures a.bove the fuel melting point. Since the main drain line is at the 'bottom of the rea.ctor, there will always be fuel in it. This line is kept from freezing or overheating by use of a circulation bypass, as shown in Fi‘gure .6; to keep the stagnant portion confined to the freeze valve area. This 'bypass prondes a certainty that the drain line to the freeze valve is elways at temperature snd open for draining. The blanket fluid transfer system is essentially the same as the fuel system. A chemical processing plant will be provided for the blanket salt, which may serve as & backup capacity for the fuel reprocessing ~ plant.. 3.4. EHeating Equipment The melting points of the process fluids used are all well above room témperature. It 1s thus necessary to provide a means of heatihg ’ all pipes and equipment- conta'.ining' thege fluids. This will, in general, be accomplishedj- by providing electric heaters to all pipes and equipflent.i Inside the reactor cell, the heaters are 1ncorp<_>rated in removable 'a_s.- | semblies that consist of the heaters and the insulation, as shown in Figure 1.7. Outside the cell, conventional methods of insta.lling heaters .. and insulation ere used, i ) L LIETING £ SPREADING - MECHANISM ———— Figo 1 o7 e b e 345 Auxillary Coollng Coollng is provided in the reactor cell to remove the. heat lost through the pipe insulation and the heat generated in the structural steel pipe and equipment supports by gamms -ray absorption. The heat | is removed by means of forced gas circulation: through radiator-type | '-space coolers. A cooling medium, such as Dowtherm, in a closed loop rémoves heat from the space coolers and dumps it to a water heat ex- - changer. 3.6. Remote Maintenance Provisions are made to carry out sll maintenance operations in ‘the reactor cell by remote means. While some small repairs may pos- sibly be made in the reactor cell, the principal requirement is to be able to remove and replace by remote means all the necessary components in the reactor cell. This will include pumps, heat exchangers, pipe, - heaters for pipe and equipment, instruments, and even the reactor ves- sel, A prime requisite for remote maintenance is a reliable method of making and breaking joints in pipe. This can be done, either by de- veloping & remote cutting and welding process or by developing a satis=~ factory flanged pipe joint (see Part 5). (The development of a remote welding process is underway at Westinghouse on the PWR project,) All equipment and pipe Joints in the reactor cell are laid out so that thsy;are accessible from above, Directly above the equipment is a traveling bridge on which can be mounted one or more remotély.opera- ted manipulators. At the top of the cell is another traveling bridge for a remotely operated crane. At one end of the cell is an air lock that connects with the maintenance area., The crane can move from the bridge in the cell to a monorail in the air lock. - 22 - @ ) When 1t hes been ascertained that a piece of equipment ghould be . : replace&, the reactor will be shut dovm and drained, and the plece of faulty equipment will be removed according to the following procedure. The manipulator will be trangported by the crane from the main’bena.nce erea through the air lock, to the cell, and pleced on the manipuletor ‘bridge. The manipulator will then be used to disconnect all instrue | men'b, electrical, and service connections from the equipment and to | unfa.sten the fle.ngee tying the ‘equipment to the system. The crane will then remove the feulty equipment and tmnsport it to 'bhe meinte-n nagnce area, The crane will then be used to move & epare 9ieoe of ejquip~ ment 1nto the cell for installation with the use of the ma.nipu]a.toro After eompletion of the replacement, the manipulator will be removed from the reactor cell by the crane, the air look will be closed, and the rea.ctor will be ready for startupo Prelimimry t-ea'bs with & Genersl Mills manipulstor have demonotruted ‘the fea.sibility of renotely removing and repla.cing the rotn.tins assembly of e liguid metal pump, It appears therefore that eat:lsfo,ctory t-chniques can be developed for remote ' ‘maintenance. Closed-circuit television equipnent 1s provided, for V:Lewing the E maintemnoe operetion in the eello A A number of camers.s a.re mounted to - ghow the opera.tion from different angles y and periscopes give a direet view of the entire cell, The mintenance area :Lo divided into hot and cold shop areas. The B . e cold shop will be used. for generel repe.ir work on equiment that can be ‘{_‘_'jhandled d.ireotly. : 'I’he hot ehop area will be used. to (1) repa.ir minte- . nsnce equipment that can not be handled directly; (2) to disassemble o Zifailed equipment 10 de‘bemine 'l:he eause of fa.ilurej; (3) to prepare hot ' '-‘;i'equipment for disposa.l, that is )’ out or disassemble la.rge equipment to. TR f-"“:f*/managea'ble size, plaoe 1n coffina, etc., end (h-) to repeir failed equip-n- ; ; B 'ment within the 1imits of 'bhat whieh ean be d.one with the equipment ree s quired for the other hot-shop opera.tionse A completely equipped. hot ' - 23 « -;shop capeble of making any and ell Trepairs to all eQfidpment does not - eeppear to be economically adriseble for the anticipated maintenance d"work for & single reactor plant. Although it is possible to remove | and replace the reactor, it is & comparatively simple and rugged piece of equipment with & low probebility of railure, and therefore & spare reactor will not be provided. ' ~ Maintenance of the intermediate sodium circuits will be done dia_ frectly. In case of an equipment failnre in one of these circuite, the = loop will be drained. At the ‘end of & feirly short period, for rosidual , fN Eh to decay, it will bve possible to remove the top slab from the second- - ary cell end remove and repdace the faulty equipment by usins the building | l_crane end direct maintenance procedures. Each secondery cell is ‘shielded, d so that the addacent cells need not be drained to make a repair. - ; 7. Fuel Fill-and-Drain Tenk The main fuel flll-end-drein syetem.mnst neet the following wa Jor design criterisa: . : (l) A preheating system must be provided that is capable of main- taining the drain vessel and its connecting plumbing at 1200°F . (2) A relisble heat-removal system must be provided that hes euf-‘ ficient capacity to handle the Puel afterheat. (3) The drain vegsel must be "ever=- safe” so that & criticel cone jdition cannot occur when the fuel is drained. The fuel dreining operation has not been considered as an emergency procedure, that is, one which must be accomplished in a relatively short ~ period of time in order to prevent & catastrophe, There are,. however,_ other 1ncentives for rapid removal of the fluid from the fuel circuit. - If, for example, there were a leak in the fuel system-it would‘be ine - portant to'drain the fuel in order to minimize the cell contamination and cross contaminafion of the systems. Further, rapid removal of the fuel at the time of & shutdown for me.intenance would have an economic edwantage in reducing “the power outage time. e v ¥ -A-cohsideration of these factors indicated that the maximum efter- heat design load should be 10 Mw for & 600 Mw reactor that had been operating'for one year and had been shut down for 10 min before the fuel drain was started, No credit was taken for fission-gas removal during '.operation, It was further estimated that 15 min would be required to lremo?e the fuel from the reactor. For the drain vessel design calculations, it was assumed that at 1200°F the fuel system volume would be 600 ££3. The design capacity of the drain vessel was therefore set at 750 ft3 in order to allow for temperature excursions and a residual inventory., An array of 12-in. - dia pipes was selected as the primary contsinment vessel of the drain system in erder,to obtain & large surface area-to-volume ratio for heat transfer efficiency and to provide a large amount of nuclear poison ma- terial (see Figure 1.8). Forty-eight 20ft lengths of pipe are arranged in six vertical banks comnected on alternate ends with mitered joints. The six banks ef pipe are comnected at the bottom with a common drailn line that comnects with the fuel system. The drain system is preheated end maintained at the desired temperature with electric heaters instal- led in smallédiameter pipes;located.exially inside the 12-in.=-dis pipes. These bayonet-type heaters can be removed or installed from one face of the pipe array to facilitate maintenance. The entire systen is instal- led in an ifisulated room or furnace to minimize heat losses. The removel of the fuel afterheat is accomplished by'filling‘boiler tubes installed between the 12-in.-dia fuel-containing pipes with water - from headers that are normally filled. The boiler tubes will normally "be dry and at the ambient temperature of about 950°F Cooling'W111 be __'°accomplished by slcwly flooding or “quenchlng" the “tubes which furnish . a heat sink for radiant heat transfer from the fuel-containing pipes to the boiler tubes.. For the peak afterheat 1oad, about 150 gpm of water is required o 'supply the boiler tubes. L UNCLASSIFIED ORNL-LR-OWG. 28535 Fig. 1.8, Drain and Storage Tank for Fuel Salt of Molten Salt Power Reactor. 26 vy Thie fill-end-drein system satisfies the design criteria in that it 1s always in & etendby condition, in which it 1s immedizstely evailable for drainage of the fuel, it:can sdequstely handle the fuel afterhest, end it provides double contaimment of the fuel. Heat removel is essen- tially self-regulating in that the emount of hest removed 1s determined by the rediant exchange between the vessels and water wall, Both the water and the fuel systems eaxe et low pressure , and & double fallure would be required for the two fluids to be mixed, The drein system temk may be ee.sfly enclosed and sealed from the astmosphere because there are no large ga.s-cooling ducts or cother major external systems connected to it. A steinless steel trey will be placed below each benk of pi;pes to catch the fuel 1if a leek develops. These treys will be cooled by water . walls to prevent any possibility of meltdown and destruction of the cell. A prelimina.ry eriticality caleulation wes mede in a drain tank as- sembly without cooling walls, A multiplication constant of 0.2 was esti- mated for a fuel containing 0.5 mole % ThFh_ and 0,125 mole % UF), at a temperature of 125005‘. 4. EHEAT TRANSFER SYSTEMS The intermediate heat transfer systems use sodium as the working fluld to transfer heat from the fuel ’co the steam system, The latter accepts the hea.t in the steam generators ’ the superhea’cers ’ a.nd the re- heaters, A diagrem of -bhe heat removal 15 shmm :Ln Fig. 1.9. A speci- fication thet the steam system components should be cempletely radiatidn- o free dictated two sodium circui‘ts for each heat tra.nsfer pa.th. Fbur systems :!.n pa.ra.'llel remove hea.t from the ree.ctor fuel; o _f_-_a,dditiona.l system ha.ndles the power generated :I.n ‘bhe 'blanket sa.lt. Each of the five systems :I.e separa.te and :l.né.ependent u;p to 'bhe point where the superheated steam f.'l.ow paths ,join a.hea.d of the turbine. o e 8¢ r"’} o (BLANKET il UNCLASSIFIED ORNL-LR-DWG, 28640 REHEAT O/LER A 220 PSS/ 690 F FOORS/ /000 < ;;Q . g ~—LPE/ P TURBINES Y LS HG ~—H PTURBINE Y | J_’JEL "‘%’l"/’ 7 x\_" A Y o - —__I~—TURBINE STOP RReRL| | M BAMET i o ~—ATTEMPERATOR A P M—EMERGENCY RELIEF ° ' ' COOLING WATER (0807 - 2% 'CONDENSER 755 By X |8rAass ] SUPERHEATER O | /%9055,’{_ 75t oo : ~—REDUCING STATION 9% | srconpary X SUPER 1 Hé‘AT[A’C‘// 13CFS 1 HEATER ~CONDENSA 7'1." PUMP 28Ps/ | isscrs] (2EaM NDER|——— I30F | {1800 Pst O B s [P ' FEEDWATER i HEATERS NV SE/?/[.S A ' STEAM CHEST: o \[j_ ~-DEAERATOR 1 X : | - FEEDWATER PUMP 265/ 22pPs/ |- ' ' 241 CFS [er/cFs 577CFs 70 BLIANKET 979 LY | A 2T PR3 }7‘0 FUEL 3B8.E6CFS FEEDWATER é % 4T HEATERS /A/&[/?/[.S‘ L apy 1 T~—8onR SODIM _, #oF 1 | /,seqooc_; éff— — Fig. 1.9. Schematic Diagram of Heat Transfer System. Each primary sodiuwm circult includes & primery heet exchanger in the reactor cell and a pump and & secondery heat ext:hanger located in a cell adjacent to the reactor cell. No control of flow ratee is required, so there are no valves, and constent=-speed centrifugal pumps ere used. With a \pmnp‘ stopped, thermal-convection flow would be avaeilable to remove afterheat from the reactor. The secondary heat exchangers are of the U=tube in U-shell, counterflow design, with the sodium of the primary eircuit in the tubes and the sodium of the secondary circuit surrounding the tubes. In order for the sodium to be at a lower pressfire‘ than that of the fuel in the primary heat exchengers, the pumps for the primarj | sodium circuits are on the higher temperature legs of the circuits, The essentlal characteristics of the various heat exchangers are described in Table 1.2. The secondary sodium circuits, except for the secondary heat ex= changers, are outside the shielded area and thus are avallable for ad- Justment and maintenance &t all times., Three paths are provided for the sodium flow from a secondary heat exchanger: a steam superheater, a steam reheater s 8énd & bypass line for control. Regulating va.lves autos matically adjust the flow to sult the load conditions, so that at very low loads most of the flow is through the bypass line, | The three sodium streams are recombined in & mixer or blender, which leads to & three-way valve. At design point, sbout one-third of the flow returns directly to the pump suction a.nd two-thirds enters the steam gen- era.tor a8 the driving streem of a Jet pump The Jet pump, located verti- ca.lly along side the steam generator 5 is assisted by thezmal-convection o i‘low ,mra in the Jet pmnp and downward in the bc::i.le::-° At 1ow power r-.1.e1@r«eil.s ’ 'bh:l.s serves to ma.inte.in 8- good recirculation rate a.nd ensure good : .,stability of control.‘ 'J.‘he three»way va.lve permits 'l',he steam generator ci:rcuit to be isolated from the rems.:l.nder of the sodium 80 tha’c at zero '_rpower the entire ’ooiler becames isothermal at the saturation temperature, amd the pressure :l.s maintained a'b the desired level. ' - 29 - | fuel ‘and Sodium to Sodium Exchangers Number required Fluid Fluid location - Type of exchanger Temperatures Hot end, °F @®l1d end, °F Change AT; hot end, °F AT, eold end, °F AT, log meen, °F -Og- Tube Data Material Outside dlameter, in, wall thickness, in. length, £t Number Pitch (4A), in, Bundle dlameter, in. Heat transfer capacityé Mw Heat transfer sxrea, ft Average heat flux, 1000 Btu/hreft2 Thermal stress,* psi Flow rate, £t3/sec Fluid velocity, ft/sec Maximm Reynolds modulus/1000 Pressure drop, psi * [ch/(l-y)] [(max AT-of wall)/2) ‘ . . , e Table l.2. Data for Heat Exchangers Primary System 4 Fvel salt Primary sodium Shell Tubas U-tube in U=-shell, countexrfliow 1210 1120 1075 925 135 195 o0 150 11T7.5 INOR=8 1.000 0.058 25.7 515 1.1 28 144 2800 175 9200 130 l‘l’ h‘6 ¢ l 10.8 19.7 %.5 523 40 15.5 Secondary System 4 Primary sodium Secondary sodium Tubes | Shell - U-~tube in U=-shell, - counterflow 1120 1080 925 825 195 255 100 65.6 Type 316 stainless steel 0.750 0.049 21.5 1440 - 0.898 36 14l 5200 95 8200 46,1 - 33,6 13.9 13.2 270 166.0 10 14.8 (: ¢ 3 Sodium to_Steam Exehangers - Number required ) Fluid - Fluid loca.tion . _ Type of exchanger _ Te'mpera.tures - Tg- Hot end, °p . Cold end, °F Change , °F AT, hot end, o AT, cold end oF AT, log mean, °F Tube Datea Materisl o Outside dlameter, in. Wall thickness, in. Length, ft ‘ -Number Pitch (4), in. Bundle diameter, in, Heat transfer capacit Heat Transfer aresa, f é Average heat flux, 1000 Btu/hre ft2 Thermal stress, psi Flov rate, ft3/sec 1000 1b/hr Fluid velocity, ft/sec Maximum Reynolds modulus/1000 Pressure drop, psi Table 1l.2. Steam Generator .;_ o l Secondary Water sodium Shell Tubes - Bayonet, countertlow 8as 621 O - 62 85 . ' 0O - 119 20k 158 2.5% cr, 1% Mo Alloy - S 2 0.180 18 362 2.75 55 82.2 2800 100 18.600 5T+5 5.6 200 5.7 (Jet pump) 410 * [aE/(J.-})] [(max AT of wall)/2] ¥ (Continued) Superhea'ber : 4 Secondaxy Steam sodium Shell Tubes U=tube in U=-shell, counterflow 1080 1000 930 621 150 : 379 | 80 309 169 5% Cr, 1% Si Alloy 0.750 0.095 25 480 1.00 23 39.2 1760 76 9000 15.5 406 9.3 = 61 164 306 6.9 10.3 Reheater , L Secondary Steam sodium Shell Tubes Straight, counterflow 1080 | 1000 1000 - 6ho 80 360 - 80 360 - 186 5% cr, 1% 51 Alloy 0.750 0,065 16.5" 800 1.00 29.7 22.6 ' 2200 35 ' 5000 16t8 399 T.9 137 163 167 3.2 10. 4 A portion of the cooled sodium leaving the steam generator circuit returns to the pump suction, which is constructed as a blender, end mixes with the stream bypassed through the three-way velve, The centrifugal pump is specified as two-speed, with the second speed being one-fourth of full speed to give essentie.'l.ly one-fourth of the full flow so tha.t the power output may be more easily regulated from 25% dovn to very low levels. The steam genere.tor , Fig. 1.10, consists of tubes suspended in the flowing sodium. In this Lewis~type boller, the water flows through & centrel tube to the bottom of each beyonet and boiling occurs during upwerd flow in the outer emnulus. Baffles in the steam dome separate the water and steam, and the water returns to & tray which collects it for recirculation, Just below the tube sheet and above the godium, & thermal barrier and & gas space are provided to permit the tube sheet to he at the satu- retion temperature and thus avoid thermal stresses. In addition, the gas space will serve &s & cushion for the initial shock in the event & tube ruptures and water leaks into the sodium. The superheater 1s a U-tube in & U-shell, counterflow exchenger, with the steam inside the tubes. As in the steam generator, there 1s a thermal barrier between the sodium and the tube sheet &t the cold end in order to minimize thermal stresses. The reheater consists of straight tubes in a straight shell. With the sodium on the shell side, the tem- perature difference between the shell and the tubes is sufficiently small that this more economicel construction can be used. The reheaters are " located mear the turbine’ to give a small pressure drop in the reheated steam. %R, H. Shannon and J. B. Shelley, "Double Reheat Cycle - Next Step?” . _Power February 1955 p 98~99. -32 - e e G e o St . o e St s Ao A UNCLASSIFIED - ORNL-LR-DWG 27480 f——MMIER |— WATER AND = STEAM TUBE SHEET e N i il ] | \!-ll THERMAL fh BARRIER l L 7 - i .10 Steam Generator, Lewis Type. N — 5. TURBINE AND ELECTRIC SYSTEM Steam is supplied to the 275-Mw-rated turbine at 1800 psie end 1000°F, The single shaft of the turbine operates at 3600 ¥pm; there are three exhaust ends, The turbine heat rate is estimated to be 7700 Btu/kwh | for & cycle efficlency of 44,3%. The electrical genera.tor end station heet retes eve, respectively, 7860 and 8360 Btu/kwh. With 6% of the electrical generator output used for station power, the supply to the bus ber is 260 Mv. These estimates are baged on Tennegsee Vs.lléy Authority heat belances for & turbine of this type ,6 with athstnients - mede for the modified steam conditions' end the different plent require- ments of the molten-salt resctor system. The condj.tions given above were selected to give the minimnn cost. Increased cycle efficiency could be obta.ined with higher tanpera.tures - end, higher Pressures, but the ilncrease in efficiency would be offset by the :anrea.ses in equipnen‘b costs associa.ted with the higher temperatures and pressures. 6. NUCLEAR PERFORMANCE The nuélea.r behavior of the particular molten—sait reactor and fuel processing cycle selected for this design study is presented in this gection. (A more detalled parametric study of the nuclear performence of . molten~salt reactors is givén in Part 4.) The reactor could utilize either 0233 or 1123 2 but, since U2 55 'is the only isotope presently available in A qua.ntity, it was selected as the primary fuel, For comparison with other reactors designed to use 023 5 8s a fuel, the performsnce of molten-sal'b reactors fueled with *>7 is given in Part k. 6’.[‘he date used were for Units Nos. 5 and 4 of the Gallatin Steam Plant, Gallatin, Tennessee. 7H. R. Reese and J. R. Carlson, “'I.'he Performance of Modern m:flaines " Mech. Engr., March 1952, p 205, Fbr the nuclear ana.‘l.ysis , the rea.ctor was conceptuslly resolve& into & spherical core having & wniform temperature of 1180°F, a thin epherical aore shell of INOR~8, & spherical snnulus of blanket fluid, snd & spheri- cal reactor shell, A blanket thickness of 2 4t appeared to be sufficient %0 prevent excessive losses of neutrons to the outside » énd a core vessel ‘thickness of 1/3 in, wvas used. A reactor shell thickness of 2/3 in, was selected for the calcula.tions, but, in meny cases, the reactor shell was neglected in order to shorten the calculetions. The remaining independent va.z'-ia.bles of significence were the concen- tretion of thorium in the fuel salt, the diameter of the corg, and the fuel salt reprocessing rate. Of prineipal interest were the oorresponding eriticel inventories of 02 5 and 02 35 end the regeneration ratio, In Pexrt 4, the results of & paremetric study of the initial states are pre- sentedj that is, the results are for "cleen" reactors, having no fission fregments or nonfissionsble 1sotopes of uranium other than 02 38 present, However, the optimum system could not be determined from such & study alone; in pexrticular, the time efter startup when proceseing is initiated, the method of processing, end the rate are important factors. The para- metric study of various fuel reprocessing schemes that is under way at present is deseribed in Part 4., This study 1s not yet complete bece.use ' the number of possible combinations of irdependent variables is quite lerge. Therefore, a typleal get of cornditions, vhich may turn out to be nearly optimum, was selec'bed for ;presen'bation. A core diameter of 8 £t end a ’bhorim concentration of 1.0 mole g4 in the ruel eelt were select.ed a.s & reasoneble compromise between the ~ desire to minimize the - inventory oi‘ u2 35 a.nd to- maximize the regenera.- tion ratio, The nuclear perfomance of the initia.l sta.te is set fourth in T&ble l-5¢ . A oonvereion ratio of 0 63 ie believed to be a.bout the maximmn tha.t can be obtained in a. homogeneous molten rluoride sa.lt system with U255 es the fuel (see Part 4). The perfoma.nce with Ue 35 would be subetan- ‘ tial]y ‘oetter, of course ’ a.nd regeneration ratios of 0.90 or higher could 555- | T&blg_l;j. Initial Ruclear Characteristics of e Typicel ‘Molten-Fluoride-Salt Reactor Core diameter: 8 ft Power: 600 Mw (heat) | Volume of external fuel eystem: 339 ft” - U-235 inventory: OO04 | ' Regeneration ratio: 0.63 . : Cation ' ‘ ~ Inventory Concentration ~ Atom Density NEutron Absorption (xg) __(mole %) (Atoms/cm”) ___ Batlos* | - ew0® Core . | BRI ~ o U-235 60k 0.25k 9,09 - Fissions | ' S 0.729 n-y o U 0.27T1 ¢ U-238 k5.5 0.019 0.6Th . 0.039 'fl‘l 2100 100 . '32.0 ) 0.%1" 1 3920 61 1982 . ( Be 3008 .37 1183 (0.102 F 24000 - 47T ¢ Core Vessel 0.052 Blanket | | ' o | - Th 30500 13 392 0.228 I 50% Tl 2139 Be 1460 16 : 482.2 - (0.021 F. 25100 l;671 ‘ \ - .Ieakagé 0.004 ‘Neutron yield, 7 1.8 % nbuirons absorbed per neutron absorbedfin flféfis. o be obtained in the clean reactors. Further, as discussed in Section 2 above, the use of graphite to moderate the fluoride regctor may result - in substantisl improvement. The neutron balance is presented in terms of neutrons ebsorbed in each element per neutron shsorbed in U235 Thus the sfim of the abéorp- | tions in thorium end U2 give, directly, the regeneration ratio, end the sum of all the ebsorptions givesM, the number of neutrons produced by fission per neutron absorbed in U255 An examination of Table 1.3 shows that sbout one-third of the regeneration tekes place in the blenket. The single, most importent loss of neutrone is to radia$ive capture in U255; if other parasitic captures and leaskages could be reduced to zero, the regeneration ratio would still be limited to 0.80 in this reactor by rediative capture in 122, fhe other importent losses are to carrier galt in the core and to the core vessel, which reduce the regeneration ratio by 0.10 and 0.05, respectively. Losses to the blanket salt and to leskage smount to less than 0,02 neutrons. | “ Of the neutrons lost to the carrier salt, the majority are captured by fluorine, and the loss is unavoldable. The lithium is specified to A be 99,99% L173 end the 116 content is estimated to be about equal to that which would be in eguilibrium with the n- reaction in beryllium. Hence, there is no point in specifying & lower concentration of I&sa The system contains nearly 10,000 kg of purified Li7, Of the neutrons lost to the core vessel, about ohe-third ere captured by the molybdemum; nickel cap~ tures account for mpst of the ramaining loss. Increasing the hardness of the neutron spectrufi by increasing the thorium concentration tends to decrease the absq;ptidhs in the carrier salt and in the core vessel, but . this decrease 1s ‘moré than offset by the decline in N of U255 at higher:' energies. | The accumulation of fission fragments end nonfissionable uranium isotopes tends to increase the inventory of U255 and to depress the re=- generation ratio. Ehe production of U255 tends to ‘counteract these L effects, N6verthe1ess, if the fission products are not removed, the fiBTu inventory of U-2? will increase repldly £ram 600 to 900 kg during the first year of operation. The regeneration ratio will fall from 0.65 to 0.53 in the same period. About TO kg of 023 > will have accumulsted, of which 85% will be in the fuel salt. The nuclear characteristice of the system et the end of the first year are presented in Teble L.k. As mey be seen, the increesing hardness of the neutron spectrmn results in a decrease of losses of neutrons to the fuel salt and to the core vesssl to O. 012, but this sa,ving is more than offset by the corresponding decline in M to 1,78 (a.veraged over all three fissionable :Lsotopes present). | If the fission products were ellowed to &cemmzlate further, the '5123 2 inventory would contlnue to rise. If, hawever, the fuel salt is reprocessed continuously at the rate of one fuel volume per year {thus holding the fission product concentretion constent), the 025 2 inventory and regeneration ratio can be held stationary, as shown in Part L, Fig. 4,10, The continual incresse in the concentrations of nonfissioneble uranium isetopes is compensated by the accumulation of 623 3. A neutron balance for the system at the end of twenty years is giveri in Teble 1.5. | As may be seen, 0256 is much more harmful than 0238 s 8ince it cap~ . tures 2.5 times as many neutrons and doez not form & fissionable isctope. Despite these losses, however, the regeneration ratio does not decrease appreciably, mainly because of the superior properties of 0233 s Wwhich provides 4§0% of the fissions. In sxmnary, once reproeessing t0 remove fission products 1is ‘begun, nuclear performance of the gystem is stabilized to a satisfactory degree for twenty years.‘ No provision for the rexhova]_. of the nonfissionable isotopes of uranium need be made. | | If desired, the trensients during the first year of operation cen’ be largely eliminsted by allowing the thcriinn congentration to decrease, rartly through burnup end partly through withdrawal. Such & case is | shown in Fig. 4.10 as a dashed line, in which the core reprocessing 1s - 38 - > o) Table 1l.k4. Nuclear characteristics of & Typleal Molten-Fluoride-Salt Reactor Power: Load factor: 3 Volume of external fuel system: 539 ft Core diameter: 600 Mw (heat) 0.8 8 ft After Operation for One Year Without Reprocessing of the Fuel Salt U-235 inventory: 890 kg Regeneration ratio: 0.53 _; Neutron Fraction Inventory Concentration Atom Dens%ty Absorption of (kg) (mole %) (Atoms/cm”) Ratios* Fissions x 1012 Core . U-235 890 0.43 13.4 ' Fissions . 0.618 0.861 n-y 0.262 U-233 61 0.029 0.926 Fissions 0.090 0.126 n-y 0.01k Pu'259 608 0-003 00101 ' Fissions - 0.009 0.013 n.""'7 .. 0 0006 Th-23%2 2100 1.0 32.0 0.299 Pa-233 8.2 0.00k 0.125 0.005 1i~Be-F 0.080 U-234 1.9 0,0009 0.029 0.001 U-23%6 62.2 0.030 0.933 0.032 Np~237 4,2 0.002 0.062 0.004 U-238 57.9 0.058 0.860 0.0%6 Fission | fragments 181 0,172 b 46 0.068 Core Vessel SN 0.0k2 Blanket L - Th-232 30500 13 392 0.206 - Pa=233 - 5.5 0.0024 0.071 U-233 . B85 0.0037 0.110 Ii-Be-F. . e 0.010 Ieskage 0.004 Neutron Yield, 17 1.78 * Neutrons absorbed per neutron sbsorbed in U-235. - 39 - T&ble 1. 5 ' After Ogeration for 20 !ears uith- . Core- diameter.g 1oad fector: | - _Volume of external fuel system‘ U=-23%5 inventory: ' 0.8 - B-rt o - Power: 600 Mv: (heat) 870 kg 339 £t° Nuclear Characteristics of 8. Tygical,HbltenzFluoride-Salt Reactor QWS vBsmoval'of Fission Products - Feutron yield, n ............ Regeneration ratig: - 0_55_ - | . | " Newtron - Fraction Inventory Concentration Atom Density Absorption of (xg) (mole %) (A$cms/cm ) Retios* Fissions o x 1019 | | dore | ‘ , ‘ U-235 872 0.410 13,1 Fission - 0.407 0,550 n-y | | 0.182 U-233 312 0.152 4.85 | _ Fission 3 0.303% 0.410 n-)’ .. ) 0 0028 Pu-~-239 52,6 0.0k 0.778 Fission 0.03%0 0.040 n-7y | 0.022 Th-232 1.00 32,0 0.255 Pa-233% 7.32 0.0032 0.102 0,003 Ii-Be-F 0.073 U"23h‘ i 1201" 00058 1 87 00026 U-236 L8 0.210 6.72. - 0.1b47 Np-237 ; 0.015 0.471 0.019 U-23%8 -0.060 1.91 0.056 Fission , _ _ fragments 0.085 2.7T5 0.045 Core Vessel | 0.043 Blanket Th-232 30500 13 392 0.195 Pa-233 50 0.0021 0.064L5 - U-253% 33.0 0.01k0 0.hk22 - 1i-Be-F | : . | 0,009 1.8k * Neutrons_dbsofbéd per nefitron absorbed in U¥255¢‘ - o - begun immedistely and the thori\m is removed at the rate of 1/900 per day, in addition to the normal burnout at the rate of 1/4300 per day. The critical inventory rises within one month to & maximm of 626 kg and then falls to 590 kg at the end of eight months. At this time the re- processing rate is increased to 1/560 ver day, end the thorium is returned to the core. Thus, the thorivm concentration falls thereafter only by burnout. The regeneration retio ie little different from that of the previous case during the first two yeers, as indiceted by the dashed line ’ but it falls steadlly thereafter. The 112 3 inventory rises slowly » but the UE5 5 inventory 1s stebllized at 200 kg after about six years. The 025 > inventory could have been stebilized &t the twe-year value by modest withdrewals of thorium; however, the regeneration ratio would have fallen faster and additions of U23 2 to compensate for burnup would have been greater. A necessary condition for the feasibility of a molten-salt reactor 1s the integrity of the core vessel. This member is exposed to high- Intensity neutron and gama fields 'y end it is therefore subject to both . radiation damage and thermal stress. With a preliminary estimate of the hea.ting in a comparsble reactor having a pure nickel core vessela as a basis and with ellowence made for such differences as diemeter snd com- position of the fuel salt, the combined gamma and neutron heating in an 8-ft~die INOR-8 core vessel in a reactor heving 0.5 mole % ThF), in the - Tuel salt and operating et & power level of 600 Mw of heat was estimated to be not greatexr than 12 w/ m’ -of metal. The rate of heat relesase in the 'blanket sa.lt was es’cimated te e not gree.ter than 50 Mw, exclusive ‘ ’of any contribution from fissions in the blanket » vhich may add up to 'a.nother 30 Mw. N | BI..., G. Alexander and L. A Maxm, Firs’c Estimate of Genma, Hea.ting in the Oore Vessel of a Molten Fluoride Convertexr, ORNIL~CF S wl2=57s - 41 . st T. FPROCEDURE FOR PLANT STARTUP The initisl startup of the plent will be accomplished in four steps: (1)‘ prelimina.ry. checking of the systems, (2)_,prehea'_ting end £1lling of ~ the fluid circuits, (3) enriching to criticality, emd (4) operating et lowpower. The integ'ity and proper functioning of the equipment will be esta.blished insofar es possible, in the preliminary checking of the system, This step will include, in addition to cleaning and leek testing, checks of instrument snd alerm equipment settings end functloning, con- tinuity and polerity of the electrical circuits ’ direction of rotation of rotary élements , and opera.tion of vaelves and auxilia.:y gystems, In the second step, the fuel, blanket salt, and sodium circuits will be preheated to eabove the melting points of the various mediums ~and then filled. The preheating loads will be divided into menageable sections that can be sutomatically monitored for hot and cold spots o0 thet thermal stresgses may be minimized. The systems will be filled at temperatures as low as practical so that full advantage can be teken of fluid circulation as & means of bringing the system to an isothermal con~ dition before enrichment. During the initial period of fluld circuletion to establish the isothermal condition, the proper functioning of the flow control equipment will be established. Also the cleanliness end metal- lurglcal stability of the contaimment system will be evaluated by analyzing samples withdrawn from the fluld systems. The operability of the fuel system withdrawing and enriching equipment will be checked with barren . salt; the high-temperature instrumentation will be checked; the draeining and refilling procedures will be verified by testing; and remote mainten- ance techniques will be tried out. This period of nonnuclear isothermsl opera.tion at high temperature will also serve to familiarize the operating crews with the system and to establish their confidence in its operability. The preheating and fillihg procedures will begin with the introq.uc- tion of water to the steam generastors. The steem genmerators and the gsodium systems will then be preheated to 350°F, which will produce & - 42 - pressure of 150 psi in the steam system, The éodim will then be pres- purized from the sodium drain tenks into the primery end secondary sodium systems, the sodium pumps will be started, and flow will be established. Both sodium systems will thé_n be further heated to 600°F by using the electric heaters end by making use of the fluid circulstion. Simulta- neously the fuel and blenket circuits will be heated with electric heaters to the same temperature. The pressure in the steam generators will have risen to approximately 1800 psi end, before further system heating is ettempted, & small loed will be imposed on the steanm generators to hold the water temperaturé to 600°F as the rest of the circults are hested to bigher temperatures. The steam genera.tors‘ will be loaded by bypassing a small steam flow around the turbine, This load will be detexmined by the smount of :éxcess power avé.ilable during the heatirig period from pumping power end externsl heat sources in the systems. The load will be low | relative to the design cepacity of the steem generators; there will probably. ‘be less than 1 Mw available for bypess steam 'generation in the five units. At this jJuncture any increase in water temperature (above 600°F) would result in overpressurization of the steem system, and if the steam generators were allowed to evaporate to dryness end go to higher tempera- tures, severe thermsl shocks would be imposed on the structures. when water was again introduced. Theréfore, the sodium flow to the steam generators will be reduced as the reactor systems »a.re elevated in tem- : pera.ture. The flow in the secondary sodium loops will be reduced by | lcwering the pump speed, wh:l.ch in turn will reduce the flow to the steam o -'genera.tors ’ and the throttling valves will be mainpulated 4o refiuce the f*'prcportion of the total flmr through the genera.tors and to ‘shunt the flow | a.round the suPerheaters'_ .': o :" G _ When the rea.ctor a.nd ;pr:!.mary sodium circuits have been preheated | to 1100 Fy the sa.lts w:l.ll then be che.rged from the dmn_p ta.nks :Lnto the process circuitry, and flcw w:l.ll be established. o ' ‘ -l|.3a- At this time &ll the 'réa.cf.o;c ‘heat transfer loops will have been filled, end the rea.étozj will be operating isothermally at 1100°F, The steem gen- o erators will be running &t temperatures less than 625°F, and the super- . heater end reheater sodium éiré.uits_ will be running et temperatures less thsn 1100°F but greater than 625°F. The heat that is trensferred through the systems by virtue of the 475°F gredient will be durqped in the steem bypassing the turbine. This heet loed nmay. ve ira.riegi by éhanging the rate . of dumping of the stéem and the sodium flow rate in the steem generstors. When 1t hes been established that the plant is performing Se.tisf_a.q?- ; torily and that the systems are tight and chemically cleen, the cfitical experiment will be started. Fuel concentrate will be added to the resctor 'through the enrichment system. Approximately 38 ft3 of LiF-BeFE-UFu mixe ture cénta.ining 2.5 mole % UF), will have to be added to the 530 £ of carrier salt to achieve & fuel concentration of 0.15 mole % UF), » As the concentrate is added, 1t may be necessary to withdraw fluid from the fuel # system so that an adequate expension volume will be availsble in the | ' expansion tenk. The reactor will be titrated to criticality at llOOo‘F," and after criticality has been a.chievéd, & thorium~enriched salt will be added to the fuel mixture. This will drive the temperature down, end the oyerating temperature msy be finally trimmed by alternste additions of fuel end thorium concentrate mixtures. By edding the thorium to the - system as a last step, 1ts worth as e poison or chemical temperg‘.ture' sliim may be evalusted before power operation. " ¥ A period of low power operation will fallow the criticality experi- ment. By virtue of the negative temperature coefficient, the reactor will be a slave to the demand loa.d; which will be imposed by _increa.s,ing" _' the steam generation rate as & resulf of increasing the rate of sodium flow through the steem system. Manuel memipulation of system comtrol - velves will be required ;mt:!.l an epprecisble fraction of design power B is obtained, sey, 30%. The rate at which the load may be incressed will be determined bir the permissiblé rete of temperature change of the com- ponents;: T o — o | The turbine will be preheated by admitting steam thmugh ‘bhe turbine control velves, a.nd when the turbine has been hesated and brought up to speed, all the stemm will be directed 'bhreugh its normal path. At low .power levels, it may be necessary to attemper the stesm so thet the tur- bine temperature limits will not be exceeded Normal plent resterts aefter power operation will follow the same basic procedures, except that no criticael experiment will he required. Since there 1s no control rod, close attention will have to be paid to the fuel system filling rate and temperature so that nuclear trensients will not be incurred. 8., RBEACTOR CONTROL AND REFUELING The kinetics of circulating fuel reactors have been studied and reported in a number of papers.9 A typicel velue for the temperature coefficlent of reactivity for & molten-saelt reactor is «k x 10™7 (Ak/k)/o F. This negative temperature coefficlent is sufficient to make the power level in the reactor a slave to the applied load for all pormal. \ope‘re.- tional power demand changes, without the use of control rods, As indi- cated in the following section (Sec. 9), 1t keeps the reactor gafe from excessive temperature excursions even under some rather a.dverse conditions. The eritical temperature of the reactor graduelly deqreases during opera.tion at ;power 8s & result of the burnup of fuel and buildup of fis- ~ sion product poisens. | At conste.nt power, all tempera.tureg in the heat exchanger syst.ems decrea.se corresponding;ly, including, _in pa.rticular, the - : temperature of the sod:l.mn retu:ming from ‘the superheater-boiler-nrehea.ter systems This temperature must be maintained et ell times above an L 9W. K. Ergen, Current Status of the Theory of Rea.ctor Inmamics, L '—onm.-cr 55-7-157 (1§53); W. K. Brgen, "Kinetics of the Circulating Fuel - Muclear Reactor,' Phys. Rev., 25, 702 {June 1954); J. A. Nohel, Ste.bilit.z " of Solutions of the Reactor ‘Bouations, ORNL~CF 5%-9-25; Ws K. Ergen end - "~ A. M, Weinberg, "Some Aspects of Nonlinear Reactor Dynamics,“ Fhysics xx, k13 (1954); F. H, Brownell end W. K. Ergen, "A Theorem on Rearrangements and Its Application to Certain Delsy Differential Equetions,"” Jowgmal of Retional Mech. and Analysis, 3 565 (1954). ‘ B . - 45 - erbitrary minimm, determined by the melting point of the fuel, The tem- perature of the return sodium is. therefore used es en indicator of the | need for additional U25 > to restore the desired temperature level. The relation which gives the mass, AM, of fissionsble ma_.terial to be added to the reactor for & given increase in the steady-state mean core tempera.ture, Ty is given by the expression, - ¢AN[= ofM ZLT vhere, B = ____/ lkk and, ae —AE-I-{- /AT For epithermal reactors, p has values between 2 and 10, usually greater than 4, and can be obtained fram criticality experiments or by computew tion. The reduction in the coolant return temperature vs the time ree quired to burn up the corresponding mass (AM) of fuel, with constant power generation of 600 Mw, 1s shown in Fig. 1.1l. For exsmple, if the fuel inventory is 1000 kg of U23 2 , Bis 9, ails < x 1077 » @nd the sodium return temperature can be allowed to drop 50 ¥, the reactor must be re- fueled at intervels no greater then 13.5 days. On this schedule, the 023 5 addition required is 10.8 kg. The effect of bulldup of nuclear poisons 1s neglected in this calculation. In the flrst year of operation, the increased inventory required to compensate for the poilsons requires more frequent fuel additions. The calculation described gbove is typical cf the conditions that exist after fuel reprocessing is initiated. 9. ACCIDENTS: CONSEQUENCES, DETECTION, AND REQUIRED ACTION | The following diséussion gives the initial results of a study of - \' - difficulties that may arise as & result of accidental occurrences in | vaxrilous parts of the reactor system. Although no plausible accldents . with inherently disastrous results have been postulated, the need for - U6 - UNCL ASSIFIED (ge2 n 40 bY) dNNYUNB o s | = O - 5.: | simula’c.or etudy is shown in Fig. 1.12, This diagrem segregates one of 9,1, An Instentaneous loss of Loa.d From & Secondsxy Sodium 01rcuit T 10, further experimental end design efforts to determine the most economical way of handling some situations is epperent. | ‘ The tra.nsient beha.v:l.or of the remctor system has been snalyzed by enalog computer techn:l.g_ues for several types of sudden changes :Ln the : 5 hea.t load on the reactor.lo The reac'bor flow disgrem essumed for the | - the core heat transfer pa.ths for individual menipulation, s 80d lumps the | others ‘together into one hest sump. The tempera.ture coefficient of re~ activity essumed was a4 x 10%7 /°F. o This is the limiting cese of an accident occurring to only one of . . the core heat transfer paths et the maximm distsnce awsy from the reactor. \ All temperatwes upstream of the feilure tend {0 become isothermal at the new reactor outlet temperature, which is slightly lower then that under full power. 'I'he. temperature change in the plping end heat exchangers is rapid and emounts to 200°F or more. The heat exchangers, as designed, will withstand the tempersture changes, but a complete stress analysis of the plping layout should be made before such & reactor plent is built. 9.2. An Instantaneous Stoppage of Sodivm Flcw in One of the Primery Heet Eb:cha.ngers This case is similer to thet discussed above, except that tempera- tures downstream from the primary heat exchenger drop quickly to a lower isothermal temperature. During the trensient the mean core temperature of the reactor rises to & pesk of at most 20° F sbove normal during & period of time epproximately 10 sec sy while the outlet temperature d:rops auto- matica.lly to its new value. These two limiting cases ghow that there is no failure of & single heat transfer path that can cause an excessive tempemture rise in the reactor, ' E. R. Mann, privete commmication, ORNL. - 148 - -6{1- ¢ . o Pump PUMP TN ~ |rransport ’:‘ Lag T=L63 sec l1210°F 20°F | REACTOR / 600 Mw 1075 °F 925 °F . Transport] i el , Lag 3 can 3 40.2 ft>/sec. 13.4 ft7/sec. 46.! ft?/sec. T=1.7T5sec. {lIst Order Lag T=2.5sec.| 1080 °F 150 Mw HEAT SINK 825 °F Ist Order - Lag : 3 z= 2. 33.6 ft.°/sec. PUMP 2.5sec. : FIG. 1.12- REACTOR SIMULATION FLOW DIAGRAM 9.3; " An Instantaneous Reduction of the Heat Flow Rate from the fieactor Core If,a,fuel pfimp should suddenly stop, the rate of heat removal from the core would be quickly reduced to a fraction of that at\full power. The heat removedfwouldfbe determined by thermal insulation iosses, byvthermal convec- tion through the core fuel circuits, and by heat transfer to the blanket ~through the core vessel wall. The latter would be very significent if the blanket pump remeined operative. During the first few seconds following & ~ sudden fuel pump faillure, forced circulation in the cbre circuits would per- - gist és a resfilf of inertial effects but at a rapidly declining rate. The sharply decreased circulation rate would result in a iarger fraction of the delayed neutrons being reléased in the reactor core. A limiting apprOXimationkof the effects of fuel ?ump stopping was studied on the simulator. In the simulator studies, 1t was postulated that during steady-state full-power operation, the heat removal was reduced instantaneously to a small fraction of full power. It was further postulated that the flow of fuel stopped instantaneously so that the fuel salt that was in the reactor stayed there. The peak‘femfieratures which could be achieved if these condi- tions could be met and the times to reach them are given in Fig. 1l.13 as functions of the reduced heat removal rate. The temperature rise indicated results from the continued fission power generation at subcritical conditions from the gradual decay rate for the neutron flux end does not take into account afterheat from fission product radioactive decay, which acts as an additional heat source. Thé curves in Fig. 1.13 should be used with caution; they are intended " only to set upper limits on fihe temperature rise. The coasting effect from the fuel's inertia and thermal-convection circulation will reduce the peak temperatures markedly, but the relationships are complex and a more extended analysis is necessary. The pesk temperatures are not a problem of themselves, but their sudden appearance will cause thermal strains. The magnitudes'of these strains and their effects on the integrity of the reactor requires analysis, but no serious consequences are expected. .50 - o 9.k Cold Fuel Slugging | If cold fuel is suddenly injected into the reactor eore when the ~ power level is very 1ow, the core may'become supercritieal on & fast period¢ This can lead to the: poWer deneity exceeding the design level before the mean eore temperature rises again to ite nermal operating rangea Two gimulstor cases vere run te see whether these higher then design level power densities ceuld 1ead to & serious temperature over- shoot in the reactor._ One case eensidered involved suddenly 1ncreasing the power demand at the beiler from 6 Mw to 600 Mw. The immediete effect is to 1ewer the N core Inlet temperature to about 1000 F. The ayerage aore temperature 1s | reduced'te ebeut 1050 F, The tempeneturee then rise asymptoticelly to normal epereting levels with overshoot at moét of & few degrees. Ehe - power level overshoots to about 900 Mw, but the oversheot in power hes no practical significenee, A sudden load 1nerease at the boiler of this ‘magnitude is impractical to obtain, so that this is a limiting case in - BO fer ag a suddEH applieatien of load is concerned. It must be eencluded _ that "eold fuel slugging as & result of load.manipulation eannet lead to any diffieulty, A second caee was set’ up in en attempt to simlate. steppage ef fuel rflbw, cooling ef the fuel in the héat exchangers to Just above its melting ‘_fpeifit, and then starting flow to. put a slug of very cold fuel in the = reactor, 1In the starting condition of the simulator study, the reactor | Was syberitical at a temperature greater than 1200°F, As the floW'was -started end cold fuel was forced- into the reactor at the normal pumping rete, 2 stepawiee increase in the reactivity of O h% vas artificelly - inserted to place the reactor in & positive period. This insertion of_e_ positive period was intended to replace & condition of starting et very low power,vsince the scaling limits of the simulator do not permit the intradudtion of'initial'bcwer levels of lese than:B Mw. Under the simu- lator eonditiens used, the reactor core. temperature again dyopped abeut 150 F and then rose asymptotically to the design tempereture with no . perceptible overshoot. It is concluded that a fuel pump starting up with cold fuel in the primary heat exchangers is unlikely to lead to high temperature excursions in the reactor. - 9. S. Rem0va1-of Afte;'heat by Thermal Convection A survey 'éxamifiation of the capability of ‘the heat transfer system for the removal of heat by thermal convection in the event that all | pumping power is lost has been made, The temperature pattern of the system reéluired for the removal of 4% of the design power by thexmal . - convection is shown in 'Fig..‘ 1.14, This study shows in a preliminary way - that thermal convection can remove enough heat from the reactor core so that loss of power to the pumps in the redioactive areas will not neces- sitate the drainege of the fuel from the reacétor. A detalled system enslysis may indicate thet slight modifications in leyout mey b: required to accomplish this, however. | 9.6, Iloss of Fuel Pump Any evefit which stops the forcéd ‘circulation of“fue}. shrough the - primary heat excha.ngers req,uires tha.t steps be ta.ken to prevent freezing of the fuel salt. The steam system is such a large, relatively low teme perature heat sink that the fuel salt would be quickly frozer. if no action were taken, There are two safety controls. First, fuel pump sfio;ppa.ge or loss of power will cut the steam tb thé ‘fiurbine » 8nd reduce the turbine output to a low level to handle ai‘terhea:b, etc. The second control, ,.triggered by a low tempera.ture :I.n 'bhe cold line of the primaxry sodium, : ;will stop 'bhe sodium pmn;ps. 9.7. Loss of Elec'bric Transanission Line comection to 'l'.he Plant In the even‘o of 1oss of electrica.l load on 'bhe pLa.nt s the turbine " ‘,'”__-:-stop-valve wi].‘!. ad,just automatically *bo prevent turbine runa.way The N "tu:r'bine ca.n be adJusted to & 5 to 10% re:bed loadll to supply lbca.l needs. '.'-E[he r:I.se in pressure in the steam system resulting from closure of the - -stop-valve will o;pen the emergency reliei’ va.lve until, to save purified x> llH L. Fa.]kenberry, TVA, pri'vate comunication. T EM, A - 54 - £ PATTERN AT 4 7% HEAT - water, the steam'bxpass valve is adjusted for the dumping of steam to the condenser._ Simultaneously, or as soon as practiceal, secondary sodium pump speeds will be reduced and valves controlling sodium flow to the ‘ boller, superheeter, and reheater and in the bypase_will adjust antomatig . cally to glve design temperatures and pressures for the smount of after- hesating being removed from the reactor. If this idling power exceeds the power required to operate the plant it will be dumped to the condenser.‘- The whole plant will be maintained in & condition ready 1o resume its electrical loed as soon as it can be re-established | It is to be noted that‘should the load loss-bewsufficiently prolonged so that the 'afterheet 1s not sufficient to provide power for local needs, the core will generate fission heat automatically. In order to maintain the power station:in a standby condition during a period in which, say, the elec- tric generator equipment is inoperative and there is a simulteneous loss of pover to the plant, emergency power generation equipment will be needed. The emergency supply must have sufficlent capacity to operate instruments, controls, feedwater punps end auxiliary equipment necessary for control end removal of afterheat. 9.8, -leak Between Fuel and Blanket Salts ~ The free surface of the blanket salt 1is above the free surfece of the fuel salt, and the blanket selt is more dense than the fuel salt, Both the core and the blanket will have e common ges pressure over them, . and both are on thejsucfiion'side of'the:pumps in their respective systems. _,Under these conditions the blanket will always be at & higher static pres- :sure than the’ core, and any 1eak between the fuel end blanket salts will | drive the blanket salt into the fuel salt and lower the critical temperan .ture of the reactor core. gj< With such a leak fihe maintenance of ayatem.tem@eratures would re- quire addition of fuel at a-rate in excess of that required for burnup end’ fission product poisoning.. To mainxein the critical temperature | | constant in a clean, 8-ft-die core with a fuel selt . contadning 0.75 mole % thorium, ebout one atom of U256 must be added for three atoms of thorium - 55 = that lesk into ‘hl*e core from the blenket. Thus, 1f the fuel accounta- bility is sufficiyently sensitive to detect a 10% excess fueling rate, inleskage to the core of more then 165 cm? per day will be detected. The fuel end blanlcet salts are chemically inert with respect to each - other, and therefore no chemical effects of the mixing are ex_pected. | if fission-product and heavy-element poisoning were to mask 'bhe excess refueling caused by a blanket leak and prevent early detection, the lesk would be detected eventual]y by cerresponding chenges in fuel end blanket inventories s 85 indicated by the level ind.:l.cators of the respective systems. ~ Once a lesk between the core and blanket was detected, the reactor would be shut down and all liquid systems would be .drained. Replacement " of the reactor vessel would be required, and this would be & 1engbhy operation, Complete rupture of the core vessel would lead e.utomatica:l.]y to & suberitical cOndition. The core surge tank would £111 as the bls.nket end core pressures tended to equalize. 9.9. Ieak Between Fuel and Sodium The relative pressures in the fuel and sodium systems will always be such that, in the event of a leak between the fuel and the sodium, the fuel will enter the sodium stream. This errengement is used because | the consequences of precipitation of urenium in the cir’cula.fiing fuel system cannot be predicted with certainty. The chemical cbnsequences of a leek of the fuel into sodium, such as could occur in a primary hest excha.nger, have been examined on the basis of thermodynamic data.la When the fuel is mixed with excess sodium, the major constituents, except IiF, will be promptly and simultemeously - réduced to their metallic states, according to the reactions: 12W.- R. Grimes, private eommication, ORNL. - 56 - v C . (1) -UFll_+Na.$NaF+UF3 AF = =36 kcal (2) -UF5 + 3Na = U + 3NaF AF = =37.3 keal (3) BeF, + 2Na =Be + 2NaF ~ AF = =29 kcal (4) ThFy, + 4YNa<==Th + 4kNaF AF = -hk keal Of the fission products contained in the fuel, the alkaline earths, the rare earths, end a considerable fraction of the alkali metals will remain in the salt phase as fluorides, while Mo, Ru, Zr, cd, Zn, Sb and Sn will be reduced to the metallic state. The anions, particularly I]'5 T and Br87, which are important for neutron detection because they are long-lived precursors of deleyed neu-. tron emitters, will appear es halide ions. Accordingly, the salt mixture » after reaction, will contain about 5% mole % NaF, 46 mole % IiF, and traces (insofar as concentration is concerned) of fission product fluorides. Such & mixture will have a melting point greater, probably, then 700 C. (1300°F) » and accordingly will be solid at the normal temperatures in the sodium circuilt. - Metals such as Sn, Sb, Cd, Rb, Cs, and Zn should be soluble in molten sodium, but all other materials introduced into the sodium by fihe fuel are moderately high melting and will be sparingly soluble in the sodium. Beryllium metal, which is present in relatively large concentrations a.nd which appears to be relatively insoluble (< 100 ppm) in sodium, will | probably be the first material precipiteted. - Sodium fluoride probs.bly dissolves to the extent of 0.2 mole % in sodium at llOOoF, and this selt, along with LiF, will exceed the solubility :l.n molten sodium and exist as. separate solid pha.ses after relatively small quan'b:l.ties of fuel have lea.ked into the sodium. '_' - o : ' o Sodium iodide and sodium bromide a.re more soluble tha.n sodium fluo- ride in molten sodium, and these precu:rsors of the delayed neutron enitiers will, accordingly, be dissolved in the molten sodium until the NaF-LiF mixture saturates the sodium end forms a seoond phase, - Since they are ‘-'57_- more soluble in the salt phase than in the liquid metal' they could, in , principle » then decrease in concentration in the molten metal due to their | extraction into the solid salt phase., This extraction process is not ~ expected to be importent, however, since ‘the emount of solid salt phase - will be small for a considersble.period, end extraction by a solid from a liquid should be relatively slow. | M'bhemore s precursors of delsyed neutrons present in the sodium a.rise only from the :t‘reshly leaked-in fuel, and therefore the concentration of precursors wil_'l. not be appreciably . _ affected even though the totel atomic: species concentration may be dimin ished by extractions. ' Prompt detection of small fuel leaks into the primary sodium clr- cuit poses a problem yet to be solved. Al cm3 /day leak will produce approximately 0.5 n/ cm e geC by precursor decay at the secondary heat exchanger, but it is doubtful that neutrons of such a source strength | can be detected in the sodium cell. Likewise the neutron actiiration of the primary sodium produces gama;ray activity which would tend to mask fission fragnenn gamma activity. Detection of large leaks would ‘oe ‘aided by comparison with the activity in the other similar secondary sodium circuits, but the determination of the size of lesk that can be detected has not as yet been mpde. . . A welleagitated stoichiometric mixture of sodium and fuel salt will result in & rapid temperature rise, estimated to be 1200°F under -adiabatic conditions. It is difficult ’ however » for such conditions to ex_isn -in - a practicel situation. Heat evolved from a small leak would be rapidly carried away by excess sodium. For the larger leaks which could occur from fatigue in bending or 'bension, the solids formed would in‘berfere . with repid mixing. Some work has been done with a NaF-ZrFu base fuel and NeX which demonstrated this smothering effect; further engineering tests - 'will ‘be required to demonstrate safety with sodium and the present fuel - salt in simuleted component equipment. , ..5_8_.. As soon as & fuel to sodlum leak had been detected and the feulty heat exchanger had thus been located, the reactor plant would be shut down, the fuel and appropriate sodium circuits drained, and the heat exchange:r replaced. '9.10. _Leak of Fuel or Blamket galt to Reactor Cell The presence of a small lesk from the fuel to the rebetor cell can be defiected"by gas=sampling techniqfies. Its léca:l:ion will be more dif- ficult to determine, 'The repair of such a leak would of course require draining the fuel salt. The provision of an inert atmosphere in the reactor cell will prevent rapid growth of lea.ks ca:used by salt-fluxed oxidation. ‘ A gross leak or ruptu:re of either the fuel or bla.nke'b circuits is a major accident. Means must be provided to vent reactor cell pressure as it is bullt up by heat release from the spilled galt , and & suitable noncritical emergency drain system that can handle afterheat on & one- time basis must be availasble. Ways are known for doing both, but the lowest cost way of accomplishing these disaster pmmw.tdéva a'teps. has not. yet been determined. 9,11, Ieaks of Water or Steam to Sod.ium The sodium in thermal contact with the water or steam is nonradio=- active. The problem of leaks peWeen the water and sodium systems has been faced by those engaged in the development of fast reactors, end thelr studies and test results will be useful in determining heet ex- changer desigl. 10. CHEMICAL PROCESSING AND FUEL GYCLE ECONOMECS 10.1. Fuel Salt Reprocessing The system for chemical reprocessing of the fuel salt is e cambina- tion of the ORNL fluoride volatility and the Kfi25 uranium.hexsfluoride reduction processes. described in Part 6. 'The salt to be reprocessed is / | transferred as described in Section 3, above, from the reactor circuit %0 & holdup vessel on & convenient schednle, such as 2 ft3 once eaeh day or 12 ft3 once each week. The holdup vessels provide containment during ~ the holdup period reqnired for decay of the short-lived activities and act as a buffer between the reactor end the chemical plant so that the operation of the reactor need not depend on the state of repair of the chemical plant. The fuel salt will be fluorinated in batches of 2 ft3 each, one batch per day. After the uranium is removed by fluorination and collected as'UFé on NaF pellet beds, the barren salt is transferred to waste storage. The UFB'Will be discharged on a twice~per-week cycle from the NaF pellet beds, which have a capacity of 10 kg of uranium. The volatility process produces liquid UEB, in cylinders, which is subsequently fed to a reduc~ tion tower to produce UF), which is combined with fresh salt for return to the reactor. The uranium losses in the chemical processing are sbout 0.1%, i.e., about 1 kg/year. 10.2. Blanket Salt Reprocessing - Chemical processing of the blanket salt is physically much the same as the processing of the fuel salt except that, after fluorination, the salt is returned to the blanket system. Because of the much lower power - density in the blanket salt, holdup for decay-cooling is not & problem. ~ Separate fluorinators for fuel and blanket salts, to prevent cross- contemination, are assumed, as are separate NaF beds, to make possible e the withdrawel of pure 0235 from the system i1f desired. The seme UFB reduction tower will serve both fuel and blanket salt processing. - 60 - et b NS i s A blanket selt processing rate that is about‘ the same as that for the fuel selt is assumed, i.e., one 2 ft3 batch per day. Thus fluorina- tion equipment of the same size will suffice. The uranium throughput ‘rate of the blanket salt processing system is , however, only sbout 10% of that of the fuel salt. For convenience s the same size of NaF bed is proposed for the two systems, elthough this meens that the UF; will be discharged from the NaF bed in the blanket salt system only once every other month, The blanket salt processing rate is sufficiently fast to hold the . U25 5 inventory in the blenket salt system to about 60 kg and to limit the fissioning in the blanket salt to about 3% of the total. 10.3. Cost Bases Fissioneble isotopes have been velued at $17/g in computing inven- tory and burnup charges and breeding end resale credits. Capltalization rates were assumed to be 4%/yr on fissioneble materials and 14%/yr on everything else. The fuel salt was estimated to cost :';5].2'78/:13‘1;5 end the - vlenket salt $2517/ft”, The varisble cost of fuel selt chemical process sing is assumed to be equal +to the cost of buying new salt to replace that process'ed.l The blenket salt is used over the life of the reactor without excessive fisslon product buildup. The fissionable materiel comsumption cost is based on feeding 93% enriched 112 3 5 to the -core system to compensate for & regeneration ratio of less tha.n unity. It is assumed that lf? 35 :Ls not availe.ble for purchese - E=h m ecms\mic pz‘iue, a.l'bhough it would 'be worth a.pproximately twice &s o .much as 025 5. 1n a.n mtemediate-neutron-energy mol'ben-sa.lt reactor due L to its higher regeneration ra.tio et 1orwer cri'bical inventories. It is o a.ssmned a.lso that isoto;pic re-enrichment of Ua 32 or 023 2 .elther by | ga.seous diffusion or by excha.nge w:l.th & price penalty, 1s not econcmice.l ~ BO tha.t the molten-sa.lt power reactor must tolere.te the nonfissiona.ble | 'uranimn isotopes a.nd 'c.he resulting 1ower regeneration rat:l.o and higher U2 3 3—1125 2 inventory - 6l = 10.4, Chemical Plant Capital Costs The budgeted capital costs for the QRNL volatility pilot plant total about $l,500 000 through fiscal 1959, This figure includes replacements and.modifications, which should not be required in & second plent, end 1t also includes solid fuel element handling end dissolution facilities, vhich would not be required in the molten-salt reactor plant. On the other hand, the $1,300,000 does not include building and service facili- ties, or any eqpipment for reducing UFg to UFh and, reconstituting fuel salt. Aflditlons and subtractions considered the reference design . chemical plant equipment and installation cost 1s. estimated to be $l,500 000, The chemical plant's share of the total reactor capital investment is _about twice this amount, when charges for bullding end site, design,_general~ expense, and contingencies are added. These capital costs are listed with other capital costs in Section 11. 10.5. Chemical Plant Operating Costs The ORNL volatility plant operating budget for three fiscal years (1957-58-59) totals $1,368,000. The molten-salt reactor chemical plant would have lower "unusual” costs (associated with development) than the pilot plant, but higher "production-proportional” coSts, apd 1s estimated to cost $500,000 per year to operate. To this must be added the cost of replacing the fuel salt processed, or reclaiming it, if this can be done for au equal or lesser cost. This is estimated to be 600 ft3 (ap-' proximately one fuel system volume) per year at $1278 per ft5, a total of $770,000 per year. A salt reclamation process might be expected to reduce this considerably, although probably not more than by a factor of 2, which nevertheless would save sbout 0.2 mills/kwh. The chemical plant operating costs are listed with other operating costs in Section 11. ' 110.6, Net Fuel Cycle Cost For the purpose of estimating fuel cycle costs, values averaged over the reactor lifetime of 1000 kg for the U253 U235 inventory and 0.5 for thereffective breeding ratio were assumed. The nuclear heat power was -62- teken to be 640 My, the net electrical output 260 Mw, and the load factor 0.80. The net fuel cycle cost is estimated to be ebout 2 mills/kwh: Ttem T - mills/kovh ¥ consumed o 2,260,000 S0 1.2k Fuel salt make@ | 770,000 0.k42 R inventory 680,000 0,37 . 2,03 If‘ & comparison of total fuel cycle costs with thoée for a solid fuel element power reactor are to be made , the chemical pla.nt capita.l cost end the chemica.l plant oPerating cost should be added. These amounts are as follows: $/yr | mills/kwh Capital cost ($3,000,000) 420,000 - 0.23 Operating cost | 500,000 0.28 They lead to a total mél cycle 'cost 61’ 2. Sy-mills/kwh. , 11. CONSTRUCTION AND POWER COSTS 11.1. Capitel Costs The information available in the preliminary design does not lend itself to & rigorous cost ana.]ysis 5 hcwever, the power cycle has been: | -sufficiently well-defined 'to pemit a segrega.tion of the major compoa . 'knents in the plant.. The plant layout has progressed to the extent that ~ the over-a.ll size may be detemined. 5 | Des:l.gn studies of some of 'bhe fuel system auxiliaries have pemitted a detailed cost brea.kdown The high-tempera’cure sodium pump | requirement.s have been ascertained to the exbent tha.t manu:facturers of this equipment *Vhave been a'ble to ma.ke prelimine.ry cost estimates. 'I'he fuel and blanket . salt pumps were estima.ted by sca.lingup costs ‘of smaller pms that have been febricated and tested at ORNIL for high-temperature reactor systems. - 63 = ORNL's experience in molten~salt and alkali metal heat transfer equipment fabrication and procurement has been drewn on to estimate the increased cost in producing reector-quality products. In some cases individuel ccmponents were found to be too mumerous for detailed cost analysis in the time availsble, and costs were essigned to entire suhsystems on the bésis of general experience. The instrumen- ,tetion, electrical equipment and auxiliary systems were treated in this menner, for exemple. | ' - It has been assumed that the molten-salt reactor plent would be con- -structed et a site similar to the one selected in a recent ORNL ges=cooled reactor study. 3 Therefore, site ecquisition, imprevement, and structure costs have been set at comperable levels, The capital cost summary is presented in Teble 1.6. It should be noted that a 40% contingency factor has been epplied to the reactor por- tion of the system. It is felt that there are a'rumber of uncertainties in some of the larger reactor cost'packeges and a contingency factor of this order is warranted. A T7.5% contingency factor was applied to the remainder of the direct costs. | The general expense or indirect costs charged to the plant represent administrative, personnel, plant protection, safety and special construc- tlon services which are largely incurred during comstruction and startup 0peraticns._ The design cost represents approximately 5% of the direct . cost subtotal before the contingency factors were applied. This capital cost sumary leads to & cost of $269 per installed killowatt of generating capacity. | : . ¥ 4 Table 1.7 presents a more detailed cost breakdown of the reactor ‘portion of the plemt. The major-emthents or items have been.listed . end the materials of construction for a particular liquid system have been indicated. 0me ORNI Gas-Cooled Reactor, ORNL-2500, Part 3 (April 1, 1958). -6l - O 10. 11. -13A, 13B. 1k, 15, 16, 18. ‘Teble 1.6, Capital Costs (FPC Account Numbers) Lend and land rights Structures and im@rovements Reactor system (1ncluding chemical plant) Steam system Turbine-generator plent Accessory électrical\equipment , Miscellaneous power plant equipment Direct costs subtotal T.5% contingency on 11,13B,1k4,15,16 409 contingenqy on 13A Contingency subtotal ‘General expense Design costs | TOTAL COST - 65 = $ 500,000 7,500,000 20,232,000 3,750,000 11,750,000 4,600,000 1,250,000 49,582,000 2,201,000 - 8,093,000 10,294,000 7,500,000 2,450,000 - $69,826,000 I.. | II. III. ‘Table 1. 7. Reactor System Capital Cost Summary - (Section 13A of Capitel Casts) 'Fuel System (INOR-8) Reactor core and blanket shell . $ 500,000 A, B. One 24,000-gmm purp, pump shielding, - 81+5,ooo - &nd motor : ) C. Four fuel-to-sodium heat exchangers - 672,000 D, System piping | | 100,000 E. Main fill-ond«drein system . _ 520,000 F. Off-gas system (includes blanket system) ‘- 568,000 G.' Enriching and withdrawal system | 100,000 exclusive of chemical plant B ' ;_ - Ho Preheating end insulation ' S 75,000 Blanket Circuit (INOR3S) | A. One pump and motor | 350,000 B. - One blanket salt~towsodium hea.t exchanger ’ C. System piping - 20,000 D. Msin fill-snd-drain system | 120,000 E. Enriching and withdrawal system 50,000 | exclusive of chemicel plant | . Preheating end insuletion - | 15,000 Intermediate Sodium System (stainless steel) (4 fuel and 1 blanket circuits) A. B. Fuel~toasodium systems: 1, four 20,000-gpm pumps end motors 960,000 2. four sodium-to-sodium heat exchangers hlS ,000 3. system piping \ 300,000 Lk, drain systems | 100,000 5. preheating and insulation 75,000 Blanket salteto-sodium system: ' 1. one 10,000~-gpm pump end motor 130,000 2. one sodimn-to-ssodium heat exchenger 45,000 5. system piping . 75,000 Lk, drein system | 30,000 5. preheating and insuletion . 20,000 =66 - 3,380,000 651,000 2,150,000 VI. VII. VIII. X XII. IV. Table 1.7. (Continued) Secondary Sodium Circuits (Cr-Mo alloy -steel) A. Fuel=to=-sodium-to-sodium systems: 1. four 15,000-gpm pumps and 2-speed drives 2. four sodium-to-water bollers 3. four sodium~to-steam superheaters i, four sodium-to-stesm reheat exchangers 5. twenty remotely~operated throttling valves 6. system piping - 7. fill-ande-drsein systems 8. hesting and insulation B. Blanket salt~to-sodium-to-sodium system: I. one 10,000~-gpm pump and motor 2, one sodium~to-water boiler 3« oOne sodium~to~stesm superheater 4, four throttling valves 5. system piping 6. fill-and-drain system T. heating and insuletion C. Sodium emergency drain system Reactor Plant Shielding (17,000 cu yd of concrete at $100/yd) Mein Conteinment Vessel, Air Lock, Reactor Support, and Cell cooling System Instrumentation _ Remote Maintenance and Hendling Equipment Auxiliary Systems (helium, nitrogen, cranes, cooling systems) SPare Parts: - Pumps - ','B‘ ‘Heat exchengers tli" Miscellaneous ' 7 Original Inventories-',,‘ A Sodium (300,000 1b x_ $0. 20/lb) ' B. Blanket salt (750 £t7 x 3.2 x 2517/1'155) - €. Fuel salt (575 3 x 1.2 x 1278/ft.5) ” Ghemical Plant Equipment N 1,000,000 336,000 252,000 380000' 350,000 175,000 150,000 145,000 40,000 38,000 80,000 50,000 40,000 200,000 3,626,000 1,700,000 450,000 750,000 1,000,000 525,000 700,000 400,000 200,000 60,000 2, 260 000 880 000 1, 500,000 $20,232,000 e e — 1l.2. Power Costs Power costs hawé been divided into three categories. These are: fixed costs, operation and maintenance costs, and fuel cycle costs. The fixed coste are the cherges resulting from the cepital investment in the plent, This emount has been set at 144 per annum of the invest- ment, which includes taxes, insurance, and financing charges. This leads to an ennuel cherge of $9,776,000 or 5.37 mills/kvh. | - The operation and maintenance costs are, in the main, dependent on the ultimate relisbility of the reactor portion of the plant. The de- velopment of practicel remote-maintenance techniques for the repair and replacement of equipment in the redioactive systems'is also vitel to | . assure reasoneble costs. No accurate determination of such costs cen 1 be made without further experience. For the purpose of this report the operation and maintenance cost breskdown given below has been assumed: | | Amnugl Charge - Labor and supervision $ 900,000 Reactor system spare parts | Pumps 250,000 Heat exchangers ' 200,000 Miscellaneous 300,000 Remote-hendling equipment | 150,000 Chemical plant operation 500,000 Conventional supplies 400,000 | Total $2,700,000 This total cost results in an incremental power cost of 1.48 mills/kwh. Net fuel cycle costs as discussed in Sec. 10 above amount to 2,03 mills/kwh. The three categories add up &s follows: - Annusl Cherge - mills/kovh Fixed cost . $ 9,766,000 537 Operating and meintenance 2,700,000 - 1.48 Fuel charges 3,710,000 2.03 Total annuel charge $16,176,000 Total power cost \ 8.88 The difference in cost between having the :'eactor plent on standby and heving 1t on the line is about 2 mills/kwh. | 12, SOME ALTERNATES TO THE PROPOSED DESIGN 12.1, Alternate Heat Trensfer Systems The possibility of replacing the fuel-toésodim-to-sodium-to-steam heat trensfer system with a fuel-*bo-gas-to-steém ‘sy'stem has been given a cursory exsmination. The attractiveness of such & system' is based on the replacement of two sodium systems in series with one gas system and in heving the gas chemically competible with both the fuel and weter or steam. Early estimates of the gas heat transfer performence indicate that the fuel volume required to transfer an equivalent quentity of heat would not be appreciably different from that requii-ed with the sodium system. These estimates were based on use of & return gas temperature below the melting point of the fuel, and the sa.fety of this procedure " must be examined further. If the gas system were operated in the 300~ to 1!-00-1)51 range, the power required to circulate the gas eould:be.kept.at a reasonable level. - The gas system would permit & reduction in the nunber of heat ex- j cha.ngers a.nd pumps, s.nd eliminate sodium va.lves. The dew point of the gas would provide & rapid method of leak deteetion in cese of e steam- to~gas lee.k A steam leak into 'bhe gas system would no-b heve the chemical hazerd that exists with a steam-to-sodium leak. If the reactor were - 69 - operated inside a pressure shell, the fuel pressi;re could be maintained slightly below the gas pressure to ensure that any leeks in the fuel system would be inwerd. The small pressure differential required between the fuel and the gas would permit the maximum fuel (gage) pressure to be maintained at a level no higher than that required by the liquid-cooled system. ‘ Gas _eooling would eliminate the need for the sodium handling eystems with their sttendent preheating problems. These sodium facilities would be replaced with gas storage and hendling equipment, Startup and shute down procedures would be simpler with ges then with liquid cooling, par- ticularly with respect to preheating end pert losd control, The amount of secondary rediation shielding required with the liquid system would. be considersbly reduced with the gas system because of the decreased in- duced activity of the coolant. | More studies of the gas cooling system are being made, and it is spparent that the bulkiness of the gas system will present handice.ps; . It 1s also probeble that a gas cooling system will prove more exjgensive. A better comparison of the gas=cooled system with the sodium-cooled system will result from a more detailed design study. An elternate o the steam—cycle described above would be the Loeffler boiler cycle. In this system (see Fig. 1.15) all the heat is transferred to the steam in the superheater. A portion of the superheated steam is recirculated by means of a steam pump to the boj.le_r,' where it transfers heat to the water by direct contact to form saturated. steam. With the seme steam conditions of 1000°F and 1800 psi, it would be neces- sary to return approximately 2 1b of steam to the boiler for every pound sent to the turbine. The advanta.ges of this system used in conjunction with the molten-salt reactor are principally connected with the control of heat flow. :I.'bh the sodimn—to-steam generator heat transfer system, the steam genera.tor represents a 1arge ca.pacity heat sink &t a tempera- ture more than 200°F below the freezing peint of the fuel bearing salt. | To prevent freezing of the fuel, careful con’crol of flow in the seconde.ry u70-i: U e NL- L'P Dsiz,ezoqzoz ~LF E/P TURBINES o REHEAT | BOILER oy LIL X FROM FUEL FROM BLANKET N&r‘—w P TURBINE i ~— AT TEMIFERATOR N EMERGENCY RELIEF e COVING WATER ~ CONOENSER ~—LDf-OUPERHEATER ~—FLOUCING STATION 1 ! 1 Y ary SUPER HEAT EXCH, o HEATER C\-conoevsare PumiPp 1£ ~SODION OFF INERT ST o FEEDWATER HEATERS W SERIES . -DEAERATOR q ~FEEDWATER PUAMP p—— STEMNT PUIIE ~—f 7O BLAONKET L —-{} TO FUEL FEEOWRATER —. HEATERS /N SERIES Fig. 1.15. MOSPR WITH LOEFFLER STEAM SYSTEM FLOW OWIGRANM sodium circult must be mainteined at low'pcwer operation. In the loeffler boller eystem, the directly coupled heat sink is dry steem in- ‘stead of water, end control of the steam circuletion is believed to be | easier then the control of the sodium flow. The elimination of the need for two speed pumps and control valves in the intermedisate circuit would therefore result in a system that would more fully exploit the inherent self-regulation of the molten-salt resctor. ” The elimination of the steam generator from & sodium circuit reduces the need for sodium flow regulation. The reduction bf sodium eguipment probebly would result in less frequent meintenance, Thus some of the serious objections to having radiocactive sodium heating the steem would thereby be leSSened and consideration could be given to the elimination of one of the intermediate circuits. In addition, since minimm tempere- tures would, &t design point operation, be sbove the melting point of sultable fluoride salts, their use in place of sodium should be exemined. If substitution could be made, chemical compatibility of the intermediate - fluid would be markedly improved both with respect‘td the fuel and the steam and these hazards would be lessened. More detailed design comparisons will be necessary to evaluate,this boiler system. Although the changes suggested above are plausible, the detailed consequences must be snalyzed and felr cost comparisons made. | 12.2. Alternate Fuels The substitution of U235 for U255 in the molten fluoride reactors would result in substantial improvement in performance. Uranium-233 is & superior fuel in almost every respect. The fission cross section in the intermediate range of neutron energies is greater than the fission cross sectlons of 0235 and Pu?59. Thus, initiel inventories are less, - and less additional fuel is reqfiiréd to over-ride poisons. Alsc, the n=-y cross section is substantially less, and the radiative capture results in the immediate formation of a fertile isotope, P, e rate of accumulation of 0256 is orders of magnitude smaller then with fl255 fuel, end the buildup of Np>2! and Pu->? is negligible. —72.-" O Prelimina.ry and iné.omplete results from e parametric study of re- - ac‘bors fueled with 0235 exe given in Part 4, Sec. 1.2, In a ty;pica.l case in which the core diameter wes 8 £t and the aoncentration of ThF) - wes 1.0 mole %, the initisl critical ma.ss was found to be on.'ly 87 kg of ‘ 0235 the inventory for a 600-Mw system was only 196 kg, the regenera.tion reatio vas 0. 91, end the long—term performance was good. In another Befte dle core system with 0,75 mole % ThF) in the fuel selt, the initial in- ventory was 129 kg, and the conversion ratio was 0.82. After operation for one yesr at & load factor of 0.8 end with no reprocessing of the core to remove fission products, the inventory rose of 199 kg, and the regene eration ratio fell to 0.7l. However, if the reprocessing required to hold the concentration of fission products constant wes stnrbed efter 1 year of operation, the inventory increased slowly up to only 247 kg &f'ter 19 years and. the regeneration ratio TOSE sl:l.gh*tl;y to 0.73. Roughly spea.k,ing, the eritiecsl :anentories regquired for the 0235 systems ere about one-third those for the corresponding 1)23 2 systems, eand the burnu,p .req_uirements are a.bout helf. The ebove described case is not optimized for U255 __ Substa.ntial - improvement can be obtained by using higher concentra‘oions of thorium and correctly ma.tching the diemeter end processing rates. -~ As discussed in Part 2, Pu:l?‘5 has eppreclable solubility in mixtures of LiF end BeFpe It should be possible to maintein concentrations of up to 0. 2 mole % sa.fely. - 'I'his is more than ample for clea.n systems having dlemeters in the range from 6 to 10 £t with no thoriwm in the core. A typical 8-ft-dia. core would have a. critical concentra.tion of 0,013 mole % PuF; end e regeneration ratio- ('l‘hFh in the blanket) of a.bout 0.35. 240 - The ei‘f’ect of accumulation of fission products and Pu on the cr:l.tica.l concentration a.nd 'bhe effect of ra.re earbh fission products on. the solu- bility of PuF. rema.in to be <’ie‘t'..srm:l.ne=.d° It does -eppear proba,ble, however, 3. %hat e molten fluoride plubonium burner ha.ving unlimited burnup end ex.hibit:l.ng substantial regeneration in -bhe blanket :I.s techn:l.ce.lly feasible, - T5 - ‘of pentavalent uranium (UF., U2F9,', etc.) are not themmally stable PART 2 - CHEMICAL ASPECTS OF MOLTEN-FLUORIDE-SALT REACTOR FUELS 1. CHOICE OF FUEL COMPOSITION The search for a liquid for use at high temperatures and low ;preé- sures in & fluld-fueled reactor led to the choice of elther fluorides or chlorides because of the requirements of radistion stebility end solu~ bility of epprecieble guentities of uranium and thorium, . The chlorides (based on the CJ.3 T isotope) are most suitable for fast rea.ct-o:r use, but the low thermal-neutron ebsorption cross section of £luorine _mak_es ‘the fluorides seem to be & wniquely desireble choice for & high-temperature £1uld-fucled reactor in the thermal- or epithermal-newtron reglon, 1.1, Choice of Active Fluoride Urenium Fluoride. Uranium hexafluoride is 8 highly volat:l.le com~ pound, and it is obviously unsuiteble as e componént of & liguid for use at high temperatures. Thae compound UOLF, o whie.h ie relatively nonvolstile, 1s a strong oxidant that would be very difficult to contain. Fluorides 1 ena. would be prohibitively strong oxldants even if they could be stebilized in solution. Uranivm trifluoride, when pure and under an lnert amosphere ’ is steble even &t temperatures above 1000°C; 293 however, it is not so steble in molten fluoride solutions.,h It disproportiona tes apyrecie.bly in such media by the reaction, ‘ ' ) Y uUF3‘———°-—3 UF, + U 1. 7. Kotz end E. Rebinowltch, The Chemistry of Ura.nimn ms-vm:f-s, . MeGraw-H11l, 1951. ®roid. | 3¢, 3. Barton, W. C Wnltley, E. E. Ke’cchen, L. G. Overholser, and " W. R. Grimes, Preparation and ProPerties of UF5, Oak Ridge National _ La.boratory (vnpublished). l*Se-e Reactor Hendbook, in press; material submitted by B. H. Clampitt 3 5. Lenger, end . F. lankenship, Ock Ridge Netionsl Leboretory. S Th - et b e ot at temperatures below 800°c, Emall emounts of UF5 are p‘emiesible in the presence of rela,tively le.rge concentramions of UFL and may be bene- fic:l.el insofer as corrosion is concerned, It is necessary, hewever, to use UFh es the major uraniferous compound :I.n the fuel. Thorium Fluoride. All the normel eompounds of thori\nn axre quadri— valent; accordingly, any use of thorium in molten fluoride melts must be aS ThFl‘.I 1.2. choice of Fuel Diluents The fluoride eomposit:l.ons that will be discussed here are 1:I.mited to those viich have & low vapor pressure &t 'm?‘c and which have & melting point no higher then 550 Cs Also, there ig ‘1ittle interest in u:anim . concentrations higher then & few per cent for the fuel of thermal reaétors ’ ‘end therefore mixtures with high UF, content will be omitted from this discussien. Of the pure fluor:!.&es ef mol’aen-;salt refic'bor interest, only ZBeF2 neets the melting point xequirement and 1t is too viscous for use in | the pure state. Thus the fluorides of interest mre ternary or quaternary mixtures containing UF), or TnF,. For the fuel,]_the relatively small smounts of UF!; required meke the corresponding binary ‘or ternsxy. mixtures of the diluents neerly controlling with regexd to physical properties . such as the melting point. Only the elkelisfietal fluorides and the fluo- - rides of bezyll:l.xm end zirconium have been given serious attention. ILead end bismu'l;h fluorides, which might otherwise be ugeful because of their low neutron a,bsorption, have been elimina,ted because they are reaflily _reducea to the metallic eta:be by structural meta.ls such es iron a.nd : Systems Gontaining UF)} Of 'bhe ternary aystems containing UFh ftwo alkali-metal fluorides, o_nJ,v the LiFo-I{F-%UFu system, chown in Fig. 2.1, gnd the LiF-RbF-:IJFh mrs'bem, ghown- in Fig. 2,2, have melting temperatures | ~ below 600 Ceat umnimn coneentratiene belew l@ mole % These two sys'cems ]e.nd the fourwcomponent systems LiFnNa.F-KFnUFh_, tor whic.h the alkali fluo- ride ternsry diegrem is shown in Fig. 2.3, end LiF—-NaF—RbF-UFh are the -75- " UNCLASSIFIED ORNL—LR - DWG 28633 UF, 1035 1000 TEMPERATURE IN °C. 950 ~LiF-4UF, P= PERITECTIC E=EUTECTIC 900 KF-2UF,— 850 P 765 800 4 70 ‘ 70 ~-P 775 E 735~ ° / ///0 sk >90 |, >40 (g | b () =;5v,..>90 | 590 5% 30 15 15 1/1L-»in. (%) ~h0 1/% in. (%) Sho CSuo >ho > 40 >ho * - Elongation recorded is that at outer fibre at time first crack appeared. » Bend angle recorded is that at which f:l.rst crack appeared..',, s o The nickel-base brazing alloys listed in Teble 3.4 have been shown - to be satisfactoxy in contact with the salt mixture LiF-KFhNaF%UFh tests conducted at 1500 F for 100 hr. Further two precious metal-ba.se brazing alloys, 82% Au-18% Ni end 80% Au-ao% Cu, were unattacked in the L:I.F-vKF-Na.F-UFll_ salt after 2000 hr at 1200°F. These two precious metel alloys were also tested in the LiF=-BeF -UFh'mixture and again were not attacked. 3.6, Nondestructive Testing An ultrasonic inspection techniqgue is eveilsble for the detection of flaws in -pla.te ’ ‘piping, and tubing. The _water-imer_sed pulse-echo . ‘ultrasound equipment has been edapted to high-speed use. Eddy current,- dye penetrent, and rediogrephic inspection methods are elso used as required. The inspected materials have included Inconel, austenitic " stainless steel, INOR-8, and the Hastelloy and other nickel-molybdenum- :I'baé';e alloys. ‘ | ' Methods ere being developed for the nondestructive testing of 'weld- ‘ments during initial construction and after replacement by remote means in & high«intensity radiation field, such as that which will 'be"";‘presen‘b if maintenance work is required after operation of a molten-salt . reactor. Thé ultrasonic technique appears to be best suited to semi~automatic a.nd remote operation, and it will probebly be the least affected by. radietion of any of the applicable methods. Studies have indicated that the dif- ficultié-s‘ encountered in the ultrasonic inspe'cticn of Inconel welds and ‘_!,';relds' of some of the austenitic stainless steels because the weld struce | tures have high ultrasonic 'a.ttenuation are not present in the inspection ' of INOR-8 welds. The high ultrasonic attenuation is not present in INOR~8 welds because the base metal and the weld metal are of the same composi- tion. The mechanical equipment designed for the remote welding operation will be useful for the inspec’o:l.on operation. ' ' In the routine inspecticn of reactor-grade construction materials ’ a tube, pipe, plate, or rod is rejected if & void is detected that is larger than 5% of the thickness of the part being inspected. In the A 4 M Table 3.4. Nickel-Base Brazing Alloys for Use in Heat Exchanger Fabrication Brazing Alloy Content (wt %) Components ~ Kiioy 52 Alioy 91 . Alloy 93 Nickel 91.2 91.‘3' | 93.3 S1licon | 4.5 b5 3.5 Boron , 2.9 2 . 9 " ~-;l_-,9 Iron and l | " | o Cexbon Balance Balance Balance - 133 - inspection of a weld, the integrity of the weld must be better than 95% of that of the base metal. Typicsl rejection rates for Inconel and INOR-8 eare given below: Re;jectien Rate (%) o Iten Inconel R INOR-B Pubing 17 .20 Pipe 12 1 Plate Rod 5 | 5 Welds i | i The rejection rates for INOR-B are ex_pected to decline as more experience 1s geined in fabrication. 4, MECHANICAL AND THERMAL PROPERTIES OF INOR-8 L.1, Elasticity A typicel stress-strain curve for INOR-8 &t 1200 F is shown in Fig. 3.1l. Data from similar curves obtained from tests at room temperature up to 1400°F are summarized in Fig. 3.12 to show changes in tensile strength, yield strength, and ductility as & function of temperature. | The temperature dependence of the Young's modulus of this materlel is illustra.ted in Fig. 3.13. - 4.2, Plasticity A series of relaxation tests of INOR-8 at 1200 end 1300°F ha.ve indiceted that creep will be an importent design consideration .for reactors operating in this temperature range. -The rate at which the stress must be relaxed in order to maintain a .eons’ca.fit elastic: ~ strain st 1300°F is shown in Fig. 3.1k, and similar data for 1200°F ere presented in Fig. 3.15. The time lapse before the,ma.tei'ial | becomes plastic is ebout 1 hr at 1300°F and sbout 10 hr at 1200°F, The time péfiéd during which the materiel behaves elastically becomes much longer at lower temperatures and below same fiemperatures , 88 - 13h - e Jooo N o -'SfRESs ~Sy N O 3 ELONGATION. /0" INCHES Pen “UieH Fig. 3.11. Stress=Strain Relationships for INOR-8 at 1200°F. 135 STRESS (psi x 1000) 10 100 90 80 70 60 50 40 30 20 10 UNCLASSIFIED ORNL—-LR-DWG 28709 TENSILE \ STRENGTH . % ELONGATION T~ \YIELD STRENGTH Mo \ \ 4 c 8 10 12 TEMPERATURE (°F x100) 149 te 18 110 {00 90 80 70 60 50 40 30 20 10 Fig. 3.12. Tensile Properties of INOR-8 as a Function of Temperature. 136 PER GENT STRAIN LE1 'YOUNGS MODULUS G N ~ UNCLASSIFIED ) e ORNL-LR-DWG 29238 " | | | | | | | 0 200 400 600 800 1000 1200 1400 A TEMPERATURE (°F) Fig. 3.13. Youngs Modulus for INOR-8 as a Function of Temperature. | 8tl UNCLASSIFIED ORNL-LR-DWG 28708 e T ~~0.2 % CONSTANT e N 24 - N 1 | = 20 N~ S \ ~_ o ~N0.1% CONSTANT € < N\ NN x 16 S 2 N T 0.05 % CONSTANT ¢ N AN — | | N w 8 \\ \\ ¥* S \\fi\ DISCONTINUED TEST S~ % 4 : | =3¢ 0.1 0.2 05 10 2 ) 10 20 90 100 TIME (hr) Fig. 3.14, Relaxation of INOR=-8 at 1300°F at Various Constant Strains. 6€1 C (x103) 28 - o C e | UNCLASSIFIED L ORNL—LR—DWG 29237 24 20j':lf 16 STRESS (psi) 12 T T T T T T TTTTT] T T T TTT] . 0.1% CONSTANT STRAIN | — bunemimr 7328 b e e e S e e Ty S e b . . . _0.05% CONSTANT STRAIN % DISCONTINUED TEST L L LIl L L L L LLLL] L L L LLLLl 2 5 10 20 50 100 200 500 - 1000 | | TIME (hr) - - 'Fig. 3.15. Relaxation of INOR-8 at 1200°F at Various Constant Strains. yet undetermined, the metal will continue to behave elasticelly indefi- nitely. It is possible to sumarize the creep dete by comparing the times to 1,0% total strain as & function of stress in the date shown in Fig. '3.16. The reproducibility of creep datae for this meterisl is indicated by.the separate curves shown in Fig. 3.17. It may be seen that quite good correlation between the creep curves is obtained at the lower stress values. Some scatter in time to rupture occurs at 25,000 psi, & stress vhich corresponds to the 0.2% offset yleld strength at this temperature. ‘Such scatter is to be expected et this high stress level, The tensile strengths of several metals ere compared with the tensile strength of INOR-8 at 1300°F in the following tabulation, and the creep properties of the several alloys at 1.0% strain are compered in Fig. 3.18: ' Tensile Strength at Material 1300°F (psi) 18-8 stainless steel 40,000 Cr-Mo steel (5% Cr) 20,000 Hastelloy B 70,000 Hastelloy C 100,000 Inconel | 60,000 INOR=8 65,000 The test results indicete that the elastic and plastic strengths of INOR-8 &are near the top of the range of strengfih'prbperties of fihe several | alloys commonly considered for high-temperature use. Since INOR-8 was designed to avoid the defects inherent in these other metals, it is ap- perent that the undesirable aspects have been eliminated without emy serious loss in strength. | h.3. Aging Characteristics Numerous secondary phases that are cepeble of embrittling & nickel- base alloy cen exist in the Ni-Mo=Cr-Fe-C system, but no brittle phase exists if the alloy contains less than 20% Mo, 8% Cr, end 5% Fe. INOR-S, - 140 = -~ . vl UNCLASSIFIED ORNL-LR-DWG 28707 105 1100°F 2 __10? g | 1300°F o o & E | EXTRAPOLATED 10 1yr 10 yr 10~ ' 10 100 1000 10,000 100,000 TIME (hr) TO 1% STRAIN Fige 3.16. Creep Data for INOR-8, ¢yl STRAIN (%) 10.0 5.0 2.0 1.0 0.5 0.2 0.1 UNCLASSIFIED ORNL- LR-DWG ‘29236 — T T 1171 T 1T T T 1T - — RESULTS FOR THREE SAMPLES 47 — STRESSED AT 25,000 psi B RESULTS FOR TWO SAMPLES o B STRESSED AT 20,000 psi | | L1 | | LIl | 10 50 100 200 500 1000 2000 TIME (hr) ' Fig. 3.47. - Creep—-Rupture Data for INOR-8. STRESS (psi) 10° 102 : UNCLASSIFIED ORNL-LR-DWG 29235 1T ____H ASTELLOY g TESTED Iy . | — T T T TIT T T T TTTTm T T TTTTTE I B -‘|7r-*'l_-|I|l|_"|*i | LU 10 - 20 50 100 200 500 4000 2000 5000 10,000 | "TIME- TO % STRAIN AT 4300°F (hr) | Flg v'3.18.".-C6rfib0.ri_$6n' of the Creep Properties of Several Alloys. 143 which contains only 15 to 18% Mo, comsists principally of two phases: - u ‘ the nickel-rich s0lid solution a.nd & complex carbide with the approximate | composition (N1, MO)SG. Studies of the effect of the carbides on creep . - strength heve shown thet the highest stremgth exists when & continuous ~ network of carbides surrounds the grains, Tests have shown that carblde o, ‘precipitation does not cause significant embrittlement at temperatures up to 11:~80°F. Agling for 500 hr at various temperatures, as shown in Fig. 3.19, impi'oves the tensile properties of the alloy. The teneile Pro= - ‘perties et room temperature, as shown in Teble 3.5, are virtually unaf- ‘fected by eging, | - 4.4, Thermal Conductivity end Coefficient of Linear Thermal Expension Velues of the thermal conductivity end coefficient of linear thema.l expansion are given in Tables 3,6 and 3.7. 5. OXIDATION RESISTANCE The oxidation resistance of nickel-molybdenum elloys depends on the ~ service temperature, the tempersture cycle, the molybdenum content, end the chromium content. The oxidation rate of the binsry nickel-molybdenum -~ alloy passes through & meximm for the alloy conteining 15% Mo, and ‘the scele formed by the oxidation is N:I.Mooh end NiO. Upon thermal cycling from above 1400°F to below 660° F, the NiMoO) undergoes & phase transforma- tion which causes the protective scale on the oxidized metal to spall. Subsequent temperature cycles then result in an accelersted oxidation rete. Similerly, the oxidation rate of nickel-molybdenum alloys contain- ing chromium passes through a maximm for alloys conteining between 2 and 6% Cr. Alloys containing more than 6% Cr are insensitive to thermal cycling and the molybdenmum content because the oxide scale is predominantly - stable Cry0,. An abrupt decrease, by a factor of about 40, in the oxida~ ~ ‘tion re.te et 1800°F is observed when the chromium content is increased . . from 5.9 to 6.2%. ) - 14k - | ‘ & (x1o3) : UNCLASSIFIED - ORNL-LR-DWG 29234 0 ' l | ] 100 |— — % 90 e L 80 o o W 70 % L, 60 (—n_.J eommmrm ANNEALED {1 hr AT 2100°F | z SO ™ emem == AGED 500 hr AT TEST TEMPERATURE ~ 40 | — | 0.2 % YIELD POINT 2 | 1 | 60 e e B I I B z 40 | ELONGATION = = -—_‘—’ < 30 |— — O = "_-'--.;1000;-;-._’f‘ 1100 1200 4300 1400 TEST TEM PERATURE (°F) an 3, 19 Effec’r of Agmg on ngh Temperafure Tensile Propemes of INOR 8 | 145 ‘Table 3.5. Results of Roam-TemperaturéwEmbritfilemenx'Tests.df'IfiOR§8 - f | Ultififife‘Tenéile_ Yield Point at | Elonéfition Heet Treatment Strength (psi) |0.2% Offset (psi)] (%) Annesled* 114,400 14,700 50 hnnealed and aged 500 hr . | .. . at 1000°F 112,000 k2,500 53 pnnealed and eged 500 hr S T o ~ at 1100°F - 112, 600 - 4k,000 - 51 Pnnealed.and ased,BOO hr Lo ~f s ‘ L at 1200°F | | 112,300 kX, 700 .51 pnnealed and a.ged 500 hr o o et 1300°F 112,000 L%, 500 k9 Annealed and aged 500 hr c .-i at 1400°F 112,400 13,900 50 * 0.0hsuin. sheet, ennegled 1 hr at 21009F and tested at & strain , rate of 0.05 in./min. - 146 - C A A it i s Table 3.6 Comparison of Thermal Conductivity Velves for INOR-8 ~ &nd Inconel at Several Temperatures . | | | | Therma;l. Conductivity /fitu/rt fisec-(oF/ft)7 Temperature (°F) : INOR-8 Inconel | 22 5456 9.kl 92 6.T7 9.92 . 572 o 11,16 10,40 | 752 - 12,10 10.89 933 | k.27 11.61 12 16.21. 12,10 1202 18,15 12,58 - 147 - - Teble 3.7 Lot 'Coefficient of Linaar-Expansian of INOR-8 for Several Temperature Ranges 70 - 400 70 - 600 70 - 800 70 « 1000 70 - 1200 70 - 1400 T0 - 1600 70 - 1800 - . | . R . | Goeff;gient of Linaar'Ex@ansiofi. Tempereture Renge (%) . (in./in,-%F) X 10'6 | 5.6 6,23 - 6,58 6.89 o3k 7,61 . 8,10 8.32 - 148 - b* The oxidation resistanée of INOR~8 is excél'lént » and continuous operation et temperatures up to 1800°F is feesible. ~ Intermittent use at temperatures &s high at 1900°F could be tolerated. - For temperatures up to 1200° F, the oxidation rate is not measurable; it is essentia.lly nonexistent after 1000 hr of exposure in static air, It is estimated that oxidetion of 0.001 to 0.002 in. would occur in 100,000 hr. of opera- tion at 1200°F. The effect of tempersture on the oxidstion rate of the elloy is shovn in Table 3.8. 6. FAERTCATION OF A DUPLEX TUBING HEAT EXCHANGER The compatibility of INOR-8 and sodium is adequate in the tempera~ ture renge presently contempla.ted for molten-salt reactor heat exchanger opei'é.tion. At higher temperatures, mass transfer could become & problem, end therefore the fabrication of duplex tubing has been investigated. - - _Satisfa.-étory duplex tubing has been made that consists of Inconel clad with type 316 stelnless steel, and components for & duplex heat exchenger have been febricated, as shown in Fig. 3.20. The febricetion of duplex tubing is eccomplished by coextrusion of ' billets of the two alloys. The high temperature and pressure used result - in the formation of & metallurgical. bond between the two alloys. In ' Subsequent reduction steps the bonded cmosite behaves as one materiel, ‘The ra.tios of the alloys tha:b comprise the composite axe controllable to | within 3% The unifomity and bond. mtegrity obta.ined in this process ‘ are :!.llustrated :I.n Fig. 3.2.1. ' 'me problem of welding INORa-B--sta.inless steel duplex ubing is - ,,be:l.ng studied. | Meriments have mdicated that proper selection of elloy e _Z’ra:tios end weld design will a.ssure welds thet will be satisfa.ctory in 'h:l.gh-temperature serivce. o ' ' - In order 0 detennine 1irrlw‘-.'.ht‘-z:c' in'berdiffusion ef 'bhe a.lloys wmlld result in a continuous brittle ; - at the interface, » tedts were made in the tanpera.ture renge or 1300 to 1800°F, As expected, & new phase " ‘i-llbgfl‘ Teble 3.8 Oxidation Rete of INOR-8 &t Various Tempereitures* Test Temzjaera;fiure. o (5F 2200 1600 Tt 100 0.00 0.25 0.8 0.52 2,70 . Welght Gain (mg/em) iififiiifi&?"‘“‘%fif4666135 Cub:i.c or. 1ogarithmicA . 0,00 : 067** Lswe 2,0%% ‘ 28.2** | “Cubie. B o Parabolic | - .—Pa.ra.'bolic. a . Linesr’ Shape of Rate' Curv'e- - U .7 mg/cm = 0.001 in, of ofidation. ) - 150 = % Exbra.polated :E'rom date. obta.ined after 170 hr et tempemture. © Fig. 3.20. Components of a Duplex Heat Exchanger Fabricated of Inconel Clad with Type 316 Stainless Steel. 151 l INCHES T £ » n ~ Fig. 3.21. Duplex Tubing Consisting of Inconel Over Type 316 Stainless Steel, Etchant: glyceria regia. 152 » T 4, i appeared at the interface between INOR-8 and the stainless steel which increased in depth along the grain boundaries with increases in the tem- pereture. The interface of & duplex sheet held at 1300°F for 500 hr is shown in Flg. 3.22. Tests of this sheet shoved an ultimete tensile strength of 94,400 psi, & 0.2% offset yleld strength of 36,800 psi, and an elongation of 51%. Creep tests of the sheet showed that the diffusion resulted in an incremse in the ereep resistance with no significa.nt loss of duetility. - - Thus , DO ma,jor difficulties would be expected in the construction - of an BWOR-B--steinless steel heat exchanger. The construction experience thus fer has involved on.ly the 20-tube heat exchanger shoim in Fig. 3.20. 7. -AVAILABILITY OF INOR-8 Two production heets of INOR-8 of 10,000 1b each and mmerous smaller heats of up to 5000 lb have been melted and fabricated into verious shepes by normal production methods Evaluation of these connnereial products . has shown them toshave properties similar to those of the laboratory heats prepared for material selection. Purchase orders are filled by the vendors in one to six months, end the costs renge from $2.00 per pound in ingot fomi 0 $10.00 per pound for cold-drawn welding wire., The costs of f,ubing, pla.te s end bar produc-bs depend to & le.rge extent on the specifications | of the finished produc‘be. o 1"""' e 8 COI»IPATIBILITY OF GRAPHITE WITH MOLTEN SAUIS | AND NIQCEL—BASE ALLOYS ' If g:-aphite could be used as & moderater in direct contect with & "_'-molten sa.l'b s 1t vould make poseible e molten-salt rea.ctor with & breed:l.ng | :-f"ra.tio in excess of one (see Part Ly, Problems that might restrict the ‘usefu.lness of ‘this approach e.re possible rea.ctions of grephite a.nd the ., - 153h o i P UNSTRESSE Interface 316 88 316 88 oos, Interface INOR-8 Fig. 3.22. Unstressed and Stressed Specimens of INOR-8 Clad with Type 314 Stainless Steel After 500 hr at 1300°F, Etchant: electrolytic H,SO, (2% solution). 154 fuel salt, penetration of the pores of the graphite by the fuel, and carburization of the nickel«alloy container. Many molten fluoride salts have been melted end hendled in graphite crucibles, end in these short-temm uses, the grephite is inert to the salt., Tests at temperatures up to lBQOoF with the ternary salt mixture Ifii&r.F-ZrFll_-UF,L geve no indication of the decomposition of the fluoride and no gas evolution eo long &s the graphite was free from a silicon impurity. Ionger«time tests of graphite immersed in fluoride salts have shown greater indications of penetration of the graphite by salts, end it must be assumed thet the salt will eventually penetrate the availeble pores . in the graphite. The "impermesble"” grades of graphite aveileble experi- menta.lly show greater reduced penetration and & semple of high~density, bonded, natural graphite (De Gussa.) ghowed very little penetration. Although q_ua.utita.tive figures are not availeble, it is likely that the extent of penetration of "1mpemee.ble" 'gre.phite gredes cen be tolerated. Although these penetration tests showed no visible effects of attack of the graphite by the salt, anslyses of the salt for carbon showed that et 1500°F more then 1% carbon may be picked up in 100 hr. -The carbon ;picku;p appesrs to be sensitive to temperature, however, inasmuch as only 0. 025% carben was found in the salt efter a 1000=hr exposure at 1500 F. | In some mstances cea.tings have heen found on the graphite after exposure to the salt in Inconel containers » 8 1llustrated in Fig. 3.23. A cross section through the eoating is shown in Fig. 3.2k, The coating wa.s found to be nea.rly pure chrmnium the.t wa.e presmnably tra.nsferred from . the Inconel conta.iner. In the tests Tun thus fe.r, ne peeitive ind:l.cetion has been found .-of ca.rburize.tion of the nickel-e.lloy containers expoeed to molten se.lts o ’--e.na a'aphite at the temperatures e.t present contemplated for power reacters s ( <1500 F) El.‘he carburize.tion effect seems to be q_uite temperature sen- sitive, however, sinee tests e,t 1500 F showed cerburizetion of Eastelloy B to & depth of 0. 005 :l.n. in 500 hr of exposure to N aF-Zth-UFh conte.ining - 155 = i i i i 0.10 IN/DIV. it Fig. 3.23. CCN Graphite (a) Before .and (b) After Exposure for 1000 hr to NaF-ZrF .~UF, (50-46-4 mole %) at 1300°F as an Insert in the Hot Leg of a Thermal-Convection Loop. Nominal bulk density of graphite specimen: 1.9 g/cm3, | | 156 ; (b) after in (a) are pores, (a) Before exposure The black areas 3.23. in (b) ig. F in ilm on the surface 157 f iC - ions of Samples Shown ith salt, illed w Cross Sect * Note the thin metall .24 3 exposure, (b) the pores are f Fig. In graphité. A test of Inconel and gra.ph:l.t'e in & theimal~convection loop in which the meximm bulk temperature of the fluoride salt was 1500° geve a maximum cerburization depth of 0.05 in. in 500 hr, In this case, however, the temperature of the metal-salt interface where the ca.rburi- - zation occurred was considerably higher then 1500°F, probably about - 1650°F. | A mixture of sodium and graphite is known to be & good carburizing agent and tests with it heve confirmed the large effect of temperature on the carburizetion of both Inconel and INOR-B es shown in Teble 3.9. Table 3.9. Effect of Temperature on Cerburization of Inconel and INOR~8 in 100 hr Temperature Depth of Cerburization Alloy ) (°r) (in.) Inconel 1500 _ 0.009 1200 | 0 INOR=8 1500 0.010 1200 | 0 Many additional tests are belng performed with é. variety of molten fluo- ride salts to measure both penetration of the graphite and carburizetion of INOR-8. The effects of carburizetion on the mechanical properties wlll be determined. -‘-158-: o 9. MATERIALS FOR VALVE SEATS AND EEARING SURFACES Neerly all metels, alioys, and.hard-facing materials tend to undergp solid-phase bonding when held togethor under pressure in molten fluoride selts et temperatures above 1000 F. -Such bonding tends to make the startup of hyflrodynamio bearings difficult or impossible, and it reducea | the chance of opening a valve that has been closed for eny length of time. Screening tests in & search for nombonding materials thet will stend wp under the moltenusalt enviromment have indicated that the most promising:materiqls are TiC-Ni end WC~-Co types of cermete with nickel or cobalt cofitents of less than 35 wt %, tungsten, and molybdenum. The tests, in general, have been of less than 1000-hr duretion, so the useful lives of these materials have not yet been determined., | 10. SUMMARY OF MATERIAL PROBLEMS Although much experimentel work remains to be done before the con- struction of & complete power reactor sysiem can begin, 1t is apparent that considereble progress hes been achieved in solving the materiel problems of the reactor core. A strong, steble, and corrosion-resistant alloy with good welding and forming characteristics is evailable. Pro- duction technigues have been developed, and the alloy has been produced in commercial quantities by several alloy'vendors.r'Finally it‘appears that, even at the peak operating temperazure, no serious effect on the alloy oceurs when the molten salt it contains 1s in direct contact with graphite. -9 - R PART & - Ik 'NUCIEAR ASPECTS OF MOI.TEN-;S'AL'I‘ REACTORS. ‘ The ability of certain molten salts to dissolve uranium and thorium salts - in quantities of reactor interest made possible the consideration of fluid-fueled reactors with thorium in the fuel, without the danger of nuclear accidents as a result of the settling of a slurry. This additional degree of freedom has been exploited in the study of mOlten-salt reactors. , , Mixtures of the fluorides of alkali metals and zirconium or beryllium, as discussed in Part 2, possess the most desirable combination of low neutron absorption, high solubility of uranium and thorium compounds, chemical inert- ness at high temperatures, and thermal and radiation stability. The following ' comparison of the capture eioss sections of the alkali metals reveals that Ii7 containing 0.01% 116 has a. cross section at 0.0795 ev and 1150 F that is a factor of & lower than that of sodium, which also has a low cross section: Element o - | Cross Section'(barns)_ ’Ii7-(conta1ning 0.01% 116-) : o 0.073 | sodiwm % - | | 0.290 Potassium o . : 113 Rubi dium o oo Cesium . - 29 ] | The.capture eross section of beryllium 1is also'satisfactorily low at all . neutron energies, and therefore ‘mixtures of IiF and BeF Y which have satisfac- tory melting points, viscosities, and solubilities for UFh and ThFu, were selected for investigation'in the reactor physics study. | Mixtures of NaF, Zth, ‘and UFh were studied.previously, and such a fuel was successfully used in the Aircraft Reactor Experiment (see Parts 1 and 2). Tnconel was shown to be reasonably resistant to corrosion by this mixture at 1500 F, and there is reason to expect that Inconel equipment would have a life of atlleast several years at 1200 F. As a fuel for a central-station power Lipreactor, however, the NaE-Zth systemjhas several serious disadyantages. 'The T w 160 = sodium capture cross section is less favorable than that of Ii?: More important, recent datal indicate.that'the capture cross eection'of zirconium is qnite-high in the epithermal'and.intermediate neutron energy ranges. In eomparison with the IiE%BeFé syetem, the NaFerFu_systen'has inferior heat transfer characteristics. Finally, the INOR alloys (see Part 3) show promise of being as resistant to the beryllium salts as to the zirconium salts, and therefore there is no compelling : reason for selecting the NaF-ZrFu system. I Resctor calculations were performed by means of the Univa02 program Ocusol, > 3 a modification of the Eyewash program,_h end the Oracle program Sorghum.. Ocusol is a Bi-group, multiregion, spherically symmetric, age-diffusion code. The_grnup-'@"' averaged cross sections for the various elements of interest that were used weré | based on the latest availahlesdata.s Wheredata'were lacking, reasonable inter#:‘ polations_based on resonance theory were made. The estimated cross sections o were made to agree with messured resonance integrals where available. Setnra-" tion and Doppler broadening of the resonances in thorium as a function offoOné - centration‘were estimated. Inelastic scattering in thorium and fluorine was | taken into scecount erudely by adjusting the value of got ;P however, the Ocusol . code does not provide for group skipping or anisotropy of scattering. Sorghum is a 51-group, two-regionx zero-dimensional, burnout code. The group~-diffusion equations were integrated over the core to remove the spatial lM’acklin, R. L., Neutron Aetivation Cross Seotions with Sb-Be Néutrons » Phys. Rev. 107, 504-8 (1957). 2U‘nivereal Antomatic Computer at NEW Ybrk University, Institute of Mathematics. - 3Alexander, L. G., et al Operating Inetructions for the Uhivac Program Ocusol-A, A.Mbdification of the Eyewash Program, 0RNL~CF-57-6 4 (1957) l"Alex,e.ruier, J. H., e.nd. Given, N. D. “A Machine Multigroup Calcula,tion.-_ ’. _The ‘Eyewash Program for Univac, ORNLF1925 (1955). 50ak Ridge Automatic Computer and Iogical‘Engine at Oak Ridge National . 6Roberts, J. T. and Alexander, L. G., Cross Sections for the 0cusol~A Program, ORNIPCF'57'6 5 (1957),,, , . _ “ 161 - dependency. The spectrum was computed in terms of a space-averaged‘gronp flux . - from group scattering and 1eakage parameters taken from an Ocusol calculation. B A critical calculation reqnires about 1 min on the Oracle; changes in concen- ‘tration of 1k elements duringia.specified time can then be computed in about 1 sec. The major assumpticn involved is that the group scattering and leakaget _'probabilities do not change appreciably with changes in core composition as : burnup progresses. This asénmption has been verified to a satisfactory degree - of approximation.; '” . | ' The molten salts may be used as homogeneous moderators or simply as fuel carriers in heterogeneous reactors. Although, as discussed below, graphite- moderated heterogeneous reactors have certain potential advantages, their technical feasibility depends upon the compsatibility of fuel, ‘graphite, and metal, which has not as yet ‘been established. For this reason, the homogeneous reactors, although inferior in nuclear performance, have been given greatest attention. ; - - o A preliminary study _in‘dicated that, if the integrity of the core vessel could be guaranteed, theunuclear economy of two-region resctors would probably be superior to that of bare-and reflected one-region reactors. ‘The two-region reactors were, accordingly, studied in detail. Although entrance and exit con- ditions dictate othernthan a spherical shape, it was necessary, for the calcu- 'lations, to use = model comprising the following concentric spherical regions: (1) the core, (2) an INOR-8 core vessel 1/3 in. thick, (3) a blanket approxi- mately 2 £t thick, and (h) an INOR-8 reactor vessel 2/5 in. thick. The diameter’ of the core and the concentration of thorium in the coré were selected as inde- pendent Variables. The primary dependent varisbles were the critical concentra- tion of the fuel (U255 U?BB, or Pu 39), and the distribution of the neutron ' absorptions among the- various atomic species in the reactor. From these, the critical mass, critical inventory, regeneration ratio, burnup rate, etc., can be readily calculated, as described in the following section. 1. Homossflpbus REACTORSE’UEIED WITH 255 y/) . , While the isotope U233 would be a,superior fuel in molten fluoride salt : reactors (see Section 2), it ig. unfortunately not available in quantity. Any realistic appraisal of the immediate capabilities of these reactors must be based on the use of: 0235 o :?.182 - The study of homogeneous reactors wes dirided into two phases: (1) the * mapping of the nuclear characteristics of the initial (1.e., "clean") states - ag a function of core diameter and'thorium concentration; and (2) the analysis of the.subsequent performance of selected_initial states with various process- -ing schenmes and,rates. The detailed results of these studies are givenfin_the following paragraphs. Briefly, it was found that regeneration ratios- of up to 0.65 can be obtained with moderate investment in U2557(1ess than 1000 kg) and that;.if the fission products. are moved (Section 1.2) at a"rate such that the equilibrium inventory is\equai to one year's production, the regeneration ratio can be maintained above 0.5 for at least 20 years. 1.1 Initial States A complete parametric study of molten fluoride sait reactors having" diameters in the range of U to lO”ft and thorium concentrations in the fuel ranging from O to 1 mole % ThFh was performed. In these resctors, the basic . fuel salt (fuel salt No. 1) was & mixture of 31 mole % BeF,, and 69 mole % I4F, which has a density of about 2.0 g/cm at 1150 F. The core vessel was composed of INOR-8. Thé blanket fluld (blanket salt No. 1) was a mixture of 25 mole % ' ThF, and 75 mole % LiF, vhich has e density of about 4.3 g/em’ at 1150°F. In order to shorten the calculations in this series, the reactor vessel was neg- lected, since the resultent error was small. These reactors contained no fis- sion products or nonfissionable isotopes of uranium other than U238 A summary of the results is presented in Table k.1, in which the neutron balance is presented in terms of neutrons absorbed in a given element per neu- tron ebsorbed in U255 (both by fission and the n-y reaction). The sum of the . absorptions is therefore equal to n, the number of neutrons produced by'fission 7fper neutron absorbed in fuel. Further, the sum of the absorptions in 023 and _thorium in the fuel and in thorium Ain the blanket salt glves directly the re- "-:generation ratio.; The losses to other elements are penalties 1mposed on the '~‘regeneration ratio by these poisons, i e., if the core vessel could be con- '-'structed of some material with a negligible cross section, the regeneration : 'ratio could be increased'by the amount listed for capture in the core vessel. | The inventories 1n these reactors depend in part on the volume of the ~ fuel in the pipes, pumps, ‘and hest exchangers in the external portion of the fuel circuit. The inventories Iisted in Table-h 1 are for systems having & Table 4.1, Initial-State Nuclear Characteristics of Two-Region, Bomogeneous, Molten—FluoridefSalt Reactors Fueled with U":"3 2 Fuel salt No. 1: 3L mole % BeF,, + 69 mole % LiF + UF), + ThF), Blanket salt No. 1: 25 mole % ThF) + T5 mole % LiF Total power: 600 Mw (heat) External fuel volume: 339 £t° Case mmber 1 2 3 y 5 6 Core diameter, £t 4 5 5 5 5 5 ThF, in fuel salt, mole % 0 0 0.25 0.5 0.75 1 025% in fuel salt, mole % 0.952 0,318 0.56L 0.721 0.845 0.938 %97 atom density* 33,8 11.3 20.1 25.6 30,0 33,5 Critical mass, kg of U>7 124 81.0 144 183 215 239 Criticel inventory, kg of U-2? 1380 501 891 1130 1330 1480 Neutron absorption ratios¥* "% (gissions) 0.7025 0.7185 0,700k 0.6996 0.7015 0.7041 22 (a-7) 0.29T77 0.2815 0.2996 0.300% 0.2985 0.2959 Be-Li-F in fuel salt 0.0551 0,0871L 0.0657 0.060k 0.,0581 0,0568 Core vessel '~ 0.0560 0.0848 0.0577 0.0485 0.0436 0.0402 Li-F in blanket salt 0.0128 0.0138 0.0108 0.0098 0.0095 0.0090 Leakage ‘ 0.0229 0.0156 0.0147 0.,0043 0.0141 0.0140 1®® in fuel sart 0.0430 0.0426 0.0463 0.045L 0.0431 0.0kl2 Th in fuel salt 0.0832 0,1289 0.16l4 0.1873 Th in blenket salt 0.5448 0.5309 o0.4516 O.4211 0.4031L 0.3905 Neutron yield, 1 173 1.77 1.3 1.73 1.73 174 ‘Medisn fission energy, ev 270 15.7 105 158 270 o5 Thermal fissions, % 0.052 6.2 0.87 0.22 0.87 0.040 - a-y capture-to-fission retio, @ 0.42 0.39 0.43 0.43 0.43 0.4203 Regeneration ratio ' 0.59 0.57 0.58 0.60 0.6L 0.62 * Atoms ( x 10719)/end. *% Neutrons sbsorbed per neutron sbsorbed in ez -16&-' 4 Table 4.1 (continued) Case number 7 8 9 10 1 12 Core diameter, £t 6 6 6 6 6 7 ThFy, in fuel salt, mole % 0 0.25 0.5 0.75 1 0.25 U°2% in fuel salt, mole % - 0.107 0.229 0.408 0.552 0.662 0.l1k 22 atom density* 3,80 8.3 14.5 19.6 23.5 %.05 Criticel mass, kg of U>? k7.0 101 179 oly3 291 79.6 Critical inventory, kg of U-2° 188 Lok TL6 972 1160 230 Neutron sbsorption ratios *» | | U222 (£1ssions) CO.TTTL 0.7343 0.7082 0.7000 0.700% 0.T748 722 (n-7) 0.2229 0.2657 0.2918 0.3000 0.2996 0.2252 Be-Li-F in fuel salt © 0.1981 0.1082 0.0770 0.0669 0.0631 0,1880 Core vessel " 0.1353 0.0795 0.0542 0.0435 0.0388 0.0951 Ii-F in blanket salt 0,016+ 0.0116 0.009. 0,008L 0.007% 0.0123 Leakage 0.0137 0.0129 0.0122 0,0119 0.0116 0.0068 " in fuel selt 0.0245 0.0575 0.0477 0.0467 0.0452 0.0254 Th in fuel salt | 0.1%21 0.1841 0.21%2 0.2438 0.1761 Th in blanket salt 0.53L2 0.4318 0.3683 0.3378 0.3202 0.4098 Neutron yleld, n 192 182 1.75 1.73 1.73 1.91 Median fission energy, ev 0.18 5.6 38 100 120 0.16 Thermal fissions, % 35 13 3 0.56 0.48 33 n-y capture-to-fission ra:bio, a_ 0.28. . 0.36 . o.b O.h2 0.k2 0.29 ::_:.'Regeneration ratio S 0. 56 - 0.61 . 0.60 - 0.60 - 0.6 0.6L ' ?3'i{:,.:,* _AtOms (x lO 19)/cm O R **'Neutrons absorbed per neutron absorbed :l.n 023 5 165 Teble 4,1 (continued) Case muber Core diameter, f’b ThF% in fuel salt, mole % U"%in fuel salt, mole 4 U255 atom density* Critical mass, kg of U235 Critical inventory, kg of U23 > Neutron absorptibn ratios¥** y=o2 (£issions) v (n-7) Be-Li-F in fuel salt - Core vessel Li-F in blanket salt Leakage PR in fuel salt Th in fuel salt Th in blanket salt Neutron yleld, g Medlan fission energy, ev Thermel fissions, % n-y capture-to-fission ratio, « Regeneration ratio * Atoms (x 1077)/cm’ 13 1 15 16 17 8 8 8 8 8 . 0 0.25 0.5 0.75 1 0.0k7 0.078 0.132 0.226 0.349 1.66 2.77 k.67 8.05 12.4 48.7 81.3 137 236 36L 110 184 310 535 824 0.8007 0.7930 0.7671L 0.7362 0.T146 0.1995 0.2070 0.2329 0.2638 0.285k 0.4130 0.2616 0.1682 0.1107 0.0846 0,149 0.1032 0.0722 0.0500 0.0373 0.0143 0.0112 0.0089 0.0071 0.0057 0.,0084 0.,0082 0.0080 0.,0077 0.0074 0.0143 0.0196 0.0272 0.0368 0.0428 . 0.20L5 0.3048 0.3397 0.3515 0.4L073 0.3503 0,3056 0.,2664 0.2356 2.00 1.96 1.89 1.82 1.76 Thermal 0.10 0.17 5.3 o7 59 k5 29 13 5 0.25 0.26 0.30 0.36 0.40 0.h2 0.57 0.6k 0.64 0.63 ¥* Neutrons absorbed per neutron absorbed in UE'5 5 . -166- Teble L.l (continued) Case number Core diameter, £t in fuel sult, mole % u25g in fuel salt, mole % " atom denstty* Critical mass, kg of U255 Critical inventory, kg of U-27 Neutron sbsorption ratios*s U235 (fissions) v">? (n-7) Be-Li-F in fuel salt Core vessel Li-F in blanket salt Leakage Uzj’8 in fuel salt Th in fuel salt 'Th in blanket salt Neutron yield, q Medisn fission energy, ev Thermael fissions, % n-y capture-to-fission ratio, @ Regenera:tion ratio * Atoms (x 10 l9(/cm 18 19 20 10 10 10 0 0.25 0.5 - . 0.033 0.052 0.081 1.175 1.86 2.88 67.3 107 165 111 176 272 0.8229 0.7428 0.7902 0.1T7L 0.2572 0.2098 0.5713 0.3726 0.2486 0.1291 0.0915 0.0669 0.011% 0.0089 0,0073 0.006L 0.0060 0.0059 0,0120 0.0153 0.0209 ~ 0.2409 0.3601 0,303 0.26l7 0,2332 2.03 2.00 1.95 Thermel Thermal 0.100 66 56 W3 0.21 0.24 0.26 0.32 0.52 0.62 : .** Neutrons absorbed per neutron absor'bed .'m U255 ' 167 21 10 0.75 0.127 k.50 258 k25 0.7693 0.2307 0.1735 0.0497 0.0060 0.0057 0.0266 0.4324 0,2063 1.90 0.156 30 0.30 0.67 22 10 1 0.205 7.28 k7 €87 0.7428 0.2572 - 0.1206 0.0363 '0.00L49 0.0055 0.0343 0.4506 0,1825 1.83 1.36 16 0.35 0.67 "volume of 339 ft5 externsl to the core, which corresponds approximatelyfto a power level of 600 Mw of heet. In these calculations it was assumed that the heat was transferred to an intermediate coolant composed of the fluorides of | 14, Be,'and Na before being'transferred_to'sodium metal. In more recent designs -'(see'Psrt 1), this intermediate seltsloop has been replaced by a sodium loop and the external volumes are-somewhst'less because of the-ifiproVed equipnent designa.ndlayout. - - | o Critical Concentration, Mass, Inventory, and Regeneration Ratio., The | data in Teble 4.1 are moyre easily comprehended in the form of graphs, such as Pig. k.1, which presents the critical concentration in these reactors as a ' function of core diameter and thorium concentration in the fuel salt. The , _data points represent calculated values, and the 1ines are reasonsble inter- polations. The maximum concentration calculated, about 35 x 10 19 atoms of , 55 per cubic centimeter of fuel sslt, or about 1 mole % UFu, is an order of ,' magnitude smaller than the maximum permissible concentration (sbout 10 mole. %). The corresponding eritical masses are graphed in Fig. 4.2. As may be seen the critical mass is.a rather complex function of the diameter and the thorium : concentration. - The calculated points are shown here also, and the solid lines represent, it is felt, reliable interpolations. The dashed lines were drawn vwhere insufficienttnumbers of points were calculated to define the curves preQ =cisely; however, they are'thought to be qualitatively correct. Since reactors -having dismeters less than 6 ft are not economically attractive, only one case with a b-ft-dia core was computed. ‘ The critical masses obtained in this study ranged from 40 to 400 kg of U255; However, the critical inventory in the entire fuel circuit is of more interest to the reactor designer than is the criticel mass. The critical in- ' ventories corresponding to an external fuel volume of 339 ft5 are therefore ~ shown in Fig. h 3. Inventories for other external volumes may be oomputed ~ from the relation, | | | 6v =M ‘(1~+ —Des) » . n . where D is the core diameter in feet, M 1is the“critical mass teken from /{f Fig. L. 2 Vé is the volume ‘of the external system in cubic feet, and I 18 the inventory. in kilograms of U235 The inventories plotted in Fig. 4.3 range _p_.-]s's'__ Fig. 4.1. 169 - 170 O © © O i o - 0 Ay - 171 from slightly above 100 kg in an 8- ft dia-core with no thorium present to ‘"1500 kg in a 5-ft~dia core with 1 mole % ThFh present.i The optimum combination of’ core diameter and thorium concentration is, o fa qualitatively, that-Which minimizes the sum of inventory - charges (including charges on Ii7, ‘Be, and Th) and fuel reprocessing costs.' The fuel costs are . directly related to the regeneration ratio, and this varies in a complex | ,_manner with” inventory of U255 and thorium concentration, as shovn in Meg. h k It may be seen that, at a glven thorium concentration, the regeneration ratio ' (with one exception) passes through a maximum as the core dismeter is varied N _between 5 and 10 ft. These maxima ‘increase with increasing thorium concen- tration, but the inventory values at which they oceur also incérease. , T Plotting the maximum regeneration ratio versus critical inventory -generates . the eurve shown in Fig. 4.5, It may be seen that a small investment "in U235 (200. kg) will give a regeneration ratio of 0, 58 that 400 kg will give a ratio ..° of O. 66, and that further increases in fuel. inventory have little effect. The effects of changes in the compositions of the fuel and blanket salts _are indicated in the following description of the results of a series of cal- ' culations for which salts with more favorable melting points and viscosities " were assumed. The fieF content vas raised to 37 mole % in the fuel salt (erl . salt No. 2) and the blanket composition (blanket salt No..2) was fixed at 13 mole % THF),, 16 mole 4, BeF s end Tl mole % IiF. Blanket salt No. 2 is & some- - what better reflector than No. 1, and fuel salt No. 2 8 somewhat better moderator. A8 & result, at a. given core diameter and’ thorium concentration in the fuel salt, fboth the critical concentration and the regeneration ratio are somewhat lower for the No. 2 salts. o ’ | Reservations concerning the feasibility of constructing and guaranteeing the 1ntegrity of core vessels in large sizes (10 ft.and over), together wvith preliminary consideration of inventory charges for large systems, led to ‘the conclusion that a feasible reactor would probably have a ccre dlameter lying in the range betveen 6 and 8 ft. Accordingly, ‘8, parametric study-in thiS'range with the No. 2 fuel and blanket salts was performed~ In this study the presence | of an outer reactor vessel consisting of 2/5 in. of INOR-8 was teken into sccount. | The results are presented in Teble h:aaand.Figs h'6 and 4.7.. In general)’ the ‘nuclear performance is somewhat‘better with the -No." 2 salt than with the No. 1 salt. = o e . AFa. o H 173 Fig. 4.5. 174 | » Table k4.2, Initial-State Nuclear Chara.cteristics of Two-Region, Homogeneous, Fuel salt No. 2 Total power: Extermnal fuel volume: MoltenflFluoride—SaJ.t Reactors Fueled with U235 37 mole % BeF, + 63 mole % LiF + UF# + ThFh Blanket salt No. 2: 13 mole % ThF) + 16 mole % BeF, + 7L mole % LiF . '?*,'Regeneration ra'bio Case number Core diameter, It in fuel salt, mole % u25% in fuel salt, mole % U255 atom density¥ Critical mass, kg of UE55 Critical inventory, kg of U255 Neutron absorption ratiog¥** 2 (fissions) v (n-7) Be-Li-F in fuel salt Core vessel ‘ Li-F in blanket salt Outer vessel Leakage P 10 fuel satt Th in fuel salt Th in blanket salt Neutron ‘y‘ield, N ' Median fission energy, ev S . ”'I'hemal fissions, b o n-y capture~to-f:lssion ratio ; ‘d,fi; 0 59 * Atoms (x 10 9)/cm . 7" SRR LT : i i.w** Neutrons absorbed per neutron absorbed 1;1 U255 L 0.8 0.8 175 600 Mw (heat) 339 £t 23 2k 25 26 27 28 6 6 6 6 7 7 0,25 0.5 0.75 1 0.25 0.5 0.169 0.310 0.423 0.580 0.084 0.155 5.87 10.91 15.95 20.49 3,13 5.38 T2.7 135 198 254 6L.5 106 201 540 790 1010 178 306 0.7516 0.71Th O.70M+ 0.6058 0.7888 0.7572 0.2u84 0.2826 0.2956 0.3042 0.2112 0.2428 0.1307 0.0900 0.0763 0.0692 0.2147 0.1397 - 0.1098 0.0726 0.0575 0.0473 0.1328 0.0905 ©0.02lF 0.0159 0.0132 0.0117 0.0215 0.0167 0.002% 0.0021 0.0021 0.0019 0.0019 0.0018 0.0070 0.0065 0.0064 0.0061 0.0052 0.0050 0.0325 0.0426 0.0452 0.0477 0.021k 0.0307 0.1360 0.1902 0.2212 0.2387 0.1733 0.2565 0. 4165 0.3521 @ 0.3178 0.2962 0.3770 0.329k - 1.86 '1 77 . 1. 7h 1.72 1.95 1.87 oo - 1ot S8 1 761 0223 0.5 ."fé211Q‘ '*¥¥{7:‘;<7}‘ 2.8 0.8 ;;' h}f- 2l S 0.33 "b'39"-:-*6ifié;; ok 037 0.3 058 ,o 57 0.62 Teble 4.2 (continued) Case pnunber Core diameter, ft ThF) in fuel salt, mole % U%3 4in fuel salt, mole % U‘e35 atom density* Critical mass, kg of U235 Critical inventory, kg of U255 Neutron absorption ratios#¥* F? (fissions) 0"’ (n-7) Be-Li~F In fuel salt Core vessel Li-F in blanket salt Outer vessel Leakage ®® in fuel salt Th in fuel salt Th in blanket salt Neutron yield, q . Median fission energy, ev Thermal fissions, % n¥7 capture-to-fission ratio, o " Regeneration ratio * Atoms (x 10'19)/cm5. 29 30 31 32 [ T 8 8 0.75 1 0.25 0.5 0.254 0.366 0.064 0.099 8.70 13.79 2.24 3451 171 271 65.7 103 Lol 783 149 233 0.7282 0.7094 0.8014 0.78LL 0.2718 0.2906 0.1986 0.2186 0.1010 0.0824 0.2769 0.1945 0.0644 0.0497 0.1308 0.0967 0.0131 0.0108 0.0198 0.0162 0.0016 0.0015 0.0017 0.0016 0.0048 0.0045 0.0045 0.0043 0.0392 0.044T 0.0177 0.0233 0.2880 0.3022 0.1978 0,30L43 0.2866 0.2566 0.3%240 0.2892 1.80 1.75 1.97 1.93 T.61L 25.65 51% thermal 0.136 11 L.3 51 28 0.37 0.4 0.25 0.28 0.61 0.60 0.54% 0.62 ¥¥ Neutrons absorbed per neutron absorbed in U235. 176~ 55 8 0.75 0.163 5.62 165 374 0.7536 0.2L64 0.1354 0.0606 0.0130 0.001h4 0.0042 0.0315 0.3501 0.2561 1.86 0.518 25 0.33 0.64 3l 8 1 0.254 9.09 267 60k 0.7288 0.2712 0.1016 0.0518 0.0105 0.0013 0.0040 0.0392 0.3637 0.2280 1.80 7.75 ll . 0.37 0.65 178 Neutron Balances andNMiscellaneous Details.:'The distributions of the neutron captures are given in Tables 4.1 and 4.2, where the relative hard- ness of the neutron spectrum is indicated by the median fission energies and the percentages of thermal fissions. It may be seen that losses to Ii, Be, and F in the fuel salt and to the core vessel are substantial, especially in the more thermal reactors (e.g., Case No. 18). However, in the thermal 235 reactors, losses by radiative capture in U are relatively low. Increasing the hardness decreases losses to salt and core vessel sharply (Case No. 5), but increases the loss to the n-y reaction. It is these opposing trends which account for the complicated relation between regeneration ratio and critical inventory exhibited in Figs. 4.4 and 4.7. The numbers given for capture in the Ii and F in the blanket show that these elements are well shielded by the thorium in the blanket, and the leakage values show that leakage from the reactor is less than 0.01 neutron per neutron absorbed in U255'in reactors over 6 ft in diameter. The blanket contributes substantially to the regeneration of fuel, accounting for not less than one-third of the total even in the 10-ft-dia core containing 1 mole % ThF) « Effect of Substitution of Sodium for 117 In the event that 117 should prove not to be available in quantity, it would be possible to operate the reactor with mixtures of sodium and beryllium fluorides as the basic fuel salt. The penalty imposed by sodium in terms of critical inventory and regen- eration ratio is shown in Fig. 4.8, where typical Na-Be systems are compared with the correSponding Ii?Be systems. With no thorium in the core, the use - of sodium increases the criticel inventory by a factor of 1.5 (to about 300 kg) | eand 1owers the regeneration ratio by a factor of 2.~ The regeneration penalty “fis less severe, percentagewise, with l mole % ThFh 1n the fuel salt, in an - f41f8 ft-dia core, the inventory rises from 800 kg to. 1100 kg and the regeneration :fiiflratio falls from O 62 to O 50._ There is scme doubt concerning the validity of _-ffithe pcint representing the 10—ft dia core for the Na-Be system with 1.0 mole % '3'5ThFh, the explanation for the apparently abberant behavicr may'be that sodium | ,;is relatively'more harmful in the 1arge, near-thermal systems. Details of gthe neutron balances are given in Table h 3 ‘“f :J'”'jf.-'_ . , N Reactivity Coefficients. By means of a series of calculations in which the thermal base, the core radius, and the density of the fuel salt are varied independently, the components of the temperature coefficient of reactivity of - 179 - 180 Table k.3, Initial Nuclear Characteristics of M-Regiog, Homogeneous, Molten Sodium-Beryllium Fluoride Reactors Fueled with l.l23 > Fuel salt: 53 mole % KaF + 46 mole % BeF, + 1 mole % ('th + UFh) Blanket salt: 58 mole % NaF + 35 mole % BeF, + 7 mole % ThF), ‘Total power: 600 Mw (heat) External fuel volume: 339 ft5 Case number 35 36 37 38 39 ko Core diameter, ft 6 6 :) 8 10 10 in fuel salt, mole % 0 ‘1 0 1l 0 1 u'°-’3g in fuel salt, mole % 0.17% 0.70L% 0.091 0.465 0.070 0.282 U™ atom density* 6,17 249 3.28 165 247 1240 Critical mass, kg of U-0° 764 308 951 48h W2 10 Critical inventory, kg of U>2° 306 1230 215 1100 o3l 1170 Neutron sbsorption ratios** | U255 (fissions) 0.T417 0.6986 0.7737 = O.70L1 0.7862 0.7081 1°? (n-y) 0.2583 0.30Lh 0.2265 0.2989 0.2138 0.2919 Na-Be-F in fuel salt 0.2731 0.1153 0.4755 0.1kl 0.6119 0.2306 Core vessel 0.1181 0.0476 0.1125 0.0392 0.0917 0.2306 Na-Be-F in blanket salt 0.0821 0.0431L 0.0660 0.0315 0.0495 0.2306 Leakage 0.0222 0.0182 0.0145 0.0116 0.0105 0.2306 U238 in fuel salt 0.0360 0.0477 0.0263 0.0484 0.0232 0.0467 Th in fuel salt 0.2418 0.3150 10,3670 Th in blanket salt 0.3004 0.2120 0.2165 0.1450 0.1550 0.1048 Neutron yield, n 1.83 1.7% 1.91 1.73 1.9% 1.75 Median fission energy, ev 1.3 190 0.20 36 0.087 Thermal fissions, % 17 o2 34 1.h 4.1 n#‘capture&d-fission ratio, 0.25 0.’4—3 0.29 0.43 0.27 0.h41 ' ’ 0.3% 0.5 0.2k 0.5. 0.8 0.52 _ Regenefation ratio , * Atoms (x 10° lg)/cm | ** Neutrons a‘bsor'bed per neutron absorbed in U235 181 8 reactor can be estimated as illustrated below for a core 8 ft in diameter and & thorium concentration of 0.75 mole % in the fuel salt at 1150°F. From the expre551on, -k = f(T)pr)_)' 4where X is the multiplication constant, T is the mean temperature in the core,. p 1is the mean density of the fuel salt in the core, and R is the core radius, it follows that ldk=1 @15) '+_1_<_a_5) d.R+l(_Lk) o, () kdf k |ao7T k | 9R aT 2p aT - - | PyR \ p,T : /R,T 'where'thetermJ%f(%%% ' represents the fractional change in k due D:R change in the thermal irepresents the change due to expulsion of fuel from the core by thermal expansion of the fluid, and the term -% gg an increase in core volume and fuel holding base for slowing down of neutrons, the tenn represents the change due to p,T capacity. The coefficient %% may be related to the coefficient for linear expension, a, of INOR-8, viz: dR =RC!. ~ arT 'Iikewise the term %T may be related to the coefficient of cubical expansion, ‘B, of the fuel salt: 7 = "PB . From the nuclear calculations, the’components of the temperature coeffi- cient were estimated, as follows: | -(0.13 ¥ 0.02) x 10"5/°F == 01“’ REILT, O s . = s i . + 0.412 ¥ 0.0005 i (V" ol v - 3 )] 14 . - 0.405 < 0.0005 W = ofu S o w Y2 0 - 182 - “tion of the resonances in thorium and U - the effective widths of the resonances would be increased at higher temperatures, The linear coefficient of expension,'a,'of INOR-8 wcs estimated to be (8.0 t 0.5) x lofg/bF,T and the coefficient of cubical expansion, B, of the fuel was estimated to be-(9.889 ¥ 0.005) x 19-3/bF from & correlation - - of the density given by Powers. Substitution of these values in Eq. 1 gives ’ld-k_ 5for, .E.&__-(380-o.oh)x10/1=‘. for the temperature coefficient of reactivity of the fuel. In this calcula- tionm, the effect of changes with temperature in Doppler broadening and satura- 235 were not taken into account. Since~ the thorium would contribute a reactivity decrease and- the U235 an increase. These effects are thought to be small, and they tend to cancel each other. Additional coefficients of interest are those for U255 and thorium. For the 8-ft-dia cores, 14 (0.17 N, (") x 10"15> N(UEBS) J k -3N(Uz§_5)N(Th) 24 Nc(U235) x 10712 and - | : | .' B | - o - 5 N(Th) [ 2k \ = N(m) [k ax, (v*?) k (om(m))L 235y ok IN(ED) sy () where | o | | | (u2 %) . . ;, 08(',5 0. os9sn(m) %1072 "EETEET"' T In tfi@se equations, N(0255) represents the atomic density of U255 in atoms, per cubic centimeter, N, (0255) is. the critical density of 0235, and N(Th) 18 the density of thorium atoms.;,., - 7K’inyon, B. W., Private Communication, ORNL (1958) 8waers, W. D., Private. Ccmmunication, ORNL (1958) - 183 - Heat Release in Core Vessel and Blanket. The core vessel Mo'f-é._hibltéh{ | salt reactor is‘ heated by gamne. radiatibn'emana.ting from the core and blanket and from within the core vessel 1tse1f. Estima.tes of the gamma heating can | be cbtained by dete.iled enalyses of the type illustrated by Alexander a.nd Ma.nn 9 The gamma, -ray heating in the core vessel of a reactor with an 8-ft- d:la core a.nd 0.5 mole % IhFh in the fuel salt has been estimated to be the following° ' Soureé' | S | Hea.t Réiea.se Rate (w/cm5) | Badioaétive_decay in core | 1 -Fissio':n,’h n-.7 capture, and o S 1nelastic ecattering in core - - 5.2 n-y capture in core vessel o ks " n-7 capture in blanket | - - 0.3 o | Total 11.k Estimates of gaxma-ray source strengths can be used to provide a crude estimate of the gemma~ray current entering the blanket. For the 8-ft-dia ‘core, the core contributes 45,3 w of gamma energy per square centimeter to the blanket and the core vessel contributes 6.8 w/cm , which, multiplied by _the gurface area of the core vessel, gives a total energy escape into the blanket of 28.8 Mv. Some of this energy will be reflected into the core, of course, and some will escape from the reactor vesael, and therefore the value ~of 28.8 Mv is en upper limit. To this may be added the heat released by cap- ture of neutrons in the blanket. From. the Ocusol-A caleulation for the 8-ft-dia. core and & fuel salt conta.ining 0.5 mole % ThF), it was found that . '0 176 of the neutrons would be captured in the bla.nketa If an energy relea.se . of T mev/capture 1 assumed, the heat release at s power level of 600 Mw {heat) 7— 1s estimated to be 8.6 M7, The total is thus 47.4 Mw, or say, 50 ¥ 10 Mw,_ to . allow for errors. .. | S | | No allovance ves made for fissions in the blanket. These would add 6 Mo - for each 1% of the fissions occurring in the blanket, Tiuis it appears that the heat release rate in the blanket m:l'g‘h'f: 'ra,rige up to 80 Mw. 9Alexa.nder, L G., end Mann, L. A., First Estimate of the Germg, Hea,ting ~ in the Core Vessel of & Molten Fluoride Converter, 0RNI.-CF-57-12-57 (19575. «-J.&h-» 1.2 Intermediate States Without Reprocessing of Fdel'Salt. The nuclear performance of a homo- geneous molten-sslt reactor changes during operation at power because of the accumulation of fission:products end nonfissionable isotopes of‘uranium. It is necessary to add U255 to the fuel salt to overcome these poisons;'and, as a result, the neutron spectrum is hardened and the regeneration ratio decreases 235 because of the accompanying decrease in 1 for U7 and the increased competition for neutrons by the poisons relative to thorium. The accumulation of the superior fuel U233 compensates for these effects only in part. The decline in the regen- eration ratio and the inerease in the critical inventory during the first year of operation of three reaotors'having 8-ft-dia cores charged, respectively, with - 0.25, 0.75, and 1 mole % ThFhare-illustrated in Fig. 4.9. The critical inventory increases by about 300 kg, and the regeneration ratio falls about 164. The gross burnup of fuel in the reactor charged with 1 mole % ThF), and operated at 60C Mw with a load factor of 80 amounts to about 0.73 kg/day. The 2 burnup falls from this value as U~ assumes part of the load. During the first month of operation, the °? burnup averages 0.69 kg/day. Overcoming the poisons requires 1.53 kg more and brings the feed rate to 2.22 kg/day. 'The initial rate is high because of the holdup of bred-fuel in the form of Pa->2. As the concentration of this isotope approaches equilibrium, the £ feed rate falls rapidly. At the end of the first year the burnup rate has fallen to 0.62 kg/day and the feed rate to 1.28 kg/day. At this time U253 contributes about 12% of the fissions. The reactor contains 893 kg of U255,770 kg of U233 7 kg of Pu 59 62 kg of U236 ‘and 181 kg of fission products.: The Ua;g end the fission prodncts capture 1.8 ,and 5 8% of all neutrons and impair the regeneratiOn ratio by 0. 10 units.r Details Vfof the inventories and concentrations are given in Table h h._idfffl,,-f ._1'fi5, - ' With,Reprocessing of Fuel Salt. If the fission products were allowed to | _.'ffaccumulate indefinitely, the fuel inventory would become prohibitively large 'nieand the neutron economy would become very poor. Ebwever, 1f the f1531°n Pr°"ii" , ffducts are removed, as described in Part 6 at a rste such thet the equilibrium ' lfl~1nventoryis, for. example, eqpal 40 the first year 's production, ‘then the in-- ‘.d:crease din- 0255 inventory and the decrease in regeneration ratio are-effectively *'farrested, as shown in Fig. 4,10, The fuel-addition Tate drops 1mmediately from 1.28 to 0.73 kg/day when processing is started. At the end of two years, - 185 - OILY Y NOILOYINTD I P AMOLNIAN/ 186 Table 4.4k, Nuclear Performence of a Two-Region, Homogeneous, Molten-Fluoride-Selt Reactor Fueled with U25 5 and Containing 1 mole % ThF) in the Fuel Salt Core diameter: 8 ft External fuel volume: 339 £t Total power: 600 Mw (heat) Load factor: 0.8 . | ",-_.,":,_—U-235 Feed Rate, kg/day L ,Regeneration Ratio 2 22?_§if”g:’ e ©1.28-0.73 Initial State After 1 year _ Inventory Absorptions Fissions Inventory Absorptions Fissions (ke) (%) (R (ke) (%) (%) Core Elements Th-~232 2,100 20.3 2,100 16.7 Pa~23%3 8.2 0.3 U=-233 6L.0 5.9 12,5 U-234 1.9 0.0 U-235 60k 5504 100 893 49.3. 86.3 U-236 62.2 1.8 Np=-237 ‘ h.2 0.2 U-238 45.3 2.2 57.9 2.0 Pu-239 6.8 0.8 1.2 Fission fragments 181 3.8 Li-7 %,920 1.9 3,920 0.9 Be-9 3,008 0.6 5,008 0.5 F-19 214,000 3.2 2lt,000 3.0 Blanket Element e | U-esza SO e e 8.7 , motal Fuel*“ " f' | *5fi;69§1'7§j7f;j17i;f'Wf;i." 965 U235 Burmip. Bate, o e | o kgfaay .0 o9 o 0.62 187 . Teble kb (continued) After 2 years After 5 years Inventory Absorptions Fissions Inventory Absorptions Fissions (ke) (%) (%) (kg) (%) (%) Core Elements Th-232 2,100 16.3 o 2,100 15.4 Pa-233 .9 0.2 - 7.5 0.2 U-233 110 9.7 20,8 201 15.3 33.0 U-234 a 6.5 0. 271 0.k U-235 863 443 7.4 818 36.9 64l U-236 115 3.1 222 5.2 ‘ Np-237 0.8 0.4 1.8 0.8 , U-238 69.7 2.3 | 9.0 2.7 : Pu-239 12.0 1.3 1.8 2h,3 2.0 2.9 Fission fragments 181 3.6 181 3.l ' A Li-7 3,920 0.8 3,920 0.6 Be-9 3,008 0.5 3,008 0.5 F-19 2k, 000 3.0 2k, 000 3.0 Blanket Element U-233 16 2L Total Fuel 990 1045 U-255 Burnup Rate, kg/day 0.58 , 0.47 U-235 Feed Rate, : kg/day 0.50 0.45 Regeneration Ratio 0.53 0.54 -188- Table 4.4 (contipued) After 10 years . After 20 years Inventory Absorptions Fissions Inventory Absorptions Fissions (kg) (%) (%) (keg) (%) (%) Core Elements , Th-232 2,100 14.6 2,100 13.7 Pa-233 Tl 0.2 6.7 0.2 U-233 266 L7.6 38.3 322 18.8 4.0 U-23k 6k 0.8 12k 1.4 U-235 831 3345 58,2 872 31.7 54.9 U-236 328 6.7 1150 7.9 Np-237 2.6 0.9 3,2 1.0 U-238 10.8 2.9 12.9 3.0 Pu-239 373 2.4 3.5 52,6 2.8 k.1 Fission fragments 181 2.7 181 2.4 Li-T 3,920 0.5 3,920 0.k Be-9 3,008 0.5 3,008 0.5 F-19 2k, 000 3.0 2l ,000 3.0 Blanket Element U-233 28 33 Total Fuel 1,129 1,232 U-235 Burnup Rate, kg/day 0.h1 - 0.38 U-235 Feed Rate, - ~ kgfday ' Ous 0.39 Regenerstion Ratio ':0.533 o 0.5%0 189 Core Diameter - 8-0" Power— 600 Mw Load Factor- 0.8 08 .‘g S S 07 Q $ o : . \ . t o] | S o é \\ ' ' - 1% 7./7/';\ T+~ L Decréasing 7'/7) o4 0 /0 20 ) 1000 // //\ — 1% Thiy 800 T — - l _ - T " ———Decreasing Thhy 600 P> & \ 0, | S /% Th\ T 400 - _ Uzss 200 — L1 o 0 /0 20 Time of operation, years Fig. 4.10. Long-Term Nuclear Performance of Typical, Two-Region, Homogeneous, Molten Fluoride Reactor Fueled with U235. 190 the addition rate is down to 0.50 kg/day, and it continues to decline slowly to 0.39 kg/day after 20 years of operation. The nonfissionable isotopes of - uranium continue to accumulate, of course, but these are nearly compensated for by the ingrowth of U772, As shown in Fig. 4.10, the inventory of N actually decreases for several years in a typlcal case, and then increases only moderately during a lifetime of 20 years. The rapid increase in eritical inventory of U235 during the first year can be avoided by partial withdrawal of thorium. In Fig. 4.10 the dashed lines indicate the course of events when thorium is removed at the rate of 1/900 per day. Burnup reduces the thorium concentration by another 1/4300 per day. The U255 inventory rises to 826 kg and then falls, at the end of elght months, to 587 kg. At this time, the processing rate is increased to 1/240 per day (eight-month cycle), but the thorium is returned to the core and the thorium concentration falls thereafter only by burnup. It may be seen that the U235 inventory creeps up slowly and that the regeneration ratio falls slowly. The increase in U255 inventory could have been prevented by withdrawing thorium at a small rate; however, the regeneration ratio would have fallen somewhat more rapidly, and more U235 feed would have been required to compensate for burnup. 2. FOMOGENEOUS REACTORS FUEIED WITH U=~ Uranium-233 is a superior fuel for use in.molten-fluoride salt reactors in almost every respect. _The fission croes seetion in the intermediate range of neutron energies is greeter than the fission cross sections of 0255 and 239 Thus initiel eritical inventories are 1ess, and 1ess edditionel fuel is required to override poisone. Aleo, the peresitic oross section is sub- ;stantially less, end fewer neutrons are lost to radietive capture. Further,.,r 37%the radiative captures result in the immediete fbrmation of a fertile isotope,,~ .71125 e rate of eecumulation of U2 is orders of’megnitude emaller then ' iwith U 35 es a fuel, and buildup of Np237 and Pu 39 is negligib1e¢.fi,r' The mean neutron energy is rather nearer to thermel in these reactors than it is in the correeponding UEBS cages, COnsequently, ‘losses to core vessel and to core salt tend to be higher. Both losses will be reduced sub- stantlially at higher thorium concentrations. | - 191 - 2.1 Initial States Results from s parametric study of the nuclear characteristics of two— region, homogeneous, molten-fluoride-salt reactors fueled with U 233 are giVen in Tsble k.5. The core diameters considered range from 3 to 10 ft, and the thorium concentrations range from 0.25'to'1 mole %. Although the regeneration ratios are less than unity,.they are very good compared with those obtained with U255. With 1 mole % ThF), in an 8-ft-dis t::ore,‘the.’l]‘?j'3 inventory vas | only 196 kg, and the regeneration ratio vas 0.91. ~ - ) ' The regeneretion ratios and fuel inventories of reactors of various diameters containing 0.25 mole % thorium gnd . fueled'with U235 or U 53 compared in Fig. h.ll. The superiority of U235 is obvious. 2.2 Intermediate States Calculations of the long-term performance of one reactor (Case 51, stle 4.5) with U233 as. the fuel are described below. The core diaemeter used was 8 £t end the thorium concentration was 0.75 mole % The changes in inventory of U253 end regeneration ratio ate listed in Table 4.6. During the first year of operation, the inventory rises from 129 to 199 kg, and the regeneration ratio falls from 0.82 to 0.Tl. - If the reprocessing required to hold the concentration of fission products and Np 57 constant is begun at this time, the inventory of U235 increases slowly to 247 kg and the regeneration ratio rises slightly to 0.73 during the next 19 years. This constitutes a substantial improvement‘over the performance with U235 3 'HOMDGENEOUS REACTORS FUELED WITH PLUTONIUM | It may be feasible to burn plutonium in molten-fluoride-salt resotors. The solubility of PuF5 in mixtures of- IiF'and BeFé is eonsiderably less then . that of UF), but is reported to be over 0.2 mole %,10 for eriticality even in the presence of fission fragments and nonfissionable isotopes of plutonium but prdbdbly 1imits ‘severely the amount of ThFh that ean | 0 .Beryllium Fluoride (in preparation), ORNL (1958). 0192 = which may be sufficient Barton; C."J.;" Bolublldty and’ Sts.hil:lty of PuFz in Fused Alkali Fluoride- o Teble 4.5. Nucleer Characteristics of Two-Region, Hamogeneous, Molten-Fluoride=-Salt Reactors Fueled with U 3 " Core diemeter: 8 ft - Externel fuel volume: 339 £t Total power: 600 Mv (heat) Load factor: 0.8 Case rumber y1 b2 'h3 Lk 45 Fuel and blanket selte* 1 1 1 1 1 Core diameter, ft 3 L 4 5 6 TuF), in fuel salt, mole % 0 o 0.25 0 0.25 U235 in fuel salt, mole % 0.592 0.158 0.233 0.106 0.048 U2 atom density¥* 21,0 6.09 8.26 3,75 1.66 Critical mass, kg of U-2° 64,9 22,3 30.3 26.9 20.5 Critical inventory, kg of U253 1620 248 337 166 82.0 Neutron absorption ratios¥¥¥ _ 2 (fissions) - 0.875% 0.8706. 0.8665 0.8725 0.881k U7 (n-7) | 0.1246 0.129% 0.1335 0.1275 0.1186 Be~Li-F in fuel salt 0.0639 0.1061 0.0860 0.1h72 0.3180 Core vessel 0.0902 0.1k0L 0.1095 0.1380 0.1983 Li-Be-F in blanket salt 0.,0233 0.023% 0.0205 0.0196 0.02L5 Leakage | 0.0477 0.0310 0.0306 0.0195 0,0160 Th in fuel salt © 0,1095 0.1593 Th in blanket salt 0.9722 0.8857 0.8193 0.7066 0.6586 - Neutron yield, n S22 2,19 2.18 2.19 2.21 /"’f:MEdian fission energy, ev r{fg,,€f }?ifh'i;fi”felu:——_ 19 2.9 0.35 . Themsal fissions, % . 0,053 8.0 . 23 16 38 ‘fig;;n-y capture to fission ratiO,:__gfi';fifO,lhrt'f 67157 0.15 0.15 - 0.13 -:-E;aRegeneration ratio Z,iy-;jjfi¢3: '-;97r,'57 O 89 .93 0.37 - 0.66 193 % Fuel'salt No. 1 3limole % BeFo + 69 mole % LiF + UFy + TuF), - . Blanket salt No. 1: 25 mole % 'm‘u + 75 m°1e % ur 7'l¥*;§Atoms (x 10 19)/¢m ‘ | o 5f*** Neutrons absorbed per absorption in 0233 Teble 4.5 {continued) . Case number 46 47 48 Lg Fuel and blanket salts¥ T 1 1 1 Core diameter;"ft ' 6 8 _ 8 10 - ThF) in fuel salt, mole % O 0.25 0.25 1 0.25 U233 in fuel salt, mole % 0.066 0.039 0.078 0.031 U2 atom density®* 2.36 150 2.95 1.10 Critical mass, kg of U-2° 29.2 M.l 86.6 63.0 Criticel inventory, kg of U2 17 931 196 104 Neutron absorption ratios*¥¥ , U%?° (fissions) 0.8779 0.8850 0.8755 0.888L v°2? (n-7) 0.1221 0.1150 0.1245 0.1119 Be-Li-F in fuel salt 0.2297 0.3847 0.1899 0.5037 Core vessel 0.1508 0.1406 0.0778 0.1168 Li-Be-F in blanket salt 0,0179 0.0141 0.0095 0.0108 Leakage 0.0157 0.0095 0.0090 0.0068 Th in fuel salt 0.1973 0.2513 0.5768 0.2852 Th in blanket salt 0.5922 0.h421l 0.334h 0.3058 Neutron yleld, 1 . 2.20 2,22 2.20 2.23 Median fission energy, ev 1.2 0.20 1.1 50% Th Thermal fissions, % | 29 43 ol 50 n-y capture-to-fission ratio, « 0.14 0.13 0.1kL 0.13 . Regeneration ratio ' 0.79 0.67 0.91 0.59 131 - 216 10 0.063 2,29 0.8781 0.1219 0.2360 0.0629 0.0071 0.0065 0.6507 0.2408 2.20 3.2 30 0.1k 0.89 51 10 0.75 - 0.0597 1.97 58.8 129 0.8809 0.1191 0.2458 0.1168 0.0187 - 0.0050 0.4903 2.21 0.68 34 0.1k 0.82 *' Fuel salt No. 1: 31 mole % BeF, + 69 mole % LiF + UF), + ThF) Blanket selt No. 1: 25 mole % TuF) + 75 mole % LiF Fuel salt No. 2: 37 mole % BeFp + 63 mole % LiF + UF) + ThF Blanket salt No. 2: 13 mole % ThF), + 16 mole % BeF, + T1 mole % LiF ** Atoms (x 10"19)/cm5. ¥¥%X Neutrons absorbed per absorption in U23j. i -194~ UNCLASSIFIED ORNL~-LR~-DWG 24923R {.0 233 /U ' /..Z. Co0.8 — - < o Z l o ¢ - U235 < & ‘ _.__.L/ W \ > o L o 0.4 — o 200 400 600 ~ CRITICAL INVENTORY (kg of U) Fig. 4.11. Comparison of Regeneration Ratios in Molten-Salt Reactors - - Containing 0,25 Mole % ThF4 and 'U_235' or U233 Enriched Fuel. 195 Teble 4.6. Nuclear Performence of a Two-Region, Hqgggeneous, Molten-Fluoride-Selt Reactor Fueled with U-2 end Containing 0.75 mole % ThF) in the Fuel Salt Core diameter; 8 ft External fuel volume: 339 ft3 Total power: 600 Mw (heat) Load factor: 0.8 Initial State _ After 1 year _ Inventory Absorptions Fissiocns Inventory Absorptions Fissions (kg) (%) (%) (kg) (%) (%) Core elements Th-232 1,572 22,2 1,572 19.1 Pa-233 9.4 0.5 U-233 129 45,2 100 199 k5.3 99.5 U-234 | 23.3 0.9 . : U-235 1.9 0.3 0.5 U-236 0.1 0.1 Np-237 ’ U-238 Pu-239 Fission fragments 181 7.9 Li-6 3,920 6.5 . 3,920 3L Be-9 3,00l 0.8 3,008 0.7 F-19 2k,000 4,0 2k ,000 3.5 Blanket element U-233 8.6 Total fuel 129 210 U-233 Feed Rate, kg/day 0.790 0.370-0.189 Regeneration Ratio 0.82 0.71 -196- o Taeble 4.6 (continued) After 2 years Inventory Absorptions (kg) . (%) Fissions (%) After 5 years Inventory Absorptions Fissions (kg) (%) (%) Core elements Th-232 Pa-233% U-233 U-23k U-2355 U-236 Np-237 U-238 Pu-239 Fission fragmefits 1i-6 Be-9 F-19 Blanket element U-23%3 Total fuel 1,572 18.9 9.0 0.5 20k L9 98.5 L4,0 1.7 Solt 0.8 1.5 0.6 0.3 0.1 0.1 1817_ TeT 3,920 33 3,008 - 0.6 2k, 000 3.l 10.7“ 220 1,572 18.3 8.9 0.k 276 43,7 95.6 89 31 17.7 2.3 LY h.2 0.2 0.5 0.1 0.3 181 7.2 3,920 2.8 3,008 0.6 24,000 3.3 16.2 050 . kefday U-233 Feed Rate, - .o - Regeneratlon Ratlo. 0.181 197 Teble 4.6 (_Cdntimzed)" | After 10 years _ After 20 years Fissions Inventory Absorptions Fiseions " ‘Inventory Absorptions (ka) (® - (%5 (k) (%) (%) Core elements s : | o - Th-232 1,572 17.8 1,572 17.2 Pa-233 8.6 ok S8 0.k U-233 2% k2.5 92.8 247 k1.5 9.5 U-234 132 b2 12 5.0 | U-235 32.5 BT 1AW 1.8 9.0 . U-236 12.5 0.6 2k 1.1 Np-237 1.7 0.2 3. 0.3 U-238 1.7 0.1 5.1 0.3 Pu-239 0.2 0.1 0.1 - 0.8 0.3 0.5 Fission. fragments 181 6.7 - 181 6.3 Li-6 3,920 2.5 3,920 2.1 Be-9 3,008 0.6 3,008 0.6 F-19 2,000 3.3 24,000 3.3 Blanket element U-233 22.2 31.6 Total Fuel 282 295 U-233 Feed Rate, | . kg/day 0.171 0.168 Regeneration Ratio 0.73 0.73 .198- the ingrowth or fission pruducts will neaessitate the addition of more Pu be added to the fuel s'alt.' Th:ls limita.tion, coupled with the condition that 259 is an inferior fuel in intermediate reactors, will result in & poor neu- | tron economy in camparisen with that of 0255-fueled reactors. Ebwever, the advantages of handling‘plutonium in e fluid fuel system.may'make the plutonium- . fueled.molten-salt reactor more desir&ble than other possible rlutonium- ‘burning systems. 3.1 Initial States Critlcal Concentration, Vass, Inventoryj“and Regeneration Ratio. The‘results of calculations of a plutonium-fueled reactor having & core dlameter of 8 £t and no thorium in the fuel salt are deseribed belcw;, The eritical eoncentratiofi was 0.013 mole 4 PuF3 which 1is an order of magnitude smaller ‘than the solubility limits in the fluoride salts of interest. The critical mass was. 13, 7 kg and the critical inventory in a 600-Mw system (339 ft5 of external fuel volume) vas only 3l.2 kg. : The core wes surrounded by the Ii-Be-Th-fluoride blanket nixture No. 2 (15 per cent ThFL) 8lightly more than 19% of all neutrons were captured in the thorium to give a regeneration ratio of 0.35. By employing smaller cores . and larger investments in Pu 2% , however, it should be posaible to increase the regeneration ratio substantially. Neutron Balance and Miscellaneous Detalls. Details of the neutron econony of a reactor fueled with plutonium are given in Tsble 4.7, Parasitic captfires in Pu?’ are relatively high; n is 1.8k, campared.with e v of 2.9. The neutron .spectrum is relatively soft; almost 60% of all fiasions are caused'by thermal .neutrons, and, as a reault, absorptions 1n 1ithium are high. z, 2 Intermediate States On the basis of the averaga value of a of Pu 59, it is estimated that Puaho will accumnlate in the system wntil it capturen, at eqnilibrium, about rhalf as many neutrons as Pu 39. While these captures are not wholly parasitie, -2kl -inasmuch as the product, qu s 1s fission&ble, the added campetition for neu- trons will necessitate an- 1ncrease in the eoncentration of the 39. Iikewise, 239 Further, the’ rare earths among the fiseion products may exert a cammon-ion influence on the plutonium end reduce ite solubility. On the credit side, .na'199~ TABIE 4.7 Initial State Nuclear Characteristics of & Twpical ‘Molten-Fluord de-Salt Reactor Fueled with Pu 239 Core Diameter: External fuel volume: Total power: Ioad factor: Critical inventory: Criticael concentration: 8 £t 5 339 £t 600 Mw (heat) 0.8 239 31.2 kg of Pu’ 219 0,013 mole % Neutrons Absorbed per Neutrons Absorbed in Pu 39 . Neutron Abéorbers 0.6%0 239 (fissions) 259 (n-7)" 0.372 116_and_1i7 in fuel salt 0.202 Be? in fuel salt 0.022 2 1n" fuel salt 0.086 Core Vessel - 0.086 Th in blanket salt 0.352 Ii-Be-F in blanket salt 0.02h fRéactor vessel 0,00k Ieakage - 0.003 NEutron Yiéid,:nA-\ 1.éh Thermal Fissions, % 59 . 0.352 Regeneration Ratio « 200 = = 2 » :"qi however, is the 0235 produced in the blanket. If this is added to the core it mey compensate for the ingrowth of Pu2h0 and reduce the Pu?39'requirement to below the solubility limit, and 1t may be possible to operate indefinitely, es with the U255-erled reactors. L. HETEROGENEOUSV GRAPHITE-MODERATED REACTORS The use of a moderator in a heterogeneous lgttice w1th molten selt fuels is potentially advantageous. First, the approach to & thermal neutron spectrum improves the neutron yleld, q, attainable, especially'with U235 and Pu 39 Second, in a heterogeneous syetem, the fuel is partially shielded from’neutrona of intermediate energy, and a further improvement in effective'neutron yield, 1, redults. Further, the optimum systems may prove to have smeller volumes of fuel in the core than the corresponding flnorine-moderated homogeneous reactors and, consequently, higher concentrations of fuel and thorium in the melt. This may substantially reduce parasitic losses to components of the carrier salt. On the other hand, these higher cancentrations tend to increase the inventory’in the eirculating fuel system externsl to the core. The same considerations apply'to fission products and to nonfissionable isotOpes of uranium. Possible moderators for molten-salt reactors include beryllium, BeO, and 'graphite._ The design and performance of the Aircraft Reactor Experiment, a ‘beryllium oxide moderated eodium-zirconium fluoride salt one-region, U235 fueled burner reactor,has been reported (see Part 1) Since beryllium and BeO and molten salte are ‘not chemically compatible, it was: neceSSery to line the fuel circuit with.Inconel. "It is easily estimated that the presence of Inconel, 'or any other prospective containment metel in & heterogeneous, thermal reactor would seriously'impair the regeneration ratio of & conyerteerreeder " Conse- quently, beryllium and Beo are eliminated from consideretiona | | Preliminary evidence indiCates thet uraniumJBearing molten salte may be "compatible with some grades of grephite and that the presence of ‘the graphite’ will not carburize metellic portions of the fuel circuit seriouely.11 It 11K9rtesz, F., Private Communieation, ORNL (1958) - 201 = e i e, et M v s P therefore becomes of interest to explore the capsbilities of thefgraphitee , moderated systems. The principal independent variables of interést.are-the 5 -~ core diameter, fuel channel diameter, lattice specing, and thorium cbncentrfiti@h. k.1 Initial States The nuclear'parameter study of graphite-moderated reactors has‘just begun _and’dnly two cases have been calculated. For convenience in comparison with - the IMFR, these first two "MSFR" calculations vere based on essentially the same geometry and graphite-to-fluld ratio as thdse of the referéhce design'IMER;la with molten salt substituted for liquid.metal.» One calculation was performed with bismuth instead of salt as & check point. The three cases are sumarized in Table 4.8, ' o _ 2Babcock and Wilcox Co., Iiquid Metal Fuel Reactor, Technical Fbasibility Report, BAW-2 (Del) (1955). | “ - 202 = TABIE 4.8 Camparison of Graphite-Mbderated.MbltenLSalt | MR MSFR-1 MSFR-2 Totael power, Mw (heat) 580 600 600 Over-all radius, in. ' 75 T5 T2 Critical mass, kg of U-233 - - -~ 9.9 9.6 27, 7 Critical inventory, kg of U-233%% U867 T7.8 213 Regeneration ratio | 1.107 0.83 Y. 07 Core o | Redius, in. 35 -3 - 34,8 Graphite, vol % | g L5 5 s Fuel fluid, vol % | 55 55 55 Fuel components, mole % ~ Bi ~ 100 IiF 69 61 BeF2 31 36'5 ThF), 2.5 ‘Unmoderated blanket Thickness, in. 6 6 13,2 Composition, mole % . ' Bl Q0 Th 20 (Th) 10 (ThF,) 13 (ThFh) LiF _ 70 71 BeFa : 20 - 16 - ~ Moderated blanket ' Thickness, in. e 36 36 ol Composition, vol % | : Graphite . . 666 66.6 100 - Blenket fluidg®x . 33,4 334 NEutron dbsorption ratio*** - " . " Thin fuel fluid LT e 04566 . U-233 in fuel fluid ST 04918 ';c 925 © . 1,000 " Other components of fuel fluid - 0.081 C0.328 0 0.106 fTh in blenket fluid .. 1.110 -~ 0.825 0.4 “U-255 in blanket" fluid L 0.083 0.071 , ‘Other camponents of blanket fluid 0.040 0,092 . 0.03%8 . Ieaksge ... .. 0.012 ~ 0.008 0.01k 'NEutron yield, q ¢:'3 _ .}-,_1? [f__.__ . 2;2 e 2.28 . 2421 ¥ With bismuth, the external volume indicated in ref 10 was used. The molten salt systems:are caleculated for 339 ft5 external volumes. ¥¥ Dame as ummoderated blanket fluid. *¥¥% Neutrons absorbed per neutron absorbed in.U-253. ..203 - PART 5 EQUIPMENT FOR MOITEN-SALT REACTOR HEAT TRANSFER SYSTEMS The equipment required in the heat transfer circuits of a molten;salt reactor consists of the components needed to contain, circulate, cool, heat, and control molten salts at temperatures up to 1300°F. Included in such systems are pumps, heat exchangers, piping, expansion tanks, storage vessels, valves, devices for sensing operating variables, and other auxiliary equipment. Pumps for the fuel and blanket salts differ from standard centrifugal pumpé for ‘operation at high temperatures in that provisions must be made to exclude oxidants and lubricants from the salts, to prevent uncontrolled escape of salté,and gases, and to minimize heating and irradiation of the drive motors. Héat is transferred from both the fuel and the blanket salts to sodium in shell- and-tube heat exchangers designed to maximize heat transfer Per unit volume and to minimize the contained volume of salt - especially the fuel salt. Seamless piping is used, where possible, to minimize flaws. Thermal expansion is accomodated by prestressing the pipe and by using expansion loops and joints. Heaters and thermal insulation are provided on all components that contain salt or sodium for préheating and for maintaining the circuits at temperatures above the freezing points of the liquids and to minimize heat loésgs. Devices are provided for sensing flow rates, pressures, temperatures, and liquid levels. The devices include venturi tubes, pPressure trfinsmitters, thermocouples, electrical probes, and floats. Inert gases are used over free- liquid surfaces to prevent oxidation and to apply appropriate base pressures for suppressing cavitation or moving liquid or gas from one vessel to another. | The deviations from standard practice required to adapt the various com- ponents to the molten-salt system are discussed below. The schematic diagram of a molten-salt heat transfer system presented in Fig. 5.1 indicates the | relative positions of the various components. For nuclear oPeration,‘an off~ gas system is éupplied, as described in Part 1. The vapor condensation trap indicated in Fig. 5.1 is required only on systems that contain Zth or a com- parably volatile fluoride as a component of the molten salt. - 20k - - UNCLASSIFIED ORNL—LR—DWG 24291 /PUMP (INCLUDES EXPANSION VOLUME) 5 VAPOR TRAP LIQUID LEVEL INDICATING DEVICE—-—I | 7 3% HIGH POINT |- - - - ~— - - SLOPE b \ HEAT I ngaT SOURCE , oo EXCHANGER O—= - L COOLANT SYSTEM FLOW MEASURING DEVICE LOW (VENTURI) POINT —t ] > o EE;ESSURE MEASURING DEVICES SHUT-OFF VALVE , \% ' - : - VENT VENT LIQUID LEVEL FILL AND DRAIN LINE INDICATING DEVICE _VAPOR TRAP ' iy " ' EQUALIZER VALVE X DUMP TANK ———— PRESSURE | ~ REGULATORS —= _ INERT GAS SUPPLY Fig. 5.1. A Molten-Sait Heat Truhsfer System. 205 1. PUMPS FOR MOLTEN SALTS Centrifugal pumps with radial or mixed-flow types of impeller have been_ used successfully to eclirculate molten-salt fuels. The units built thus far and those currently being developed have a vertical shaft which carries the impeller at its lower end. The shaft passes through & free surface of liquid to isolate the motor, the seals, and the upper bearings from direct contact - with the molten salt. Uncontrolled escape of fission gases or entry of unde- sirable contaminants to the cover gas above the free-liquid surface in the pump are prevented either by the use of mechanical shaft seals or hermetic enclosure of the pump and, if necessary, the motor. Thermal and radiation shields or barriers are provided to assure. acceptable temperature and radia- - tion levels in the motor, seal, and bearing areas. Liquid cooling of internal pump surfaces is provided to remove heat 1nduced by gamma and beta radiation. The principles used in the design of pufips for normal liquids are appli- cable to the hydraulic design of a molten~-salt pump. Experiments have shown that the cavitation performance of molten-salt pumps can be predicted from tests made with water at room temperature. In addition to stresses induced by normal thermal effects, stresses due to radiation must be taken into’ account in all phases of design. The pump shown in Fig. 5.2.was developed for 2000-hr durability at very low irradiation levels and was used in the'Aircraft Reactor Experiment for circulating molten salts and sodium at flow rates of 50 to 150 gpm, at heads up to 250 ft, and at temperatures up to 1550°F. These pumps have been vir- tually trouble-free in operation, and many units in addition to those used in the Aircraft Reactor Experiment have been used in developmental tests of various components of molten-salt systems. The bearings, seals, shaft, and impeller form a cartridge-type subassembly thatris-removable from the pump tank after opening a single, gasketed Jeint a- bove the liquid'levelQ The volute, suction, and discharge connections form parts of the pump tank subassembly into which the removable cartridge is in- serted. 'The,upper portion of the shaft and a toroidal area in the lower part of the.bearing housing are cooled by circulating oil. Heat losses during operation are reduced by thermal insulation. | - 206 - ‘(3 20T . . 7 3 > - ‘ UNCLASSIFIED ORNL.LR-DWG. 6218R DRIVE SHEAVE UPPER SEAL OIL RETAINER - UPPER SEAL LEAKAGE DRAIN BEARING CLAMP RING SLEEVE BEARING HOUSING S SPACER SLEEVE TOP SHAFT SEAL ASSEMBLY SHIELD ‘ SEAL FACE RING OlL INLET BEARING (UPPER) _INTERNAL SHAFT COOLING B.H. BREATHER LUBE OiL. GUIDE TUBE HEATED GAS VENT AND LIQUID INJECTION NOZZLE O RESISTANCE HEATER CONNECTION LAVA SPACER BEARING SPACER BEARING (LOWER) . ~SEAL RING OVERFLOW e b e L L 'SEAL FACE RING | ‘ © L w o SUNGER . LOWER SEAL ASSEMBLY— o " SPACER AND HEAT DAM OIL :DRAIN HEADER RING RESISTANCE HEATER CONNECTION DRAIN ) COOLANT CONNECTION * PUMP BODY ASSEMBLY. L FLANGE ~ - CLAMP THERMOCOUPLE. GLAND——— = PUMP TANK COVER FLANGE STILLING WELL LIQUID LEVEL DISCHARGE ELBOW' PUMP TANK LEVEL INDICATOR FLOAT FLEXIBLE CONNECTOR IMPELLER HOUSING VARIABLE INDUCTANCE LEVEL INDICATOR ANTISWIRL INLET BAFFLE DISCHARGE LEG Fig. 5.2. Sump-Type Centrifugal Pump Developed for Use in the Aircraft Reactor Experiment, A In a2ll the units bullt thus far nickel-chrome alloys have been used in _the construction of all the high-temperature wetted parts of the pump to minimize corrosion. The relatively low thermal conductivity and high strength of such alloys permitted close spacing of the impeller and bearings and high thermal gradients in the shaft. ' ' Thrust loads are carried at the top of the shaft by a matched pair of | pfeloaded angular-contact ball bearings mounted face-to-face in order to pro- vide the'flexibility required to a#oid.bindings and to accomodate thermsl distortions. Either éingle—row ball bearings or a journal bearing cafi be used successfully for the lower bearing. | The upper lubricant-to-air and the lower lubricant-to-inert-gas seals are similar, rotary, mechanical face-type seals consisting of a stationary graphite member operating in contact with & hardened-steel rotating member. The seals are olil-lubricated, and the leakage of oil to the process side is approximately 1 to 5 cmB/day. This oil is collected in a catch basin and removed from the pump by gas-pressure sparging or by gravity. The accumulation of some 200,000 hr of relatively trouble-free test operation in the temperature range of 1200 to lSOOoF with molten sglts and liquid metals.as the circulated fluids has proved the adequacy of this basic pump design with regard to the major prdbiem of thermally induced distortibns. Four différent sizes and eight models of pumps have been used to provide flows in the range of 5 to 1500 gpm. Several individual pumps have operated for periods of 6000 to 8000 hr, consecutively, without maintenance. 1.1 TImprovements Desired for Power Reactor Fuel Pump ; The bdsic.pump described above has bearings and seals that are oil-lubricated and cdoled, and in some of the pumps elastomers have been used as seals between parts. The pump of this type that was used in the ARE was designed for a rela- tively low level of radiation and received an integrated dose of less than 5 x 108 r. Under these conditions both the lubricants and elastomers were proved to be entirely satisfactory. The fuel pump for s power reactor, however, must last for many years. The radiation level aenticipated at the surface of the fuel is 105 to 106 r/hr. Beta- and gamma-emitting fission gases will permeate gll aveilable gas space above the fuel, and the daughter fission products will be depoéited on all exposed surfaces. Under these conditiofis, the simple pump described ebove would fail within a few thousand hours. | - 208 - C Considerable improvement in the resistance of the pump and motor to radiation can be achieved by relatively simple means. Iengthening the shaft between the impeller and the lower motor bearing and inserting additional shielding material will reduce the radiation from the fuel to a low level at the lower motor bearing and the motor. Hbllofi, metal O-rings or another metal gasket arrangement can be used to replace the elastomer seals. The sliding seal just below the lower motor bearing, which prevents escape of the fission-preduct gases or inleakage of the outside atmosphere, must be lubricated to ensure continued operation. If oil lubrication is used,' radiation may quickly cause coking. Various phenyls, or mixtures of them, are much less subject to formation of gums and cokes under radiation and could be used as lubricant for the seal and for the lower motor beariné; This bearing would be of the friction type, for radial and thrust loads. ' These modifications would provide a fuel pump with an expected life of_the . order of a year. With suitable provisions for remote maintenance and repair, these simple and relatively sure improvemente would probably suffice for power reactor operation. ' Three additional improvements, now being studied, should make possible a fuel pump that will operate trouble-free throughout a very long life. The first of these is a 'pilot bearing for operation in the fuel salt. Such a bearing, whether of the hydrostatic or hydrodynamic design, would be com- pletely unaffected.by radiation and would permit use of a long shaft so that - the motor could be well shielded. A combined radial and thrust bearing just below the motor rotor would be the only other bearing required. The second improvement is a 1abyrinth type of gas seal to prevent escape of fission gases up the sheft. There are no rubbing surfaces and hence no ‘need for lubricants,l g0 there can be no radietion demege. “The third innovation is a hemispherical gas-eushioned bearing to aet as ‘8 combined thrust end radial bearing. It would heve the advantage of requiring no auxiliary lubrication supply, and it would cedbine well. with the 1ebyrinth type of gas seal. It would, of couree, be unaffected.by redietion, _;' : 1. 2 A Propoeed Fuel Pump A punp design embodying these 1ast three feetures is shown in Fig. 5.3, It is designed for operation at a temperature of 1200 F,Aaiflow rate of 24,000 gpm, and s head of TO ft of fluid. The lower bearing is of the hydrostatic type and is - 209 - - o _ Fig. 5.3. Improved Molten-Salt Pump Designed for Power Reactor Use. Operating Temperature, 1200°F; Flow Rate, 24,000 gpm; Head, 70 ft of Fluid. 210 eI lubricated by the molten-salt fuel. The upper bearing, which is also of the hydrostatic type; is cusfiiohed‘by helium and serves also as a barrier against passage of gaseous fission products into the motor. This bearing is hemi- spherical to permit accomodation of thermally induced distortions in the over-all pump structure. | The principal radiation shielding is that provided between the source and the area of the.motor windings. Iayers of beryllium and boron for neutron . shielding and a heavy metal»for gamma radiation shielding are proposed. The motor is totally enclosed in order to eliminate the need for a shaft seal. A coolant is circfilated in the area outside the stator windings and between the upper bearing and the shielding. Molten-salt fuel is circulated over the sur- faces of these parts of the pump which are in contact with the gaseous fission products to remove heat generated in the metal. 2. HEAT EXCHANGERS, EXPANSION TANKS, AND DRAIN TANKS The heat exchangers, expansion tanks, and drain tanks mfist be‘especially designed to fit the particular reactor system chosen. The design data of items suitable for a specific reactor plant are described in Part 1. The special problems encountered are the need for preheating all salt- and sodium-containing components, for cooling the exposed metal surfaces in the expansion tank, and for removing afterheat from the drain tanks. It has been found that the molten salts behave as normal fluids during pumping and flow and that the heat transfer coefficients can be predicted from the physical properties of the salts. 3. VALVES The prdblems associated with valves for molten-salt fuels are the consis- 'tent alignment of parts during transitions from rocm temperature to 1200° F, the selection of materials for mating surfaces which will not fusion-bond in the salt and cause the valve to stick in the closed position, and the pro- vision of a gas-tight seal.; Bellcws-sealed, mechanically operated, poppet valves of the type shown in Fig. Se A have given reliable service in test systems. A number .of corrosion- and fusion-bond-resistant materials for high- temperature use were found_fihrough exfehsive screening tests. Molybdenum "~ UNCLASSIFIED ORNL—-LR-DWG 21292 1 SYSTEM CERMET SEAT AND POPPET BELLOWS PROTECTOR BELLOWS DUMP TANK c Fig. 5.4, Bellows-Sealed, Mechanically Operated, Poppet Valve for Molten=Salt ervice. ' 212 against tungsten or copper and several titanium or tungsten carbide-nickel cermets mating with each other proved to be satisfactory. Valves with very accurately machined cermet‘seats and poppets haue operated satisfactorily in 2-in. molten-salt lines at 1300°F with leakage rates of legs than 2 cmB/hr. Consistent positioning of the poppet and seat to assure leak tightness is achieved by minimizing transmission of valve body distortions to the valve stem and poppet. ' ' If rapid valve operation is not required, a simple "freeze" valve may be used to ensure a leak-tight seal. The freeze valve consists of a section" of pipe, usually flattened, that is fitted with a device to cool end freeze e salt plug and another means of subsequently’heating and melting the plug. L. SYSTEM HEATING Molten-salt systems must be heated to prevent thermal shock during fill- ing and to prevent freezing of the salt when the reactor is not operating to ‘produce power. Straight pipe sections are normally heated by an electric tube-furnace type of heater formed of exposed Nichrome V wire in a ceramic shell (clamshell heaters). A similar type of heater with the Nichrome V wire installed in flat ceramic blocks can be used to heat flat surfaces or large components, such as dump tanks, etc. In general, these heaters are satis- factory for continuous operation at 1800°F, Pipe bends, irregular shapes, and small components, such as valves and.pressureameasuring devices, are usually heated with tubular heaters (e.g., General Electric Company "Calrods") which can be. shaped to Fit the component or pipe bend. In general, this type of heater should be limited to service at 1500 F. Care must be exercised in the installation of tubular heaters to avoid failure due to a hot spot caused by insulatioh in direct contact with the heater. This type of failure can be avoidednby installing a thin sheet of metal (shim stock) between the heater ' and the insulation. ," Direct resistance heating in which an electric current is passed directly :,through 8 section of the molten-salt piping has also been used successfully. Operating temperatures of this type of heater are 1imited onlyiby the corrosion and strength 1imitations of the metal as the temperature is increased. Experi - ence has indicated that heating of pipe bends by this method is usually not uniform and can be asccompanied by hot spots caused by nonuniformity of liquid flow in the bend. - 213 w 5. JOINTS Failures of‘some system conponents may be expected during the desired ' operating life, say 20 years, of a molten- salt power-producing reactor, con- sequently, provisions must be made for servicing or removing and replacing such_components., Remotely controlled manipulations will be required because there will be a high level of radiation within the primary shield. Repair work on or preparations for disposal of components that fail will be carried out in separate hot cell fac1lities. | The components of the system are interconnected by piping, and flanged connections or welded joints may be used, In breaking connections between a component and the piping the cleanliness of the system must be preserved, and in remaking a connection proper alignment of parts must be re-established. The reassembled system must conform to the original leak-tightness sPecifica— tions. Special tools and handling equipment will be needed to separate components from the piping and to transport parts within the highly radio- active regions of the system. While an all-welded system provides the highest structural integrity, remote cutting of welds, remote welding, and inspection of such welds are difficult operations. Speclal tools are being developed for these tasks, but they are not yet generally available. Flanged connections, which are attractive from the point of view of tooling, present problems of permanence of their leak tightness. Three types of flanged joints are being tested that show promise. One is a freeze-flange joint that consists of a conventional flanged-ring joint with a cooled annulus between the ring and the process fluid. The salt that enters the annulus freezes and provides the primary sealy, The ring.provides_ a backup seal against salt and gas leakage. The annulus between the ring and frozen materiasl can be monitored for fission product or other gas leakage. The design of this joint is illustrated in Fig. 5.5. A cast-metal-sealed flanged joint is also being tested for use in vertical runs of pipe. As shown in Fig. 5.6, this joint includes a seal which is cast in place in an annulus provided to contain it. When the connection is to be made or broken the seal is melted. Mechanical strength is supplied'by clamps or bolts. - 21 o - UNCLASSIFIED ORNL-LR-DWG 27895 »' l-*GAP (~Yi6 in) V//’ N} SOFT-IRON OR COPPER SEAL RING 7 ~— AIR CHANNEL FOR COOLING AN FROZEN-SALT SEAL WELD OF FLANGE TO LOOP TUBING —s— SALT FLOW A A RING INSERT TO PROVIDE LABYRINTH FOR SALT LEAKAGE FROZEN-SALT SEAL Z - — NARROW SECTION TO REDUCE HEAT — TRANSFER FROM THE MOLTEN SALT IN THE LOOP TUBING £ A 4 1 Y 0 % 1 e — INCHES 7 © \__ARINLETTO COOLING CHANNEL 'f BETWEEN FLANGE FACES INDICATE REGION OF FROZEN-SALT SEAL; 2222 INDICATE REGION OF TRANSITION FROM LIQUID | TO SOLID SALT Fig. 5.5. Freeze-Flange Joint for 1/2-in.~OD Tubing. 215 UNCLASSIFIED ORNL-LR-DWG 27898 i SEAL MATERIAL INSERT SHOWN BEFORE BEING \\ FUSED TO FORM SEAL { 1/2 0 ‘/2. 1 WELD OF / T INCHES FLANGE TO LOOP TUBING ~” - Fig. 5.6. Cast-Metal-Sealed Flanged Joint. 216 A flanged joint containing a gasket (Fig. 5.7) is the third type of Joint being considered. In this joint the flange faces have sharp, cir- cular, mating ridges. The opposing ridges compress a soft metal gasket to form the seal between the flanges. 6. INSTRUMENTS Sensing devices are required in moltenJSalt systems for the measurement of flow rates, pressures, temperatures, and liquid levels. Devices for these services are evaluated according to the following criteria: (1) they must be of leak-tight, preferably allewelded, construction; (2). they must be capable of operating at maximum tempersture of the fluid system; (3) their acouracies must be relatively unaffected by changes in the system temperature; (4) they should provide lifetimes at least as great as the lifetime of the reactor; (5) each must be constructed so that, if the sensing element fails, only the measurement supplied by it is lost. The fluid system to which the instrument is attached must not be jeopardized by failure of the sensing element. 6.1 TFlow Measurements Flow rates are measured in molten—salt systems with orifice or venturi elements. The pressures developed across the sensing element are measured by comparing the outputs of two pressure-measuring devices. Magnetic flow- meters are not at present sufficiently sensitive for molten-salt service because offthe,poor'electrioal_conduotivitj of the salts. 6.2 Pressure Measurements . - Measurements of system pressures require that transducers operate at g safe margin above the melting point of the salt, and thus the minimum trans- ' ducer operating temperature is usually about 1200° F. ‘The pressure transducers "~ that are available are of two types (1) a pneumatic fOrce-balanced unit and (2) a displacement unit in which ‘the pressure is sensed by displacement of a Bourdon tube or. diaphragm. The pneumatie force-balanced unit has the disadvan- teges that loss of the instrument gas supply (usually air) can result in loss ‘ of the measurement and that failure of the bellows or diaphragm would open the process system to the alr supply or to the atmosphere. The displacement unit, - 217 - 8ic ~ UNCLASSIFIED ORNL-LR-DWG 27899 WELD OF FLANGE TO LOOP TUBING e ey — INCHES ANNEALED COPPER RING | | Fig. 5.7. Indented=Seal Flange. | RAISED TOOTH 4 b O 4 1 on the other hand, makes use of an isolating fluid to transfer the sensed pressure hydrostatically to an isolatedzlow-temperature output element.. Thus, in the event of a failure‘of-the_primary diaphragm, the trocess fluid would merely mix with the isolating fluid and:the closure of the syetem would be unaffected. | o 6. 3 Temperature Mhasurements Temperatures in the range of 800 to 1300 F are commonly measured with Chromel-Alumel thermocouples or platinum-platinum-rhodium thermocouples. The accuracy and life of a thermocouple in the temperature range of interest are functions of the wire size, and, in general, the largest p0381ble thermocouple should be used. Either beaded thermocouples or the newer, magnesium oxide- insulated thermocouples may be used. 6.4 Iiquid-Ievel Measurements Instruments are available for both on-off and continuous level measure- ments. On-off measurements are made with modified automotive-type spark plugs in which 8 long rod is used in place of the normal center conduc¢tor of the spark plug. In order to obtain a continuous level measurement, the fluid head is measured with a differential pressure instrument. The pressure required to bubble a gas into the fluid is compared with the pressure above the liquid to obtain the fluid head. Resistance probe and float types of level-indicators are available for uee in liquid-metal systems. | 6.5 Muclear Sensors , Nuelear sensors for molten-salt reactors are. similar to. those of other reactors and are not required to withstand high temperatures. Existing and 'well-tested fission, ionizatiOn, and ‘oron trifluoride ‘thermal-neutron detec- tion chafibers ere available for installation at all points essential to reactor | ioperation.; Their disadvantages of limited life can be countered only'by dupli- cation or replacement and provisions can’ be’ made for this. It should be ;pointed out that the relatively large, negative, temperature coefficients of reactivity provided by'most circulating-fuel reactors, make these instruments unessential to the routine operation of the reactor. ‘ w 210 w PART 6 BUILDUP OF NUCLEAR POISONS AND METHODS OF CHEMICAL PROCESSING ~ Even though nearly pure 1?2 or 177 is used as the initial fuel - for & reactor, undesireble products quickly build up as & result of the fissioo.process. First, each uranium nucleus thet undergoes fis- 'sion'splits into two "fission product"” elements. The'fission products are‘nuclear poisons to vanying degrees that depend on_their_atamio- _ number and mass-and on the mean neutron‘energy of the reactor. A sec- ond source of poison is the even-numbered isotopes of uranium, which. are not fissioneble. A certain emount of U2 is fed, along with the 177, even in highly enriched uranium, and ‘as the U727, with its high neutron absorption cross sectiofi,»is burned out, the percentage of U238 rises. Simllarly, a certain fraction, a, of the captures in the fissionable isotopes results in radiative ‘capture of the neutrons, and, instead of fissioning, the next higher uranium isotope (UEBLL 256) is formed. It is necessary to examine the rates and extent to which these updesirable constituents build up so that changes in neutron or U economy may be understood and so that desirable chemical reprocessing rates may be determined. 1. BUILIUP OF EVEN-MASS-NUMBER URANIUM ISOTOPES The buildup of U222, 12, 1220, and UP>® as nonfissionsble iso- topic diluents in U233 and U235 is, as stated above, significant in | fuel cycle economics. Although_U232 does not bulld up enough to af- fect the neutron balence significantly, its hard-gamma-emitting daughters are produced fast enough to be a biological hazard in the handling of U255 and thus adversely affect the resale value of the U253.l It has been assumed that & molten-salt reactor will process 1. A. T. Gresky and E. D; Arnold, Products Produced in Continuous Neutron Irradietion of Thorium, ORNL-1817 (Feb. 6, 1956). - 220 - and burn all the U23 > it produces, hence nhe 0232 problem has not been considered in detail.- : | Rediative captures in Pa.25 > and. Ua5 > lead to isotopic contamina- tion of the U= wi‘bh'Uzsl} . With no processing to separate these | isotopes, end none seems feasible , the 0231} builds up until it is be- ing produced and burned at the same rate. Based on cross sections for neutrons at a velocity of 2200 m/sec, as taken from BNL-325 (cf 25 532, a 23 = 0.10, o ih - 92), & thermal reactor at steady state would have ~ 58% as much U234 as U233 with 025 capturing ~9% as many neutrons as 0253 In the epithermal mol‘ben-salt reactor de- scribed in part 1 and hereafter called the reference-design reactor (a 23 ny_ 0.16, o, 25 / o’eu. ~s 4.67), the steady-state 023lP concentra- tion is ~75% of the U255 concentration, with U23 absorptions equal to ~14% of the 023 3 absorptions. Neutron capture in 02 produces fissionable 0255 but capture in thorium would be preferable ) since U25 5 is a better fuel than 0255 Rediative capture in ye?? yields 1125 6 which may be considered ~to be an isotopic poison, since neutron absorption in U23 6 effective~ ly ylelds Np 21 instea.d of a fissionable isotope. In a thermal reac- tor (o7 = 0. 19, a/ = 582, 5-26 s Te5), the v22% yould build up until :I.t was present to the exbent. of ~15 times the amount of 023 > with v 6 capturing ~16% as many neutrons as 2, Normally, in any actual thermal reactor, resonence captures in U25 6 will reduce the steady-state ratio of U0 to 0235 to less then 15. In the epi- thermal reference-design rea.c'bor (a o 0.k, '0’25/“‘ 26_ ~ 1.3), the U‘?5 would build up only to NSO% of the 1123 5 at steady state, but ab that point the 1323 6 would be capturing N}O% a.s many neutrons as U 3 5 . Isotopic separation of Ua3 5 and 023 6 may be feasible in a power rea.ctor economy beca.use of the 1arge amounts involved and be- cause’ 11; is importan-b in a. breeder-converter economy. Separation in a sepa.ra.te ca.sca.d.e would cost at least nine times as much as separa.- tion of 6235 and 02 ) bu‘b by feeding the U 25 5 0256 mixbure into - 221 - existing cascades (either by adding top stages or accepting lower pro- duction rates) less expens.ive processing probably could be achieved. A study of the gaseous-diffusion problem is being made 3 and an . analysis 2,3 of its bearing on the nuclear fuel cycle has been reported At present 3 the government buy-back prices for 1.123 > include the same pena.lty for di- - lution with U23 b and. U23 as for U258 It has been assumed in the ref- erenee-design reactor that U23 build.up must be tolerated. . ‘l‘his is. probably a realistic assumption, since the fuel becomes a mixture of 22 3 %1556, 8. The assumption would be pessimistic, however, for another type of feasible molten-salt reactor which would burn U23 > in the core and which would make 0233 only in the blanket that would be sold externally at a premium price for another molten-salt reactor which would burn only U 235 » since in such a case the IJ23 > ~could be "traded in“ on fresh diffusion plant materia.l when the U25 content warranted. For a steady-state reactor operating with highly enriched 023 5 feed (e.g., 93% 0255 6% 0238 1% U23h) without any isotopic reprocess- ing, the U2‘7’8 at steady-state will capture 6/93 as meny neutrons as the U23 > fed. (as distinguished from the U 235 ‘ouilt up from U 235 via U25h) In a thermal reactor (o’ 25 691+ o, 28 _ 2.73) at steady state there yould be ~16 times as mich 112 as 1255, 1n actual thermal reactors the U238 to-U 255 ratic would not get this la.rge because of resonance captures in 0238 . In the reference-design molten salt reactor (I 5"25/5’2 o 1. 5), the U238 can build up only to Jlo% of the U235 fed (thus in the molten-saltreactor, = is a worse isotopic con- tamihant than "X 4n ‘amount, number of neutrons captured, and in being a poison rather then a fertile ma.teria.l_). For the reference-design 2. E. D. Arnold, Effect of Recycle of Uranium Through Reactor and " Gaseous Diffusion Plant on Buildup of Important Transmutation Products in Irradiated Power Reactor Fuels, ORNL~210k (August 21, 1956). 3« J. O. Blomeke, The Buildup of Heavy Isotopes During Thema.l Neutron Irradiation of Uranium Reactor Fuels, 0RNL-2126 (Jan. 11, 1957) - 222 - : 'In the referenee-design molten—sal'b react.or ebou‘b l% of ’the Pa molten-salt reactor 1t was assumed that 0238 buildup would have to be accepted. The rea.lism or pessimism of thie e.ssumption is about the seme &s discussed for IJE5 in the preceding paragraph 2. PROTACTINIUM AND NEPTUNIUM POISONING Neutron capture in pa’dd or Np239' has the same result as nonfission capture in 1123 5 or Pu 59 i.e., & fiseiona:ble atom is effectively lost, as well as & neutron. Although neutron loss to Np 231 does not involve loss of & fissionsble atom, this loss can be more important than losses to Pa®>> end Np-~” in reactors fed with highly enriched U-~”, Although neutron capture by eny of these three isotopes yields a fertile atom, at present prices for fertlile and fissile materials the gain is negli- gible compared with the loss. The average ratio of neutron captures to beta decays by 1’&'.25 3 in a reactor is given to a good approzd.mation by: | — (Pa) OO’-I-6P(1+O.’)—('T -—(—-T' where P = reactor power level, Mw (thermal), (1 + @) = average ratio of absorptions to fissioms in fissile material, R = regeneration ratio, | M('Ih) - mass of thorium in syetem, K& ’I'he P a.nd o refer to 't.he whole syetem 'I'he other parameters can refer M 'ei'bher to the whole system or 'bo the :r:‘uel e.nd ‘bla.nket sys'hems separately. 233 captures a neu'bron before it can d.ecay to 0255 In a U25 3 025 5 breed.er-eonverl:er reaetor w:l.th highly enriched 1125 5 | 3‘me.keup Np239 poieoning is rela.tively unimporta.nt, ‘but, i the breeding- convereion ra.tio ie poor, the Npe 7 poisen:lng ca.n 'become q_ui“be high 1f 'i'b is hot removed by chemical 'proceseing. 'I‘his :Ls especially 'brue in resonence reectors 5 n which U"25 (en& henee N;p 37, ‘ot stea.dy state) - 223 = yields may be twice as high as in thermel reactors. In the reference- design molten-salt reactor, processed at the rate of once per year, | Np257 polsoning is zero.initially;:dbout'ogs% after one year, about 2% after 20 years, and sbout 2.5% at steady state. 3. FISSION PRODUCT POISONING A 600-Mw reactor operating at a load factor of 0.80 will produce about 183 kg/yr of fission products. About 22 atom percent of these fission products have decay chains such that they appear as krypton or xenon isotopes with half-lives of 78 min or more and thus are sub- Ject to physical removal from a molten-salt reactor as rare gases by purging with helium or nitrogen. These "removable" fission products contribfite'dbout 26% of the total'fission broduct poisoning at 100 ev. (Eor half-lives of 3 min or more, the yield'and poisoning percentages -~ are 30 and 31, respectively. For half-lives of 1 sec or more the velues are Ll and 38%, respectively.) The comparable poisoning per- centage in a thermal reactor is much higher because of ‘the very large thermal neutron absorption cross sectig? of Xe 35.' In a thermal re- actor, however, burnout limits the Xe poisoning to & meximum of about 5%, while in a resonance reactor ad jacent nuclei do not have greatly differing cross sections, and burn-out is relatively ineffec- tive in limiting total poisoning. Thus, to a first approximation, in resonance reactors poilsoning increases almost linearly with time if fission products are not removed. About 26 atom pércent of the long-lived fission products are rare earths. At 100 ev they contribute sbout 40% of the total fis- sion product poisoning. The remaining ru52% of the fission products contributes ~hl% of poisoning at 100 ev and comprises a wide variety of elements, no one of which is outstandlng from the nuclear poisoning point of view. In a thermal-neutron U 255 burner reactor the fission product poisoning, = /2 ,js appro:dmately eq_ual to the equilibrium Xel3 2 | poisoning (0-5%, depending on flux level) plus the equilibrium.Sm?hg - 224 - .poisoning (o1, 2%) rlus the ‘contribution from ell other fission pro- ~ducts. From data.presented in ORNL-2127 (ref. 4) for thermal 0255 burners operated at constant power and constant U‘?35 inventory, with no fission product removel, the poisoning from “all other fission products” is calculated to be ~3% at 100% burnup (i.e., when the total amount of U235'burnea is equal to the "2 inventory), ~19% at 1000% burnup, and ~51% at 10,000% burnup. - Thus it is possible, although not necesserily economicel, to run & thermel, fluid-fueled reactor for many years without processing thé fuel to remove fission product poisons. The penalties for not processing would'be higher U255 inventory charges and lower breeding-conversion ratios. At & load factor of 0. 80, a 600-Mw thermal reactor burns sbout 218 kg of 35 per year (183 kg fissioned, 35 kg converted into U236), and therefore with a L36-kg U-35 inventory the fission product poisoning would incresse from O to 5% initislly end then to 20 to 25% after 20 years. | Even in thermal reactors, resonance captures in fission products make the poisoning somewhat worse than the nubers given above. The megnitude of the extre poisoning depends on the ratio of the neutron flux at resonance energies to the thermal neutron flux, which is de- termined in part by the effectiveness of the moderator. In resonance reactors, the flssion product poisoning is considersbly worse than in thermal reactors because of the higher average fission product ebsorp- tion cross—section relative to U235 " At & load factor of 0.80, a 600-Mw reactor using 100-ev neutrons would burn about 275 kg of 0235 per yeer — (183 kg fissioned, 92 kg converted into U256 - In such e reactor with . an inventory of 550 kg of U25 the fission product poisoning would 'increase approximately 1inearly from zero, initially, to AJ52% after 2 years. S -.EQE'J. 0. Blomeke and.M. F. Todd. Uraniumea'"J | Fission-Product Production &8 & Function of Thermal—NEutron‘Elux, -Trradletion Time, and Decay. - Time, ORNL-2127 (Aug.:.19, 1957). | . 225 - For U233-fueled reactors, the fission product poisoning is about the seme as for U‘?35 in thermal reactors, but in the resonance region | the higher U233 cross section reduces the poisoning effect‘by & factor of 2 compared with U255. Thus & lOO-eV'breeder-converter burning half-" end-half U235 and U235 would have a fission product poison level of ~6% if the fuel were processed at the rate of twice per (100%) burnup. The reference-design molten-salt reactor has & medien fission | energy of n10 ev, with ~10% of the fissions at thermal energles. It may be considered, to a first approximation, that sbout one-third of | the fissions are at thermal energy and sbout two-thirds are at an ener- gy of 100 ev, for comparison'with the analyses presented above. At a losd factor of 0.80, the 600-Mw resctor will bumn 105 kg of U and n:125 kg of ye? per year, and it will produce ~ 183 kg of figsion pro- ducts, m13 kg of u2 , and m3h kg of 11236 It should be processed at a rate of three to four times per 100% burnup, and the total fission product poisoning will be 6 to 8%. L, CORROSION%PRODUCT POISONING Chemical analyses of fuel mixtures circulated in INOR-8 and Inconel loops have indicated that the principal corrosion-product poisons will be the fluorides of chromium, iron, and nickel. These are relatively light elements and,.per atom, their capture cross sections in the reson- ance region are lower than those of the fission products. Further, ex- trapolations of short-time tests indicate that the concentration of the corrosion products will be much lower than that of the fission products. Corrosion produot poisoning'has therefore been neglected. 5. METHODS FOR GHEMICAL PROCESSING The "ideel" reactor chemical processing scheme would remove fission products, corrosion products, Pa233,rnp237, and Np 239 as soon as they were formed. After the Pa 255 and the Np239 had decayed to U233 and Pu 59 they would be returned to the reactor {or sold), along with the urenium and plutonium that would also be recovered in the process. This ideal chemical plant would have low cepital and operating costs, would hold - 226 ~ up only smell amounts of fissioneble eand other high-priced materials, ‘and would discharge its waste streams in forms that could be inex- pensively disposed of or, possibly, sold as by—prodnctssrlPresent technology, however, . does not offer such an ideal pro cess for sny reactor. More practical goals for processing a molten-galt reactor are (1) continuous removel of most of the gaseous fission products by purging the fuel with helium or nitrogen ges; (2) an in-line removal of rare earth, noble metel, or other. fission products by freezing-out part of the selt streem, pleting out fission products on metallic surfaces (either naturelly or electrolyticelly), exchenging the reare earths for Zcerium, or scavenging by contacting the salt with & solid ~such as BeO to remove certa.in constituents of the salt by edsorption : or exnhenge, &nd (3) continuous or batch removal of the salt from the resctor &t an cconomically qptimum.rate to seperate the uranium, plutonium, end selt from the remaining fission products and corrosion_ products by the least expensive method availeble. Present technology does not make all these methods immediately eveildble, but there is . reason to expect that continuation of the current development program would make them eveilsble in the nineteen~sixties. ’ Operation of the Aircraft Reector Experiment end of molten-selt " in~-pile loops have indicated that gaseous fission product removal can be achieved a.n.d theat Ru, Rh, and P4 plate out on metal surfaces. Provisions for degassing nre included in the molteneselt reactor, but et present the possible reduction in fission product poisoning - as a result of plating out in the heat exnhengers and the possible /freduction of corrosion.by formation of & protective surface sre not _f“;being considered in economic studies. Similarly, possible increases “in fission prodnct poisoning as e result of the pdating out of noble - metals in the core ere nnt being teken into eccount. | Mnthods for removing-fission products from.the selt so that the selt can be reused are being inwestigeted in current research and - 227 = development programs, but for the present 'mol‘beh—salt rea;ctoif s'tlidy' it is assumed that only the uranium will be recovered from the salt by the ORNL fluoride volatility process and that the salt will be stored for future recovery of thorium and lithium. Adequate tech- nology already exists for the preparation of fresh fuel stafting with nonradioactive UF6 and fluoride salts, and methods for remote reconstitution of fuel from recycled uranium (snd recycled salt, 1f possible) are being developed. 5+l. The Fluoride Volatility Process A program of development and pilot-plant demonstration of a fluoride volatflity process for recovering uranium from molten fluoride salts by oxidation with F2 to form UF6 is currently under | wey. Recovery of the uranium in the slightly radioactive ARE fuel (Na.F-ZrFu—UF,_l_) was completed in February 1958., and a demonstra‘bion» ) of the recovery of uranium from a highly radioactive STR fuel ele=- ment (which, first, will Be_di’ssolved in NeF-ZrF) in the presence of HF) is scheduled for fiscal year 1959. A blo"ck flow sheet of the process as adapted for molten-salt reactor fuel i'eprocéssing is shown in Fig. 6.1. The uranium-bearing molten salt is trans- ‘ferred to the fluorination vessel in batches. Fluorine, diluted ‘with N , is bubbled through the salt st 450°C until its U content 2 is reduced to ~10 ppm. The UF6, N, and excess F,. pass out of the fluorinator through a 100°C NaF pellet bed, wh:i.charemoves the UF from the gas stream. In the pilot plant the excess F2 is disposed of by scrubbing it with a reducing KOH solution, but in a production Plant the fluorine might well be recycled to the fluorination step. The UF6 is desorbed from the NaF bed by raising the temperature of the bed to LOO°C and sweeping it with more F,-N,. The UF, then - passes through a second NaF bed and finally is collected in cold traps at -40 to -60°C. The volatility plent product is ligquid UF obtained by isolating the cold traps from the system, heating to gbove the triple point, and draining into the product receiver. 228 - 1 in i it et b b 6¢¢ Ru Cold Trop UFg =1 Cold Trap - - R . (Desorption) LiF-BeF,(+U,Th,FP) N from reactor -~ © . 0 L Fluorination - | F, Hold-up Vessel” o Vessel : - ' ~450°C (+Th, >99%FP) ~ to salt recovery, storage, or waste - disposal. UFg = . Product NaF NaF Bed . Bed - NaF Waste Absorption on first bed at 100°C. Desorption through both beds at 100-400°C. ——— F, Disposal Ha UFg — UF, Reactor = Powder Removal System I 1 URs Product Fige 6.1. Proposed Uranium Recovery Flowsheet for Molten Salt Power Reactor — | ! | | | | | } | i 1 | | Y L__J b Sintered Metal ‘Fiiters Ha,F2 ' Dns;iosal Ckhemical Trap - Most of the decontemination in the voletility process is achieved in the fluorination step, since most of the fission end corrosion pro=- ducts remain in the salt. The volatile conteminants (Xe, Kr, I, Te, Mo, most of the Ru, end part of the Nb end Zr) either pass through the NeF ‘bed while the UF¢ is retained or remein on the bed when the UF6 is de- sorbed (Nb, Zr). The I, Te, Mo and Ru are removed almost entirely by a cold trap, and the remainder is scrubbed out of the gas system along fl_with the excess F2 The Xe and Kr follow the N to the plant off-gas- | system. The Nb end Zr slowly'build up on the NeF bed, which is replaced_ vhen poisoned. Replacement is infrequent, however, because a micrometal- lic nickel filter between the fluorinator end NaF bed removes most of the Nb and Zr. | The ARE fuel was processed in batches of 1.k ft3 of salt, each batch containing w10 kg of U-2” (the capaclty of the NaF beds), at the rate of two batches per week. The not-economically-recoverable uranium losses in processing the ARE fuel were approximately 0.015% in the waste salt and O. 075% on the NeF beds. The reference-design molten-galt reactor includes a volatility plant approximately the same size as the ORNL pilot plant. About the seame uranium processing rate is planned (two 10-kg batches per week through the NeF beds), but & higher salt throughput rate (2 £t°/day each of fuel and blanket fluid) will be used because of the lower uranium con- centration in the salt (compared with the ARE salt). This higher salt throughput rate is within the capabilitlies of the present ORNL pilot plant, which could fluorinate one 1.k ft3 batch per shift, if necessary. For the molten-salt reactor processing plent, separate fluorinators of gbout 2 £t5 cepecity each for the fuel and blanket salts will preveht Cross contaminetion and require only one fluorinetion per day for each. The possibility-of continuous fluorinstion and the consequent re- ‘duction in the size of the equipment are being considered. In addition, information needed in the adasptation of the process to the particular fuel and blanket salts to be used in the molten-selt reactor is being - 230 - obtained » @nd simpler means for reconstituting the reactor feed mater- lals are being studied. 5.2. K-25 Process for Reduction of UFg to UFy The continuous, reduétion of highly' enriched. UF6 to _UFM_ is__wgli proved as a nonradioactive proc:eesss.5 The process, used and developed | at K-25, is indicated in Fig. 6.13 the UF product from the volatility process is the feed mate:ial. The reduction takes place in a U"E'6-F2-H2 flame in a Y-ghaped reactor. The Fa is added to give the proper fleme temperature. The reaction products are UF,+ and I-IF-HQ gas. Micrometal- lic filters are used to recover eny UFI;. wvhich may be entrained in the exlt gas. A vibrator 1s used to shake free any UFh which clings to the filter or tower walls. A chemicel trap with a t',‘z?:..'i'xclL or an NaF pellet bed 1s used to recover any unreacted UF6 in the exit gases, although the amount so collected is negligibly small in normal oper- ation. The HF in the exit gas is either scrubbed with a KOH solution spray or sorbed on an NaF bed. | This process has not been used a‘b 8 high level of radioactivity. ,S-:ane the UF6 from the molten-salt reactor volatility plant will be somewhat radiocactive » the major actiflty probebly being II23 T , & pilot plant demonstration will be required, but no serious difficulties are anticipated. For the molten-salt reactor chemicel plant, Vequipment of the same size as that used in the ‘K—-25 facility is assumed. A fa- cility of this Size would have excess capacity on a continuous basis, but it is essumed that it would be operated intermittently, probebly on ;thé once-or-twice-a-week schedule 'fv'.'sed.' for the discharging of the UF from the voletility plent NeF beds. | 5.3. -Salt Recovery by Dissolution in Concentrated HF . Leboratory work ha.s"ti:éefi initiated on & process for. recavering LiF 5. §. H. Smiley and D. C. Brater, "Conversion of Uranium Hexafluoride . to Uranium Tetrafluoride," Chepter in "Process Chemistry, Vol. II," ’ Prog‘resx)s in Nuclear Energy Serieés, Pergenmon Press (to be published - 231 - and BeF from contaminated molten~-salt reactor fuel by dissolution in | concentrated hydrofluoric acid (>90% HF, balance H 0) The urenivm would first be removed by the volatility process. Thorium, corro:s_ion products, and most of the fission products are insoluble in the sol- vent and would be separated by filtration » centrifugation, or other. solid-liquid separa.tion methods. | The proposed flowsheet is shown in‘_' | Fig. 6.2. Lithilm'_fluoride is quite. soluble in anhydrous HF end quite in- soluble in H20. By itself, BeF is Just the opposite, elthough its solubility in HF is significantly increased by first saturating the HF with LiF. A small emount of water in the: EE‘ also lncreases the 2 elements end some fission products, including the rare earths, are insoluble in both solvents. The solubility of the 63 mole % LiF-37 mole % BeF fuel carrier salt mixture is greater than 100 g/kg in ~90% HF. In an experiment in iwhich mixed fission products (rare earths, strontium, and c.es.iunjl) were added to the salt mixture and the mixture was dissolved and filtered, the salt was decontaminated from rare earths by & factor of ~1000; that is, the activity remaining in the salt was 1/1000 of the activity of the mixture prior to dissolution. No decontami- nation from cesium was obtained. The results for strontium were inecon- clusive, but the decontamination_ appeared to be intermedisate between that from rare earths and that from cesium.- In further studies, fission product and heavy element solubili-~ ties will be investigated, as well as possible hydrolysis of the s’_alt by reaction of the salt with water in the solvent. | It may'be_ found possible to avoid hydrolysis of the salt byusing aphydrous HF or to overcome it by trea.ting the reclaimed. salt with HF. ' 5 h Rare Earth Removal by Exchange with Cerium A study of the removel of high cross-section rare earths from - molten-salt reactor fuels by exchenging them for a low cross-section - 232 BeF solubility markedly. Fluorides end oxides of most of the heavy - . - UNCLASSIFIED ORNL LR DWG. 28749 .........--.-.—.——---UF’6 LiF- BeF, Sait Fluorinator U,Th,FP ~450°C from Reactor Salt HF,H,0 vapors _ Solvent Th,FP] Condenser >90% HF, <10% H20 Dissolver 32°C ~ 0% Salt Sall Solution Solid Solution Fiash in HF-H, 0, Liquid Evaporator + Th,FP Solids Separotor 100-400°C Continuous e e - = Batch - Th,FP Solids Molten - Some Salt & Salt Solvent HF H, UF,, ThF, HF H,O o _ » Fuel ~\éVcste ’r - Make-up a _ ~vapord or. Purification " Th Recovery . - LiF-BeFa-ThFa-UFs : : = to Reoctpr Fig. 6.2. Proposed Salt Reclamation Flowsheet Based on Dissolution in 90+%, HF 233 rare earth, possibly cerium, has also been initiated. The process might be carried out in either of two ways. First, the salt might be saturated with ~1 mole % cerium at an elevated temperature, end then cooled to within ~30°C of the liguidus temperature of the pure selt, at which temperature the rare earth golubility is only 0.2, Thus ~80% of the cerium would precipitate and carry with it ~80% of the fission product rare earths. After a solid-liquid separation, ~this process could be repeated, if desired, to give high-percentage removal of fission product rare earfihs. Second, & perhaps better way to'accomplish the same end wnuld'be to cool the core salt to near its liquidus temperature and éontact it with solid CeFj, prob- ably in a columnar bed of pellets. In principle, the fission pro- duct rare earths could.be‘exchanged'for cérium to ahy desired extent in this menner, depending on the ratio of salt to cerium used, ‘the pellét size (determining the surface area), and the contact time. The attractiveness of the éxchange process is enhanced by the low ebsorption cross section of cerium compared with the sbsorption cross sections of most fission product rare earths. This potential. - advantage is reduced somewhat for e reasonsble processing rate, in that, effectively, several cerium atoms are exchanged for one fission product iare earth atom, since the ceriufi solubility is ~ 0.2 mole % near the liquidus temperature, whereas the fission-product rare earth concentration in the core salt is only ~0.04k% for a processing rate of once per year (and proportionately less for shorter processing perlods, as would be desirable given an economical processing method). In the reference-design moltenésalt reactor the advantage is :educed' further by the fact that for resonance-energy neutrons the cerium cross section is not as much lower than the cross sections of the other rare earths as it is at thermal energies (see Table 6,1). Thus ‘&t 100 ev the poisoning due to 0.2% cerium would be equal to that from | . | fissiOnQProduct'rare earths for & processing rate of once per six | months, iQe., the total would be equivalent to fission-product rare . _earfih roisoning for a once-per-year yrccessing rate by the volatility processe . ' - . | W - 234 - Table 6!1-. Abgorption Crose Sectlons of Various Isotopes at Seversl Energies* Abgsorption Cross Section {barns) Fission Isotope Yield (%) At 0.025 ev At 100 ev’ At 25 kev 1t 6.6 8.4 10.9 0.050 gl 0 6.5 0.63 0.61 0.021 prtil 6.4 11.2 14,5 0547 c:ell"2 6.2 1.0 0.64 0425 mlb 5.9 280.0 12.9 Ndw* 5.1 4.5 7.9 Nalhfi 4.0 52,0 22.8 N’dll‘s 2.1 9.2 7.8 T 2.3 60.0 35.8 su T 0.09 R .Ndll‘l's 1.8 5.2 39 smt*9 1.h 66,000.0 48,6 N+ 0.7 2.8 2.8 smt?t 0.5 10,000.0 5Lk Sm-2 0.3 140.0 26.2 0.860 B o> 0.1 420.0 90.2 e 0.1 5.5 20.0 0,465 adt?”: 0.06 70,000.0 54.6 el 0,05 | 35,7 el 0.02 160,000.0 62.5 'Weigh‘bed average of S | - ebove 2,100.0 12.0 0.5 | ”o.'—é'r | 0.61 - 0.069 Na.tura.‘l. Ce #0,025-ev cross sections from BNL-BSES e.nd its Supplement No. J., yields and 10Q~ev cross sections from P. Greebler, H, Burwitz, end M. L. Stom, Nuclear Sci. and Eng., 2, 334 (1957); 25-kev cross sect:lons from R. L. Macklin, ORNL 3 personal connnuni.ca:bion. - - 235 - e et T R L i e e el kbl et b om it | ‘The cerium exchange processing method would rasise the liq_uidus temperature of the core sa.lt by 30 c or more , end remove plutonium : and trivalent urenium slong with the rare earths. These disadvantages * do not e.ppear to be serious at present, and the potentia.l advantages . 7- of this processing method for thermal (gra.phite-moderated) molten-sa.lt | rea.ctors would seem to justify continued development work. - 5.5. Radioa.ctive Ws.ste Disposal At 8 load factor of 0.80, 600-Mw reactor will produce ~183. kg/y'r of fission products, About 23 wt % of the fission products cen be re- moved as Xe-Kr gases, but the remaining ~1hl kg/yr must be removed by chemical processing The chemical processing waste streams include fused salt 5 NeF pellets ’ 2 N2 s and I-IF--H2 gases, Most of the nongaseous : fission products rema.in in the fuel salt residue a.fter fluorination and. ms.y be stored in this form. Most of the rema.ining fission products are removed by periodically flushing out the micrometallic filter between the fluorinator and the NeF bed and the cold trap between the NaF bed and the F, disposal unit. The NaF bed is replaced infrequently, if and vhen it becomes poisoned with niobium and zirconium. Any remaining fission products in the gas streams are scrubbed out with the F2 and HF, or vented to the reactor off-gas system. For optimum costs, high-power molten-salt reactors should have moderately high inventories of enriched uranium and the fuel should be proces_sed about tiwlce per fuel-inventory burnup, & compromise be- | tween the rate-proportional cost of processing (assumed e,t present to be equsl to the cost of buying new salt) and the savings due to im- 'proved regenera.tion ratio made possible by processing If a fission- | -~ ‘eble ms.terial inventory of 600. kg or more and a fuel ‘processing rate B V'of once per year or greater are assumed and it is considered that the fuel selt will be discarded after its uranium is removed by the voletility process , the wa.ste salt volume amounts to 600 f't5 per | year or more (sbout 1 ft /Mw-year, or 1 £t / kg of fissionsble mater- ia.l processed, or 3 to 4k £t /kg of total fission products) - This - 236 - volume is quite 1ow in comparison with the wastes of other power reac'bors s for example, STR wastes are ~90 f‘b3 /kg of U23 2 processed in the present agueous process and ~3 ft3 /kg of U23 > in the proposed volatility process, and the figures are gbout 4 times larger on & cubic foot per kilogram of fission products basis. | - 237 = . - - \O 03 OV Bl o - — o 11. 13. 1h, 15, 16, 17. 18. 19. 20, 22. 23, 24, 25, 26. . D. S. Billington . F. F. Blankenship E. P. Blizard . A. L. Boch C. J. Borkowski G. E. Boyd E. J. Breeding R. B. Briggs C. E. Center (K-25) R. A. Charpie F, L. Culler L. B. Emlet (X-25) D. E. Ferguson A. P. Fraas J. H. Frye, dr. W. R. Grimes E. Guth C. S. Harrill H. W. Hoffman A. Hollaender A . S, Householder W. H. Jordan G. W. Keilholtz M. T. Kelley J. A. Lane R. S. Livingston ORNL-263% Reactors-Pover TID=-4500 (1h4th ed.) INTERNAL DISTRIBUTION 27‘. 28, 29. 30. 31. 32. 33- 3k, 35, - 36. 37. 38. 39. Lo, Ly, ho, k3. L, k5. ke, h-48. 49-60. 61. 62-63. H. G. MacPherson W. D. Manly J. R. McNally K. Z. Morgan J. P. Murray (Y-12) M. L. Nelson A. M. Perry H. W. Savage A. W. Savolainen H. E. Seagren E. D, Shipley M. J. Skinner A. H. Snell J. A. Swartout E. H. Taylor A. M. Weinberg C. E. Winters Biology Library Health Physics Library Reactor Experimental Engineering Library Central Research Library Laboratory Records Department Laboratory Records, ORNL R.C. ORNL - Y-12 Technical Library, Document Reference Section EX‘I'ERNAL DISTRIBUTION 6)4- Division of Research and Development, AEC, ORO 65-5T1. Given distribution a8’ shown in TID-hSOO (14th ed.) under Reactors-Power " category ..239..