ORNL Central Files Number 57=4=27 (Revised) C=-84 - Reactors-Special Features of Aircraft Reactors % / b Contract No. W-ThO5-eng-26 A PRELIMINARY STUDY OF MOLTEN SALT POWER REACTORS H. G. MacPherson L. G. Alexander D. A. Carrison J. Y. Estabrook B. W. Kinyon L. A, Mann J. T. Roberts F. E. Romie F. C. Vonderlage DATE ISSUED: April 29, 1957 DEC 31957 OAK RIDGE NATIONAL LABORATORY Operated by UNION CARBIIE NUCLEAR COMPANY A Division of Union Carbide and Carbon Corporation Post Office Box X Oak Ridge National Laboratoxry iim Internal Distribution 1-20. ILaboratory Records External Dlstrlbution »-«m a'n Tyt F **‘flfl; ' 21-23, Alr Fbrce nggggsggnflissile Division 2k-25. AFPR, Boeing, Seattle 26. AFPR, Boeing, Wichita 27. AFPR, Curtiss-Wright, Clifton 28. AFPR, Douglas, Long Beach 29-31. AFPR, Douglas, Santa Monics 52. AFPR, Lockheed, Burbank 33-34., AFPR, Lockheed, Marietta 52. AFPR, North American, Canoga Park 36. AFPR, North American Downey 37-38. Air Fbrce Special Weapons Center 39. Air Materiel Command . Air Research and Development Command (RDGN) L1. Air Research and Development Commend (RDTAPS) 42-55. Air Research and Development Command (RDZPSP) 56. Air Technical Intelligence Center o>7=59. ANP Project Office, Convair, Fort Worth 60. Albuquerque Operations Office 61l. Argonne Natioanl Laboratory 62. Armed Forces Special Weapons Project, Sandia 63. Armed Forces Special Weapons Project, Washington 64. Assistant Secretary of the Air Force, R&D 65-70. Atomic Energy Commission, Washington Tl. Atomics International T2. Battelle Memorial Institute T3=T4. Bettis Plant (WAPD) T5. Bureau of Aeronautics T76. Bureau of Aeronautics General Representative . BAR, Aerojet-General, Azusa 78. BAR, Convair, San Diego 9. BAR, Gleann L. Martin, Baltimore 80. BAR, Grummen Aircraft, Bethpage 81. Bureau of Yards and Docks 82. Chicago Operations Qffice 83, Chicago Patent Group 84. Curtiss-Wright Corporation 85. Engineer Research and Development Laboratories 86-89. General Electric Company (ANPD) 90. General Nuclear Engineering Corporation 91. Hartford Area Office 92. Idsho QOperations Office 93. Knolls Atomic Power Laboratory 9k. Lockland Area Office 95. Los Alamos Scientific Laboratory 96. Margquardt Aircraft Company Pimovaiggy-ny g8. 99. 100. 101. 102, 103, 10L, 105. 106. 107. 108, 109. 110. 111-11k, 115, 116. 117. 118, 119, 120. 121, 122, 123-124, 125-1k2, 143-167. Netional Advisory Committee for Aeronautics, Cleveland National Advisory Committee for Aeronautics, Washington Naval Air Develorment Center Naval Air Material Center Naval Air Turbine Test Station Naval Research Laboratory New York Operations Office Nuclear Development Corporation of America Nuclear Metals, Inc. Office of Naval Research Office of the Chief of Naval Operations (OP-361) Patent Branch, Washington Patterson-Moos Pratt & Whitney Aircraft Division San Francisco Operations Office Sandia Corporation School of Aviation Medicine Sylvania-Corning Nuclear Corporation Technical Research Group USAF Headquarters USAF Project RAND U. S. Naval Radiological Defense Laboratory University of California Radiation ILaboratory, Livermore Wright Air Development Center (WCOSI-3) Technical Information Service Extension, Osk Ridge T TABLE OF CONTENTS Page SECTION I - Summary, Recommendations and Acknowledgements eeeceeseses 1 SECTION II - Survey and Analysis of the State of Molten Salt Power Reactor TechNOlOZY esesccecsccsssscsccssscscsscecnsa [ A.. Ma'terial.s UL L B AL B BN B BN BN BN BN B BN NN N BN BN BN BN BN R BE BN BN B BN BN RN BN BN BN BN BN RN NN NN OB B B N NN R NN NN BN RN NN BN 7 Fuel Carrier Evaluation cesesecescoccccsccccoscancnnans [ Blanket Material Evaluation .eeceseccccsccscevesscccaee 13 Intermediate COOLANtS cveceseccccccrsscsscsssssescecess Ll Container Materials seeceevessssoscascassccssssacosnocese L1 MOQerator MEterials ceececceccecccsosrersssssscscsecoces 2l \n-l-"ylml—' B. Materia]—s S 5 560000000500 H S 0C S80S PO PO OISO 0SS SO SONE BSOS PCPOS SO 25 e« PUlDS ceeerceconscsaccsssssacsscssnssssosssssccscscescense 26 « Healt EXChBNgers .ceccecsscsscsssssssssssscssccscesssscasa 29 +« Reactor Vessels sceeesscevessscesssscssssscssssscscscoe 39 e Other VesSSelsS cescscessscsssssssccssssssssssssccaccssns 30 o JOInts and VaBlVeS ecceceososcessssssssscssssscnsscsnscsee 39 . Instrument and Control COMPONENtS ceecceescescoccsccsess HO O\ AN O C. Component SYSLEmMS eeeeecsccccosscccsssscsssccscsssssssscccnes 45 1. Salt and Liquid Metal Charging and Storage Systems .... 45 2, Off=0as HANATINg seeececcccccncccacosscosesssssonsscsce UO 3. Inert Gas SYStEM ceccecccccocrccnsoscoscaccccssocsnsoes U4S 4. Heating and Cooling of COMPONENLES +eesceccocconoossensa US D. NuClear Considerations @ * 000 OO 0O O et PO IE DSBS OO0 OES RN SSasS 50 Previous Work and Early Consideration sececececcccecscss 50 One Region ReACtOTS sececccescococssssssscccscsscsccsaas 5 « TwO Reglon ReacClOI'S ceeecesscscccccssnsscsssscsssscssses 59 . Reactivity Effects in Typical ReQCLOT .cveecccsscecccess 69 W E. Reactor Operation, Control and Safety ecceseccecssccccccaass 70 l. The Control Problem of Nuclear Power Reactors seceseeee 1 2. F‘lleling S & & 0 0O 0O 0PV B OB O PO e OO OO PAEARDE OO R O PRSP ESIPSOES 76 3. Criticality Staz't-up o2 0 60 0 000 00 O "B OO O S BSOS PO SE NSO eI e 77 F. Build-up of Nuclear Poisons and Chemical Processing s.ccce.. 80 1. Fission Product POiSORING ececcecescoccscncssscsssnsecse S0 2. Pa255’ NP237 and NP239 POisoning * 0 &9 0SB E SO S SO0 SEeD RS 82 3, Corrosion Product POLiSONING sececcacsscsccsascscccscccs 83 R UMENTED m* k., Chemical Processing and Fuel Reconstitution .cecosecceeo 83 5. Build-U.P Of E'Ven-MaSS"NUInber Uranim ISOtOPes eeceeces o 88 6. Rwioa’ctive Waste Disposal 24 8 8 8 8 0 &9 0 00 "B OSSOSO S ODODOOSBSND 91 G. F‘llel CycleEcOnomics ® 0 8 & 88 85 0 08 900 S PSS OSSO SO S OO0 S SE S S S S DOOOCO 95 l e Cost mse s ® 0 0 O 8 0 00 0 02O SO 80O OO S OB OO eSO AE SN S SSSBSEESEOOO0DOODO0OO0OS 95 2. "Steady State" Neutron Balances and Comparative Fuel costs O 0O PR OO O F PP OE PP EP S OO NP ® OO0 MO8 0D 914' SECTION III - Reference mSi@ ReaCtor @0 600 00 0QC0OOSHDAISOIIODPDOROODODE O 98 A. IntrOduCtion 0SSP O8O 00N P EE OO 0P OSSOSO SO SDONO0EDGOS NS00SO RS 98 B, Heat Generation, Transfer, and Conversion Syste® .ccecsccoe 103 l. Reactor * 0 OO0 00RO ESECE OSSN OSSO0 000GS OO0 800000 105 2 @ Heat mchmlgers ® O 5 8 00 5 88508 PSS BSOSO NSO SDRPOSOS O OO0OO0OCSS OSSO loh 5 * stem cycle ® D O 8 8 8 80 8 8 8 50 800800 S PRS0 000 S 9S8O0 O0Aa S8 00008 DH lll C. Components and Component SysStell scceecscessesssscosscesscocea L1l3 1. PUNDPS eecoscrssvsssscssssssassansoscssosssssncsssssoscses 113 2o VALVES vesecessesscsssscessossssscsesssossssassnccscnace LoD 3, Pipes and TUDES eeeeecesessossscossosconsoasaocsososos LLU h, PFill-and-Drain Tanks .eesecccoescsccsocoasscssosaonooo LLI 5. Gas Supply Systems (Helium, Nitrogen, and Compressed . O I 6. OFf=CaS SYSTLEM eeesercocccoooncsancnoossassocoscoosose LLT 7. Preheating and Temperature Maintenance .ccoecococecooscos 1LY D. Pl‘allt Ial.yout OO0 6 8 00 000 ¢ PO P PD SO DT OO SO SS90 9P S BsCOEROCO0 0000000 118 E. Chemical Processing and Fuel Cycle EcCOROMICS soeees00000ess 122 1. Core ProceSSiNg ececsssossecscascsssosssscscsnsconcescns Lol 2. Blanket Processing cceevecsscsccssosncenssccooecconces 122 3, Chemical Processing COStS ccecesccscscsscosssocssccsoo L2k k, TFuel Cycle ECONOMICS veeeescecesccascscsncoscocsacosse 125 F. COSt AnahlySiS ® 6 0 5 60 600 0 0" 0P PSSO S SO S0 SN CO0E0OCC0CSeE SO R 0000 6N 126 TNET0AUCHLION eeesevoesacoscecoessacssvsssssanssoncoosana 12O Materials and Components Development COStS ccoecoccsos 127 Design and Construction COStS seccescnesccscseccessccs 129 Cost of Power from the Reference Desigh Reactor ccsceo 130 Fobr o Appendix 8 & 00 ¢ 90 ® 00 S 0O D OO OO O OO B BN OO PP RFOO S SEDeEP0R00000eeB000O0D0O0CSEe®BD 155 o Fig. No. wVie LIST OF FIGURES Title Fission Cross Sections and Eta for U233 and U235 in the Ocusol=A Program. Reference Design Reactor Heat Transfer Clrcuit Showing Simulator Constants. Change in Coolant Inlet Temperature for Intermediate Heat Exchanger Due to Fuel Burn-up, for a Typical Fused Salt Circulating Fuel Reactor at a Power Density of 200 watts/cm?. Pused Salt-Fluoride Volatility Uranium Recovery Process. UFg Reduction Process Flow Sheet. Schematic Diagram of Heat Transfer System. Reference Design Reactor. Temperature-Heat Diagrams for Heat Exchangers. Steam Cycle Diagram. Plan View of Power Plant. Section Through Reactor and Power Plant. Page 56 5 86 89 101 105 108 120 121 -1 - A PRELIMINARY STUDY OF MOLTEN SALT POWER REACTORS SECTION T Summary, Recommendations and Acknowledgments Molten salts provide the basis of a new family of liquid fue1 power reactors. The wide range of solubility of uranium, thorium and plutonium com- pounds makes the system flexible, and allows the consideration of a variety of reactors. Suitable salt mixtures have meltinglpoints in the 850-950°F range and will probably prove to be sufficiently compatible with known alloys to pro- vide long-lived components, if the temperature is kept below 13000F° Thus the salt systems naturally tend to operate in a temperature region suitable for mocern steam plants and achleve these temperatures in unpressurized systems. The molten salt reactor system, for purposes other than electric power generation, is not new. Intensive research and development over the past seven Years under ANP sponsorship has provided reasonable answers to a majority of the obvious difficulties. One of the most important of these is the ability to handle liquids at high temperatures and to maintain them above their melting points. A great deal of information on the chemical and physical properties of a wide variety of molten salts has been obtained, and methods- are in operation for their manufac- ture, purification and handling. It has been fOund that the simple ionic salts are stable under radiation, and suffer no deterioration other than the build-up of fission products. The molten salt system has the usual benefits attributed to fluid fuel systems. The principal advantages claimed over solid fuel elements are: (1) the lack of radiation damage that can limit fuel burn-up; (2) the avoidance of the -2 - expense of fabricating new fuel elements; (3) the possibility (partially demon- strated in the ARE) of continuous gaseous fission product removal; (4) a high negative temperature coefficient of reactivity; and (5) the ability to add make- up fuel as needed; so that provision of excess reactivity is unnecessary. The latter two factors make possible & reactor without control rods, which automati- cally adjusts its power in response to changes of the electrical load. The lack of excess reactivity can lead to a reactor that is safe from nuclear power excursions. In comparison with the aqueous systems, the molten salt system has three outstanding advantages: it allows high temperature with low pressure; explosive radioclytic gases are not formed; and it provides soluble thorium and plutonium compounds. The compensating disadvantages, high melting point and basically poorer neutron economy, are difficult to assess without further work. Probably the most outstanding characteristics of the molten salt systems is their chemical flexibility, i.e., the wide variety of molten salt solutions which are of interest for reactor use. In this respect, the molten salt systems are prac- tically unique; this is the essential advantage which they enjoy over the U-Bi systems. Thus the molten salt systems are not to be thought of in terms of a single reactor - rather, they are the basis for a new class of reactors. Included in this class are all of the embodiments which comprise the whole of solid fuel element technology: straight U235 or Pu burner, Th-U or Pu-U thermal converter or breeder, Th-U or Pu-U fast converters or breedérsq Of possible short-term interest is the U255 or Pu stralght burner: Dbecause of the inherently high temperatures and because there are no fuel elements, the fuel cost in the salt system can be of the order of 2 mills/kwh. Moreover, the molten salt system 1s, except for the molten Pu alloy system, probably the only system which will allow plutonium to be burned at high temperature in liquid form. Ll -3 - The state of present technology suggests that homogeneous converters using a base salt composed of BeF, and either Ii7F or NaF, and using UFh for 2 fuel and ThFh for a fertile material, are more suitable for early reactors than are graphite moderated reactors or Pu fueled reactors. The conversion ratioc in such an early system might reach 0.6. The chief virtues of this class of molten salt reactor are that it is based on well explored principles and that the use of g simple fuel cycle should lead to low fuel c¢cycle costs. With further development, the same base salt (using Li7F) can be com- bined with a graphite moderator in a heterogeneous arrangement to provide a self-contained thorium-U233 system with a breeding ratio of about one. The ¢hief advantage of the molten salt system over other liquid systems in pursuing this objective is, as has been mentioned, that it is the only system in which a soluble thorium compound can be used, and thus the problem of slurry handling is avoided. Plutonium is an alternate fuel in the fluoride salt system. Only moderate breeding ratics are expected in thermal or epithermal reactors; but a small, highly concentrated flucride reactor may be fast enough to provide a breeding ratio of one. Eventuzally the use of chloride salts might provide a fast plutonium reactor with & breeding gain, although this would require use of separated 0137° The plutonium system needs additional research to determine the stability of Pu compounds and to provide a suitable chemical processing system. The present report is primarily intended to summarize the state of the molten salt art as applied to c¢ivilian power. The report is divided into three parts. Section I is the summary and recommendations. Section II is a survey of the state of molten salt technology. Section IITI is an analysis of one possible T molten salt reactor embodiment - a two reglon converter based on the fuel 69 Ii F - 30 BeF2 - 1 UFh and the blanket composition Tk LiTF - 26 ThFh. ~This embodiment, -4 . called the reference design, has been examined carefully, primarily to bring to focus the problems which may arise if a full-scale molten salt system were to be built soon. The conclusion which we can draw from our study of the molten salt situation is that a large-scale molten salt reactor - either a straight U burner or a non-breeding converter - could be built, but that prior to building, two important questions should be answered: 1. Will any molten salt reactor produce economical power? Our study shows the answer is probably yes, provided longevity of components can be assured. Hence the issue depends on the second question: From what we know about materials compatibility, how likely are we to develop a salt and a container metal which will last for many years of operation? This is the central issue in the civilian molten salt nuclear power reactor program. The information gathered by the ANP project, added to our general knowledge of the mechanism of attack on metals (particularly INOR-8) by fluorides, suggests that the outlook for a solution tc this problem is very good. However, very little long- term testing at power reactor temperatures (e~ lEOOoF) has been done; our recommendations, therefore, center around thé necessity for acquir- ing this long-term data as soon as possible. Should these tests demon- strate;the long-term compatibility of-materials, there will still be required the development of reliable large-scale components. Experience on ANP indicates that this part of the development should not present major difficulties. Recommendations In view of the preceding, we recommend: 1. The long-term corrosion resistance of the proposed alloys in the salts that could be used in a power reactor should be established. This will involve the operation of a number of pumped loops incorporating a temperature gradient, to be operated at the temperature of interest for periods of at least a year. 2. The effects of radiation and fission product build-up on the com- patibility of the salt and alloy should be thoroughly investigated. At least two in-pile pumped loops simulating the condition in a molten salt reactor should be operated for a long period of time. These in-pile loop tests should bhe supple- mented by small-scale studies of the behavior of fission products. 3. It is recommended that a modest reactor study effort be maintained. Different embodiments of molten salt reactors would be examined, so that if favor- able results from Items (1) and (2) are obtained, it will be possible to recommend a specific reactor, probably a burner or converter, for design and construction. Also, the problem of remote maintenance, which 1s shared by all circulating fuel reactors, could be examined in further detail. Y}, Since there is always a time lag between the initiation of research and the availability of practical developments, it is recommended that a modest research program aimed at longer term possibilities be maintained. Objectives would be (a) the incorporation of a solid moderator, (b) utilization of plutonium in the molten salt system, (c) better alloys, and (d) improved fission product removal systems, Acknowledgements Many members of ORNL and other organizetions have helped in the work of the group and have shown great interest in its progress. It is difficult to single out individuals for mention, but the following people have been serving on a project steering committee: A. M. Weinberg J. A. Swartout R. A. Charpie S. J. Cromer W. K. Ergen W. Re Grimes W. H. Jordan W. D. Manly Others who have heen especially close to the project are: E. S. Bettis E. A. Franco-Ferreira J. L. Gregg F. Kertesz W. B. McDonald E. R. Mann P. Patriarca M. T. Robinson H. W. Savage -7 - SECTION II Survey And Analysis Of The State Of Molten Salt Power Reactor Technology In the research and development program cerried out by the ANP for the construction of the ARE and future high performasnce reactors, much technical information hfis been derived which is applicable to power reactors. The purpose of this section 1s to abstract the information that is most pertinent to the con- struction of power reactors, and to provide adequate references to document properly the summaries given. This has been supplemented by studies of nuclear characteristies of homogeneous one and two region power reactors. It will be seen that most of the information required to design a practical power reactor is available. However, long-term tests of materials and components are lacking, and they must be supplied by & power reactor research and development program. A. Msterials l. Fuel Carrier Evaluation The applicability of molten salts to nuclear reactors has been ably reviewed by W. R. Grimes and others Y, g/, by Crooks et al 2/, and Schuman E/o 1/ Grimes, W. R., et al, "Molten Salt Solutions"”, Proceedings of the Second Fluid Fuels Development Conference, ORNL-CF-52-4-197 (1952), p. 320 et seq., Secret _ , Grimes, W. R., et al, "Fused Salt Systems", The Reactor Handbook, Vol. II, Engineering, RH-2 (1955) p. 799 et seq., Unclassified 3/ Crooks, R. C., et al, "Fused Salt Mixtures as Potential Liquid Fuels for Nuclear Power Reactors", BMI-864 (1953), Secret L Schuman, R. P., "A Discussion of Possible Homogeneous Reactor Fuels", KAPL-63k4 (1951) & - The most promising systems are those comprising the fluorides and chlorides of the alkali metals, zirconium, and beryllium. These appear to possess the most desirable combination of low neutron absorption, high solvent power, and chemical inertness. In general, the chlorides have lower melting points, but appear to be less stable and more corrosive than the fluorides. The use of chlorides in & homogeneous fast reactor would be preferable except that the strong (n,p) reaction exhibited by 0135 would necessitate the separation of the chlorine isotopes. The fluoride systems appear to be preferable for use in thermal and epithermal reactors. Many mixtureé have been investigated, mainly at ORNL and at Mound Iasboratory. The physical properties of these mixtures, in so far as they are known, have been tabulated by Cohen et al 2/. Phase studies are exten- sively reported é/o 117 has an attractively low capture cross section (0.0189 varns at 0.0759 ev); but Ii6, which eomprises'To5 percent of the natural mixture, has a capture cross section of 542 barns at this energy. The cross sections for several compositions are shown in Table I; also shown are the thermal cross sections of Na, K, Rb, and Cs. Table I CAPTURE CROSS SECTIONS OF AIKALI METAIS AT 0.0759 ev (llBOOF) Element Cross Section, barns Lithium 6 0.1 % Ii 0.561 0.01 " 0.0731 0.001 " 0.0243% 0.0001 " 0,019k Sodium 0.290 Potassium 1.13%0 Rubidium 0.401 Cesium 29 &2 Coben, S. I., et al, "A Physical Property Summary for ANP Fluoride Mixtures", ORNL-2150 (1956), Secret C-84 §/ The Atomic Energy Commission, The Reactor Handbook, Vol 2, Engineering, RH-2 (1955), Secret, and ANP Quarterly Reports -9 - The capture cross sections at higher energies presumably stand in 7 has approximately the same relation as at thermal. It is seen that purified Ii an attractively low cross section in comparison to the other alkali metals, and that sodium is the next bept alkali metal. The fluorides of Ii, Ne, K, and Rb melt at 1550, 1820, 1560, and 1460°F, respectively Z/° Binary mixtures of these salts with UFH form eutectics having melting points and compositions shown in Table II. Table II BINARY EUTECTICS OF UFh AND AIKAILI FIUORIDES Alkali Fluoride Mole % UFL in Eutectic Melting Point, °p IiF 26 915 NaF 26 11%0 KF 14 1345 RbF 10 1330 With the possible exception of the first, these combinations are too high melting to be attractive as fuels; however, the eutectics of UFh with IiF and NaF might be suitable for use in the blanket of a two region plutonium breeder- converter. DBinary mixtures containing less than 1.0 mole percent UFh do not exhibit liquidus temperatures below 1450°F. IAF and NaF form an eutectic melting at 1204°F 8/. Small adaitions of UFh raise the liquidus temperature slightly. The ternary eutectic melts somewhat below 8hO°F and contains approximately 30 mole percent UFh' This system is attrac- tive only as a blanket material. 7/ The Atomic Energy Commission, The Reactor Handbook, Vol. 2, Engineering, RHE-2 (1955), Secret 8/ TIbid., p. 948 ~ 10 - The Na-Zr fluoride system has been extensively studied at ORNL and a phase diagram published 2/, An eutectic containing about 42 mole percent Zth melts at 9lOOF° Small additions of UFh lower the melting point appreciably. A fuel of this type was successfully used in the Aircraft Reactor Experiment. Inconel is reasonably resistant to corrosion by this system at 1500°F° Although long~term data are lacking, there is theoretical reason to expect the corrosion rate at 1200°F to be sufficiently low that Inconel equipment would last several years. However; in relation to its use in a central station power reactor, the Na-Zr fluoride system has several serious disadvantages. The Na capture cross section is less favorable than the 117 cross section. More important, recent data 19/ indicate that the capture cross section of Zr is intolerably high in the epithermal and intermediate neutron energy ranges. In addition, there is the sc-called "snow" problem, iaené ZrF) tends to evaporate from the fuel and crystallize on surfaces exposed to the vapor. In comparison to the Ii-Be system discussed below, the Na-Zr system has inferior heat transfer and cooling effectiveness. Finally, the expectation at Oak Ridge is that the INOR-8 alloys will prove to be as resistant to the Be salts as to the Zr salts, and that there is, therefore, no compelling reason for selecting the Na-Zr system. The capture cross section of beryllium appears to be satisfactorily low at all enefgies° A new phase diagram for the system IiF-BeF. has recently 2 been published ll/. A mixture containing 31 mole percent BeF,, (Mixture Th) 9/ The Atomic Energy Commission, The Reactor Handbook, Vol. 2, Engineering, RH-2 (1955), p. 952, Secret Macklin, R. L., Private Communication, ORNL (1957) &k Eichelberger, J. F. and Jones, L. V., "Iiquid Cycle Reactors, Fused Salts Research Project - Report”, ML-CF-57-1-10 (1957), p. 3, Secret (Supersedes Figure 6.2.26, p. 950 of The Reactor Handbook) - 11 - reportedly liquifies at 968°F; however, Cohen et al lg/give 941°F as the 1iquidus temperature of Mixture Th. Other physical properties are listed by Cohen, who gives 7.5 cp for the viscosity at 11120F. Further additions of BeF2 increase the viscosity lé/n A new ternary diagram for the system LiF-BeFE-UFh has recently been published l&/. Additions of UFh to the compound LieBth (1iquidus tempera- ture BTBOF) lower the liquidus temperafure appreciably. A mixture melting somewhat below 840°F (possibly as low as 820°F) can be obtained, having about 5 mole percent UFho The ternary eutectic melts at 805°F and contains about 8 percent UFh’ 22 per- cent BeFé, and 7O percent LiF. The system LiF-BeFé is attractive as a fuel carrier. Substantial concentrations of ThFu in the core fluid may be obtained by blending Mixture T4 with 3 IiF - ThFh, and a liquidus temperature diagram for the ternary system has been determined li/_ The liquidus temperfiture along the join between Mixture T4 and 3 IiF ° ThFh appears to lie below 950°F for mixtures con- taining up to 10 mole percent ThFh. The liquidus temperature thereafter rises slowly at first, and then more rapidly. ©Small additions of UFh to any of these mixtures should lower the liquidus temperature somewhat. No data on the system NaF-BeFE-ThFh are available; however, tpe solubility of ThFh and other physical properties are expected to be nearly as good as for the Li-Be system. 12/ Cohen, S. I., et al, "A Physical Property Summary for ANP Fluoride Mixtures", ORNL 2150 (1956), Secret C-8k4 Barton, C. J., Private Communication, ORNL (1957) N Eichelberger, J. F. and Jones, L. V., "Liquid Cycle Reactors, Fused Salts Reactor Project - Report", ML-CF-57-1-10 (1957), p. 6, Secret (Supersedes Figure 6.2.2, p. 930, Vol. 2 of The Reactor Handbook) & The Atomic Energy Commission, The Reactor Handbook, Vol. 2, Engineering, RH-2 (1955), p. 962, Secret - 12 - Mixture T4 has moderating power substantially less than beryllium or carbon; gzt stands in the relation 0.176, 0.064, and 0.037 for beryllium, graphite | and Mixture Th, respectively. Nuclear calculations on these systems were performed by means of the Univac program Ocusol lé/, The ages from fission to various energies for Mix- ture 74 were computed and listed in Table ITI, together with the corresponding capture-escape probabilities. Table III NUCIEAR PROPERTIES OF MIXTURE Th-A (69% IiF,* 31% BeF,) o 2 Fission Neutrons Ener ev e, cm Capture-Escape Probability 1234 207 0.973 112 298 0,971 10,16 396 0.96k 0,0759 591 0,848 * 11 isotopic composition: 99.99% 117 Cohen et al 17/ give 1.3 x IO'u/oF for the mean volumetric coefficient of thermal expansion for Mixture T4 in the liquid state, presumably in the range from 1100 to lSOOOF. This may be compared to the coefficient of Mixture %0 (50 NaF, 46 ZrF) , L UFh), which is 1.58 x 10‘”/°F. The heat capacity of the liquid is given as 0.67 Btu/1b-CF and the density as 120 1b/ft° at 1150°F. 16/ Alexander, L. G., Carrison, D. A. and Roberts, J. T., "An Operating Manual for the Univac Program Ocusol-A, A Modification of Eyewash", ORNL-CF- (in preparation) 17/ Cohen, S. I., et al, "A Physical Property Summary for ANP Fluoride Mixtures", ORNL-2150 (1956), Secret C-8k - 13 - The stability of alkali fluorides and zirconium fluoride toward heat and radiation seems to be well established by the work at Oak Ridge. Beryllium fluoride is thermally stable at temperatures of interest; preliminary in-pile tests lé/ indicate that BeF2 is as stable toward radiation, including fission fragments, as Ztho The compatibility of the systems under consideration with container materials and adjacent fluids is dealt with in later sections, as is also the problem of processing irradiated fuels. Costs are listed in Section II-F. On the basis of presently available information, the fuel carrier salts which have been considered appear to stand in the following order of pre- ference: IiF-BeF,; NaF-BeF.; LiF-NaF-BeF,. The LiF-BeF. system has slightly 2 2 2 7 better moderating power, lower parasitic absorption (if high purity Ii' can be obtained), and adequate solubility for ThF) and UF). It may prove to be more corrcsive than the NaF-BeFé system, and the cost is greater. 2. Blanket Material Evaluation The Ii-Be-Th fluoride mixtures recommended above as fuel carrier appear to be suitable for use in the blanket of a two region reactor. There is evidenceig/ that these mixtures when containing no UFh are much less corrosive than fuel bear- ing mixtures. As mentioned above, a mixture containing 10 mole percent ThFh has (according to the diagram on p. 962, Vol. 2 of The Reactor Handbook) a liquidus tempersture of 9520F° If a safety margin of 100°F 1s specified, the minimum blanket inlet temperature would be 1032°F. 18/ Keilholtz, G. W., et al, "Solid State Division Quarterly Progress Report Ending May 10, 1952", ORNL-1301 (1952), Secret 19/ Blakely, J. P., "Corrosion Results of Be Salts in Thermal Convection Loops”, Memo of April 6, 1956, to C. J. Barton, ORNL - 14 - It might be possible to dispense with the BeFé and use a mixture of IiF and ThFho A phase diagram for this system is given 29/, The compound 53 LiF - ThFL melts at 107OOF, and may possibly be a satisfactory blanket fluid. The density was estimated by the method of Cohen gl/ to be 4.55 g/ce at 11120F, and the melt conteins about 2700 grams of thorium per liter of solution. The viscosity has not been reported, but is not expected to be greater than 7 cp at 1100°F., The corrosion rate in Inconel is low gg/. Additions of NaF to this com- pound should lower the liquidus temperature appreciably, perhaps as much as IOOOF. 3. Intermediate Coolants From the standpoint of simplicity, it would be desirable to transfer the reactor heat directly from the circulating fuel to the steam. This, however, has several serious disadvantages, among them being the induction of radiocactivity in the steam by delayed neutrons; the danger of contamination of the power-producing equipment by leakage of fuel into the power loop, and the danger of nuclear or other accidents in case of leakage of water into the core system. It therefore seems desirable to employ intermediate coolants. Among the intermediate coolants considered were water, organic liquids, liquid metals, and molten salts. High pressure, and nuclear and chemical compati - bility with fuel eliminate water. The organic liquids have poor thermal stability above '1100°F. Among liquid metals, sodium (or NaK), mercury, lead, and bismuth 20/ Cuneo, D. R., "ANP Chemistry Section Progress Report for October 9-22, 1957", ORNL-CF-56-10-121 (1956), Secret (Supersedes Fig. 6.2.31, p. 958 of The Reactor Handbook) 21/ Cohen, S. I. and Jones, T. N., "A Summary of Density Measurements on Molten Fluoride Mixtures and a Correlation Useful for Predicting Densities of Fluoride Mixtures"”, ORNL-1702 (1954), Secret 22/ Doss, F. A., "Supplement to WR Salt Mixtures in Thermal Convection Loops", Memo of October 5, 1956, to W. R. Grimes - 15 - were considered. JIead and bismuth appear to be excessively corrosive (mass transfer effects). Mercury has poor heat transfer characteristices and has special problems of containment. | Sodium has relatively good heat transfer characteristics, can be readily pumped, but is chemically incompatible with both UFh bearing salts and water. The reaction of sodium with a Zr based fuel in a pump loop with a simulated leak was investigated by L. A. Mann gé/’ 22/. It appears that slow addition of sodium to the system IiF-BeFé—UFh would result first in the reduc- tion of the UFh to UF This would probably not result in the formation of a 3° precipitate at ¢oncentrations of UFh under consideration. The UF3 and ThFh would be reduced next, and then the BeFe. Solid phases containing uranium metal would probably be formed shortly after the reduction of the thorium begins. Molten salts considered for intermediate coolants include Mixtures T4 (three variations), 12 and 84. A study of these, together with metallic sodium, was performed by means of a simplified systems analysis. The results, together with relevant physical properties, are listed in Table IV. It is seen that Mix- ture TU-A, which is the base recommended for the fuel mixture, has a melting point too high for safety, being only 3hoF less than the proposed intermediate coolant inlet temperature (IOOOOF). Mixture T4-C has a satisfactorily low melting point, but the viscosity (14.0 cp) seems excessive. Mixture T4-B appears to be suitable from standpoint of both melting point and viscosity. It would also be completely compatible with a ffiel based on Mixture T4. Ieakage of Mix- ture TL-B into the fuel circuit would not result in the formation of precipitates, 23/ Mann, L. A., Private Communication, ORNL (1957) 2/ Grimes, W. R. and Mann, L. A., "Reactions of Fluoride Mixtures with Reducing Agents", ORNI-1439 (1952), p. 118 - 16 - would not contaminate the fuel provided the lithium were of the same isotopic » composition as that in the fuel, and could only decrease the reactivity by dilu- tion of the fuel. It should be possible to tolerate small, continuous leaks in normal operation. On the other hand, salts containing beryllium are incompatible with sodium metal, which displaces beryllium from the fluoride compound. The reaction is expected to be energetic and rapid, but not explosive, since no gases are formed 22/. It was estimated from the heat of formation data given by Quill et al Eé/ that the addition of one mole of sodium to BeFé would release about 30 Kecal of heat at 1000°F. This is sufficient to raise the temperature of the stoichio- metric mixture about 13000F above the initial temperature. In addition, the beryllium metal formed would deposit throughout the system and might lead to embrittlement of the material of construction. The consequences of the leak- age of sodium metal into IiF-BeFé thus could be serious. . Mixture 12 (a Flinak) appears to be completely inert toward sodium. From a heat transfer standpoint, it appears to have a slight advantage over Mix- ture T4-B, as shown in Table IV, where the required heat transfer areas are compared. Mixture 12 may be slightly more corrosive than Mixture T4 22/, but the corrosion in the intermediate coolant loop is not expected to be critical because of the lower temperatures prevailing there. It has fairly good compatibility with a fuel based on Mixture T4 (LiF-BeFe),_ Small leaks of Mixture 12 into the core system probably would not result in the formation of precipitates. The potassium would poison the nuclear reaction, as would also any Ii6 present. Iarger lesks might lead to the 25/ Grimes, W. R., Private Commumnication (1957) 26/ Quill, L. L. (editor), "Chemistry and Metallurgy of Miscellaneous Materials- Thermodynamics”, National Nuclear Energy Series IV-19B, MeGraw-Hill Book Co., New York, N. Y. (1950) W _ - 17 - precipitation of binary compounds of XF with UFh and ThFh. Precipitation of UFh outside the core would tend to decrease the reactivity in the core; the precipitation of ThFh would have the opposite effect. The precise course of events cannot at present be predicted, but it seems doubtful that the reactivity increase due to the precipitation of ThFh could override the decreases due to the precipitation of UFh and the addition of X and 1160 On the basis of these considerations, Mixture 12 would appear the safest choice of intermediate coolant. By comparing the results listed for metallic sodium in Teble IV with those for the salts; the penalty imposed by the use of salts as intermediate coolants can be assessed. The heat transfer areas and fuel holdup volumes are significantly less with sodium. The power consumption for pumping fuel and sodium is excessive for the case where the tube pitch is the minimum allowable. Doubling the tube pitch reduces the pumping power to a negligibly low level without increasing the heat transfer area or fuel volume excessively. 4., Container Materials The feasibility of the molten salt reactor system depends in large measure c¢n the existence of a suitable container material. Any corrosion of the container metal must be slow enough so that components will be long-lived. The container material must be obtainable in sufficient quality and quantities from commercial vendors, must be fabricable into suitable shapes, and must have satisfactory strength, creep characteristics; and other physical properties at the operating temperatures to be encountered. Several hundred high temperature static and dynamic tests have been carried out since 1950 to determine which materials were most practical for con- tainment of the fluoride salts EZ/. Of the pure metals, molybdenum; columbium 27 ORNL-1491 - 18 - Table IV COMPARISON OF INTERMEDIATE COOLANTS Fuel Inlet - 1100°F Coolant Inlet - 1000°F 0.378" x 0.039" Tubes on minimum allowable pitech (0.495 in.) Basgis: Mixture Number Composition, mole % IiF NaF KF BeF2 Melting Point, °F Density; p o 1b/ft5 at 110073 Viscosity, p cp at 11000F Heat Capacity, Cp Btu/1b-OF at 1292°F Thermal Conductivity, K, Btu/hr-ft-°F Molecular Weight, M gm/mole Relative Heat Transfer Surface Relative Pumping Power, g * Fuel Holdup Volume, ft5 Compatibility with Fuel Compatibility with Na Induced Radiocactivity * Percent of heat transferred Th-A 69.0 31.0 966 120 T3 0.67 4.2 2.4 Good Poor Low T4-B 62.7 37.3 842 120 9.0 (0.67) (4.2) 33.7 1.09 0.068 125 Good Poor Low Th-C 56.8 43.2 797 124 14.0 Good Poor Low % (1) Flow area on shell side same as for salts Fuel Outlet - 1200°F 12 46.5 11.5 42.0 849 131 5.0 0.45 2.6 41,2 1.00 0.076 11k Fair Good Moderate (2) Flow area on shell side twice as great as for salts 8L 35.0 27.0 38.0 640 125 8.1 3.2 38.3 1.15 0.076 151 Poor Poor Moderate Coolant Outlet - 1125°F Na 0.30 57 High - 19 - and nickel were outstanding in corrosion resistance, but were eliminated for reasons of fabrication difficulties and/or physical property shortcomings. From the numerous alloys tested, Inconel was selected for extensive additional testing and study in both thermal convection and pumped loops containing large temperature gradients. Iater work has indicated the greater desirability of a nickel-molybdenum alloy, INOR-8, and ORNL at present is active in trying to bring it into status as a commercial alloy. Both Inconel and INOR-8 alloys contain chromium. Their mechanism of corrosion §§/ in NaF-Zth salts has been determined to be the diffusion of chromium to the hottest metal surface, its solution in the salt there, and the diffusion of chromium into the colder metal surfaces. In this way, subsurface voids ar: created in the hottest region, from which the chromium is removed. The usual mechanism of the corrosion is the same for both Inconel and INOR-8 alloy, and it is therefore presumed that the general relstionships of corrosion rate to time, temperature, and other test conditions found by analysis of hundreds of thermal convection and pumped loop tests for Inconel in the NaF-Zth salt will also hold for INOR-8. With cther salts there is the possibility of the deposition of free Cr metal in the cold region, thus changing the rate limiting mechanism. In tests of Inconel against Fuel-30 (50% NaF, 46% ZrF) 4, UFh)’ . approximately the same corrosion rates were found for both thermal convection loops and pumped loops in which a 200°F to 3000F temperature difference was maintained between the hot and cold sections. The depth of corrosion was found to depend primarily on the metal wall temperatures 22/. Most tests were carried 28/ ORNL-2106, Parts 1-5, p. 96 29/ ORNL-2217 - 20 - out at maximum metal wall temperatures of 1600--170001"‘° 1000-hour pumped loop tests at 16OOOF wall temperature typically glve corrosion depths of 5 to 7 mils, and this increases to 9 to 10 mils at l'TOOOFo Thermal convection loops with maximum wall temperatures estimated to be as low as 15500F give 2000 hours corrosion of 7 mils, as compared to corrosion of 13 mils at 1640°F. Thus the existence of a temperature coefficient is known, but its magnitude is uncertain for extrapolating down to wall temperatures of lEOOOF. With & maximum wall temperature of about 1600°F, the corrosion in the usual pumped loops is about 3 mils in the first 15 hours, and 3 to 4 mils per 1000 hours after that. The longest tests were run for 3000 hours (1li4-mil attack) and 8300 hours or sbout one year (25-mil attack). The time dependence is consis- tent with the theory that in the first 15 hours, the salt becomes saturated with chromium, and that after this initial period, the rate limiting step is that of diffusion of the chromium into the walls at the cold temperature. Attempts to show an increase of corrosion rate with an increase in area of the cold wall section were inconclusive, however. Early thermal convection loops 5—0-/ operated at 15OOOF showed a depend- ence of corrosion on the difference in temperature between hot and cold legs, with corrosion being decreased by a factor of 2 by lowering the temperature difference to 15OOF. Tests with pumped loops él/, while not conclusive, suggest that tempera- tu:e drops of greater than 200°F noticeably increase the attack, but that decreasing the temperature drop from 200°F to 100°F does not correspondingly reduce the attack. 30/ ORNL-1729 31/ ORNL-2221 - 2] - The UFL content of the salt definitely contributes to the corrosion é@{ probably because of the reduction of UFh to UF, by oxidation of chromium. A 3 reduction of uranium concentration from 4 mole percent to the less than one per- cent suitable for a power reactor will substantially reduce the corrosion rate. These data hafie been interpreted to indicate that with the 1200°F peak temperature contemplated for a power reactor, the maximum depth of void formation would not exceed 10 mils in one year or 20 mils in three years with Inconel. The thinnest hot metal section will occur at the heat exchanger tubes, and these could be designed to allow for a 3-year operation, as far as weakening due to corrosion is concerned. Tests on salts other than Composition-30 against Inconel have in most cases yielded higher corrosion rates. In particular, lithium fluoride salts are about twice as corrosive as Fuel-30 when tested against Inconel at a metal temperature of 1600°F 22/. Under these conditions the IiF salts form a dendritic growth of Cr in the cold leg. A great many tests have been run on low chromium, low iron, nickel- molybdenum alloys, but most of these have been at 1600°F wall temperature or hi gher, and have been for fixed periods of 500 to 1000 hours. These nickel- molybdenum alloys uniformly give low corrosion depths of the order of 1/2 to 2 mils under these conditions, and are thus estimated to be at least 5 times as registant to corrosion as Inconel is to Fuel-30. Most of the tests on the nickel- molybdenum alloys have been with the more corrosive salts containing lithium fluoride; within the limits of the low corrosion rates, there does not seem to be much dependence on the particular molten salts used. 32/ ORNL-1864 33/ ORNL-2217 - 22 - If this superiority of performance can be extrapolated to lower tempera- tures and longer times, the nickel-molybdenum alloy INOR-8 can be expected to produce reactor components which should last many years before failure due to corrosion. Obviously, what is needed here is some long-term testing under the temperature and radiation cbnditions desired for a power reactor. Other alloys now being examined by ORNL may prove even better for the power reactor than INOR-8. These alloys do not contain chromium, and the ~mission of chromium may result in complete thermodynamic stebility of the container =:d salt. There remains the question of mass transfer attack as a result of tempera.- ture gradients. One of the alloys being tested substitutes niobium for chromium in the INOR-8 type alloy and is currently being called INOR-9. It is not clear, however, just how much work on new alloys will be undertaken without active sponsorship from a power reactor program. Other Properties - Inconel has the following nominal composition: Ni - TT% C - 0.08% Cr - 15% Mn - 0.25% Fe - 7% S - 0.007% Si - 0.25% Cu - 0.2% Its melting point is 25&0-2600°F, Some of its properties 25/ in the range of temperature suitable for a power reactor are given in Table V. Inconel is a commercial alloy and fabrication techniques are well established. In the experimental program on INOR-8 carried out at Osk Ridge, a range of composition was considered. Table VI gives the properties of INOR-8 as of January 15, 1957 22/@ 34/ PWAC-56L 35/ ORNL-CF-57-1-108 - 23 - Table V REPRESENTATIVE PROPERTIES OF INCONEL Ultimate strength: Strain Ultimate strength at 1350°F 0.04"/" /min 37,500 psi 0.2 "/"/min 45,000 psi Stress to rupture: 1000 hour stress rupture strength 1200°F 14,000 psi 1350°F 6,000 psi Other physical properties: elastic modulus 25.5 x 106 psl at 12009F Poisson ratio 0.32 at 1200°F thermal conductivity 12 Btu/hr-£t-°F at 1200°F mean thermal expansion coefficient 8.8 x 10°°/°F 32%F - 1300°F - mn e Emam Ay amomm Table VI PROPERTIES OF INOR-8 (This table is based on experience with the following compositions) Mo Cr Fe Other oM Composition Range 10-24 3-10 4-10 0.5 Al; 0.5 Mn; 0.06 C Bal. (Wt percent) Number of Compositions 18 Number of Heats 25 Fabricability: Between Inconel and Hastelloy B, depending on the composition. Oxidation Resistance: Rates for 6 percent Cr alloy O.4 to 1.6 mg/cm?/lOO hr at 15000F. Oxide scale is borderline with respect to stability (7 percent results in a stable scale). Joining: Weldability between Inconel and Hastelloy B. Brazeable in dry H2. Stress Rupture: 11 tests, 3 stress levels, argon, Fuel No. 107, and 2 conditions. Rupture life at 8000 psi and 1500°F: 242 - 900 hr. Corrosion: Fuel No. 30, 1000 hr, 1500°F, 1 mil Fuel No. 107, 1000 hr, 1500°F, 1/2 - 2 mils Type of attack in molten salts - subsurface voids. Mass transfer detected in sodium loops. - 24 - (Table VI - continued) . Tensile Properties: Alloys containing 20 percent Mo or less show no tendencies to be brittle; 9 tests, > temperatures, and 3 conditions. > Test Temp. Yield Point, 0.2% Ultimate Elonagation F Offset, psi T.S5. psi percent RT 46 ,000-47,000 114,000-13%0,000 y7-51 1300 31,000-32,000 63,000-T73,000 13-21 As a result of the data leading to this tabulation, the following is chosen as the nominal composition of INOR-8 for larger scale tests: Alloy INOR-8 (nominal composition, weight percent) Mo: 15-17 Mn: 0.8 max Fe: 4-5 Si: 0.5 mex Cr: 6-8 Ni: Balance c: Oool‘l"‘o 008 — mm mWe mm wma vemm sems 5. Moderator Materials The highest breeding ratios for molten salt thermal reactors will be obtained with the use of & supplementary moderator. Although the salt Mixture-Ti4 . has a slowing down power of 60 percent that of normal reactor graphite, its slow neutron absorption is about 4 times that of graphite, and the lower neutron absorp- tion of the mixed moderator and salt is definitely beneficial in breeding ratio. Moderator materials_that can most readily be considered in this high temperature system are beryllium metal, beryllium oxide, and graphite. The former two require canning for protection from the salt, and since the canning materials have rather high neutron cross sections, their value in obtaining an improved neutron economy is questionable. Uncenned graphite is the best hope for a moderator that will provide good neutron economy. Several graphite samples have been tested which show very little pene- tration by the molten salt in short-term tests éé/, It is ressonable to expect 36/ ORNL-2221, p. 158 O - 25 - that one or more of these grades could be made commercially if longer term tests confirm their penetration resistance, particularly under the effects of radiation. The other problem associated with the use of graphite in direct contact with the salt is whether or not the metals of the system will become carburized. Preliminary tests with Inconel and Fuel-30 show no detrimental effects in 100 hours 21/, Further testing with other metals is underway, and there is hope that a suitable metal system for use with bare graphite and a molten salt will be developed. However, much more experimental work is required to demonstrate the practicality of a molten salt-unclad graphite system. B. Materials This section is a survey of the factors involved in selecting com- ponents, including an appraisal of the experience in actual operations. All equipment and controls which are parts of, or directly affect, the reactor: molten salt systems, liquid metal systems, and radicactive gas systems, require components of much higher quality and greater dependability than has been generally accepted as "commercial practice" in steam-electric plants. It should be strongly emphasized that the high quality demanded of these systems is actually necessary, and must be achieved because of the difficulty of maintenance of highly radioactive systems. ZExperience at the development laboratory level has shown that, with proper attention to detail, the standards of quality required are prac- tical of attainment. Specifications and procedures have been rather thoroughly developed and detailed. The importance of the strictest possible quality control and inspection of absolutely every step can hardly be overemphasized. 37/ ORNL-2221, p. 187 - D6 - All items not involved in handling molten salts, molten metals, or radioactive solids, liquids; or gases require only the same standards of quality as conventional installations. This includes turbines, condensers, feed-water heaters, water pumps; generators, and electrical gear, together with their instrumentation, controls, and auxiliary equipment, and most of the building and building services. 1. Pumps The choice of pumps for liquid metals is among three proven types: (1) electromagnetic; (2) canned rotor; and (%) gas-sealed. The frozen seal 38/ type may be sufficiently proven within the next two or three years to add to the list, when more actusl operating data are available. Electrcmagnetic pumps of several different designs have beefi success- fully operated with little or no trouble I>r several years. Among the larger users are: General Electric Company at KAI'[; Aircraft Reactor Engineering Divi- sion (ARED) at ORNL; North American Aviation Company (Atomics International); Westinghouse; Mine Safety Appliances Company, Callery, Pennsylvania; Brookhaven National Iaboratory; and Argonne National Iaboratory. Among other users are Battelle Memorial Institute; Allis-Chalmers Company, and Babcock and Wilcox Company. Most applications have been at temperatures below 900°F, and at rather low capacities of only a few gallons per minute. Some experience exists at ORNL and other places at temperatures up to lSOOOF and higher. EM pumps of 5000 gpm at 100 psi and even higher capacities are advertised for sale by General Electric Company, Westinghouse Electric Company, Allis-Chalmers Company, and perhaps others. Very recently, EM pumps with efficiencies greater than 20 percent are reported to have been developed 22/. o 38/ NAA-SR-1804 (TID-7525); ANP Quarterly Reports 39/ Information from Broockhaven National Iaboratory, etc. - 27 - Canned rotor pumps for liquid metals have been developed in the past three or four years by Westinghouse, Allis-Chalmers, and others for high perform- ance operation. Some difficulties inherent in canned rotor pumps are: friction problems in bearings during start-up and shutdown operations; low efficiencies; thin and relatively fraglle walls between stator and rotor, and the requirement for keeping the temperature of rotor and stator low. Both EM pumps and canned rotor pumps have the advantage of being "sealless", that is, there is no communicsation via packing or gas passages to the surrounding atmosphere. This is a major engineering advantage in pumping oxygen-~-gsensitive liquid metals, obviating any need for more or less elaborate sealing mechanisms. Each has the disadvantage of low efficiency and lack of long~time proof testing with liquid metals at temperatures above 700 or 800°Fo Gas-sealed centrifugal pumps &9/ are at present the most adequately proven pumps for moving either liquid metal or molten salts at temperatures above 8OO°F° Efficiencies are good; no different than for centrifugal pumps in general. t The limits of head and capaclty are similar to those of the ordinary centrifugal pump. More than usual "overhang" (distance from nearest bearing to impeller) is required, so that the seals and bearings can be maintained at a low temperature. Gas-sealed pumps were used exclusively for molten salt and liquid metals in the ARE operation, and in the scores of high perfofmance, high temperature (lOOOoF to 1700°F) heat exchanger tests, pump tests, pumped system tests, in-pile loop tests, etc., operated by ARED for pumping both molten salts and liquid metals Ei/o 40/ Complete details of design and performance are available from ORNL-ANP 41/ See ANP Quarterly Reports - 28 - The largest capacity high temperature pumps used to date by ANP are about 1200 gpm at 370 feet of head at 12000F and higher temperatures. Pumps of this type are advertised commercially in capacities to 20,000 gpm. Their degree of reliability is not known at ORNL. The two most vulnerable parts of gas-sealed centrifugal pumps are (1) the seal and (2) the bearings. To date, at ORNL, bearings are lubricated by cireulating oil, & small part of which is also used to lubricate the seal. Both the seal and the nearest bearing are usually within 18 inches of the pumped liquid. Therefore, pumps for radiocactive fluids require shielding of the bear- ing and sesl region and provision for replacement of the lubricant as it gradually becomes damaged by radiation. Provisions for such replacement are included in the ART pumps and lubricant systems designs Eg/, Gas-sealed centrifugal pumps have been operated by ANP at 1200°F for durations up to 8000 hours without bearing, seal, or other pump mainten- ance. The operation of such pumps is now considered by ANP to be routine and trouble-free. While some work is continuing there on the further refinement of seals and shielding, the chief present use of such pumps in the Experimental Engineering Division is to serve tests of other components requiring pumped hot liquids {e.g., heat exchanger and radiator tests). Possible future improvements in gas-sealed pumps for high temperature liquids include the use of hydrodynamic bearings and liquid centrifugal seals. Hydrodynamic bearings could reduce or remove the need for compromising between "overhang” from the nearest oil-lubricated bearing and radiation damage to the lubricant. Before using in a reactor system, a development program involv- ing tests at operating temperatures, flows and heads would be required. A difficulty is the close clearance, without rubbing, required between rotating and static parts. 42/ "Design Report on the ART", ORNI-2095, p. 29 et seq. o A - 29 - Iiquid seals similar to mercury seals sometimes used in Laboratory bench tests would allow relaxation of tolerances now required in the mechanical seals employed. A problem is the length of available sealing annulus required for handling possible pressure surges. The combination of a liquid seal with & hydrodynamic bearing would appear a good future prospect for power reactors. Where two or more gas-sealed pumps operate in parallel;, free flow between the liquid-to-gas interfaces in the pumps must be provided in order to prevent surging of the level in one pump significantly above the levels in the other pumps. The close coupling can be accomplished by providing large connect- ing flow channels to maintain approximately equal levels in all the parallel pumps of any one system, or, preferably, by containing the parallel pumps in one single reservoir. Pump design is well enough understood that larger pumps of a proven type can be designed so that they can be confidently expected to deliver the flows and heads calculated. However, a modest development program should be expected and a series of thorough prove-in tests will be required before pro- curing larger pumps on a routine basis. s 2. Heat Exchangers Heat exchangers for a power feactor will differ greatly in general design from those used in the ANP reactors. The removal of space and weight limitations allows the heat exchangers to be designed to minimize stresses and reduce construction difficulties. On the other hand, the ANP date on heat transfer coefficients and fluid flow, the metallurgy involved in fabrication, and the chemistry of corrosion are invalusble in the design of a power reactor heat transfer system. - 30 - To minimize thermal stresses, the heat exchangers should normelly be of U-shaped construction. The two tube sheets are then side by side, and the individual tubes are free to expand longitudinally with very little consequent stress in the tube-to-tube sheet joints. Countercurrent flow, which has two advantages, is possible with the U-shaped configuration. Counter flow exchangers, in any given application, require less heat transfer area, and have lower maximum local temperature dif- ference between the two fluids than for any other configuration. The maximum temperature drop through the tube wall;, and thus the tube wall thermal stress, is consequently lower. Although thermal stresses due to temperature differences between the internal and external walls of the exchanger tubing tend to be relieved by creep under steady operating conditions; the unavoidability of some thermal cycling requires that such stresses hbe considered. They can be minimized by using thin tube walls which yield a small temperature difference between the inner and outer surface. The small temperature drop through the wall is desirable in itself, because it allows higher steam temperatures for the same fuel temperature. These factors must be balanced against choosing the tube wall thickness for maximum reliability. Tube walls greater than 20 percent of the diameter are poor from the tube fabrication standpoint, but for small diameter tubes, this 20 percent limit should probably be approached closely to minimize weakening effects asso- ciated with high temperature immersion in the molten salt. Sample calculations have shown that where minimum fluid volume has great importance; the tube diameter should be as small as the technology will allow for economical and reliable manufacture. The ease and reliability of manufacture as a function of heat exchanger tube diameter have not been accu- rately assessed. However, ANP experience with Inconel indicates that simple - - - 31 - fusion welds can be made for tubes between one-quarter inch and one-half inch in diameter, and that for appreciably larger tubes, filler metal will have to be added while welding. The present indications are that 5/16-inch and 3/8-inch diameter tubing may be desirable. ANP experience indicates that the weakest places in a heat exchanger are in the neighborhood of tube-to-tube sheet joints, where stresses tend to be a meximum. Thus special attention must be given to the design and metallurgical aspects of this problem.. For the molten salt heat exchangers, the following types of jolnt are all possibilities, and the one chosen wi;l be determined by further development. a. Face welded joint The tube sheet is machined so that through most of its thickness, the hole diameter is appreciably larger than the tube diameter. The remaining thick- ness of the tube sheet is drilled to match the tube 0.D. closely, and a trepan is machined to leave a thin section of the tube sheet to be welded to the tube end. This type joint can be fusion welded for tube wall thicknesses no greater than 0.060 inch. The presence of a crevice on the outside of the tube, where it enters the tube sheet, precludes the use of such a Jjoint with water or steam on the shell side of the exchanger, since chlorides and other impurities will collect in such cracks and cause embrittlement. No comparable phenomenon has been observed with sodium or molten salt heat exchangers, however, and such a . crevice may not be detrimental in them. b. DBack brazed joint Brazing, or seal welding with back brazing, has been used extensively in the ANP program. This avolds stress concentrations in the joints which might lead to cracking, but the metallurgy of the braze may cause difficulty for - 32 - long-life applications. Diffusion of braze metal constituents into the tube wall may cause embrittlement. For the salts, 100-hour tests on gold-nickel alloy brazing have indicated corrosion resistance and little tendency to diffusion Eé/o ¢c. Butt-welded joint In this joint, the tubing is butt-welded to a nipple which may be forged on the tube side of the tube sheet. The resulting joint is perhaps the most satisfactory of all, but the fabrication is the most difficult, particularly in small size tubes. Further testing is required before this type joint can be considered as proven technology for the desired tube sizes. All salt-to-salt heat exchangers should be constructed entirely of INOR-8 for resistance to corrosion. It is probable that salt-to-sodium heat exchangers can be made from INOR-8 also; however, this is subject to confirma- tion of low mass transfer of nickel in sodium circuits by tests over long periods of time. In the event that excessive mass transfer rates are encountered,; a duplex construction, with INOR-8 on the salt side and 316 stainless steel on the sodium side, can be used. Construction with duplex tubing is more difficult and probably less reliable, but has been used in other applications. The situation in regard to sodium-to-water or steam boilers or super- heaters is being intensively studied by several organizations at the present time, and it is anticipated that considerable progress in defining the best solution will be made in the next two years. The basic difficulty is that leaks develop, probably starting as small stress-corrosion cracks originating from the water side. Both impure water and impure sodium are very corrosive, so that a 43/ ORNL-1934 -33- small leak quickly spreads. The problem is either to achieve perfection in leak-proof construction, or to design the system so that small leaks can be detected quickly. One way of accomplishing leak detection is to have a double tube construction, with a leak detecting fluid between the tubes. In one case, mercury is used as the intermediate fluid; others are trying an inert gas be- tween tubes that meke a partial metallic contact. Either of these solutions requires large area boilers and superheaters because of the decreased heat transfer coefficients. Another approach is to use single walled tubes, but to use double tube sheets, with the tubes sealed to each tube sheet. The idea behind this is that leakage will be most probable into the space between the two headers and can be detected there. This approach has been used in a steam generator on the HRT project, and can be adapted to sodium-to-water boilers and superheaters. Inconel has been suggested as a good material for the boilers and superheaters, since it is not ordinarily subject to stress-corrosion cracking, and the consequences of a small leak should be considerably lessened. Alternately, ferritic steel, such as 2 1/4 Croloy, could be used in the boiler and either a ferritic or an austenitic stainless steel, such as type 316, could be used in the high temperature service of the superheater. This is standard power plant practice and can be adapted directly, since both metals are compatible with gsodium. If the austenitic steel is used for the superheater, it must be pro- tected from any but dry steam. For this reason, a once-through boiler, delivering slightly superheated steam to the superheater, is indicated. 3, are: Reactor Vessels The design requirements of the core vessel of a circulating fuel reactor (1) geometry to fulfill nuclear requirements; (2) compatibility of container - 34 - material(s) and liquid(s); {(3) a fluid flow pattern which will, without excessive fluid pressure drop, prevent local stagnation or eddies that might cause harmful temperature transients; (4) low neutron poisoning by materials of construction; and (5) adequate safety against core shell failure from mechanical or thermal stress level or stress cycling. Fortunately, criteria (1) and (2) are relatively easily fulfilled in a circulating homogeneous molten salt fuel reactor system. Criterion (3) can be fulfilled for low power density cores (less than 250 w/cc) by employing the data and experience of the aqueous homogeneous reactor project and by tests in transparent prototype mockup of actual geometry and fluid flow types. Criterion (5) can be fulfilled by good choice of basic geometry and competent stress analysis and adequate design safety factors. Criterion (h) can be satisfied more easily for single region than for two region reactors because little or no material of construction is needed in the region of high flux. Possibly further improvements in this respect may be made in the future (e.g., if improvements in technology allow for use of graphite as the core shell material). One of the main advantages of circulating molten salt systems is the low pressure level and the consequent low mechanical stress. It should be noted here that the extreme complication of design of the interior of reactor cores which has already become traditional with solid fuel elements; moderators, control rods, etc., is virtually eliminated by the use of & homogeneous fuel, moderator, and, to some extent at least, reflector. a. Choice of number of regions Involved in the choice of number of regions are: (1) transmutation ratio potentiality; (2) ability to manufacture "pure" v or Pu; (3) flexi- bility of operation and experimentation (different composition fluids in core - 35 - and blanket); and (4) cost of inventory, fabrication, and processing. These aspects are discussed elsewhere in this report. More auxiliary equipment (storage tanks, fill-and-drain tanks, pumps, heat exchangers, instruments, ete.) will, of course, be required for a two reglon system. b. Single region reactor vessels For low neutron leaksges, single region reactors are large. In terms of spheres, calculations indicate that the optimum diameter is between 10 and 20 feet, probably about 1k feet &E/. Uranium concentration decreases with diameter increase and increases with increasing concentration of thorium. Note that the external volume of fuel is independent of reactor size, but external holdup of uranium is linear with concentration of uranium ifi the fuel. Container shape is an important fabrication parameter. Precise, final geometry will depend on optimized compromises determined by studies of: (1) nuclear characteristics; (2) fluid dynamics, including temperature transient studies; (3) stress analysis; (4) availability and fabricability of materials of construc- tion; and (5) costs. In general, spheres are optimum from nuclear considerations, cylinders for fabricability, spheres or spheroids for low stress, cylinders for flow pattern optimization. Stress considerations include weight, pressure, flow, and wall temperature patterns, and must therefore result in a compromise. Possibly the best geometrical compromise is approximately a right circular cylinder with ellipsoidal heads. One important design decision is the choice of relative locations of the entry and exit passages for the fuel. They may be placed at the same end 44/ The volume of & lh-foot sphere is 1437 ft2. Welght of fuel at 135 1b/ft” = 194,000 1b. Weight of l-inch shell at 500 1b/ft° = 26,000 1b. Power density at 600 Mw = 17.5 w/cc - % - of the reactor by making use of concentric pipes, or they may be at opposite ends for "straight-through" flow. The former appears to be better for avoid- ing stresses between core shell and blanket shell in a two region reactor. For a single region reactor, the latter appears to offer fewer complications. c. Two region reactors Because the transmutation ratio is such an important factor in the mills/kwh cost of electrical povwer, and because in two region reactors the core sfiell thickness has an important influence on the transmutation ratio (since it is a neutron absorber), it is very desirable to design for the thinnest core shell compatible with safety against failure. It is not believed that thermal stress will be a significant factor in determining the thickness of the core shell 52/ in low power density homogeneous reactors. Pressure stresses will probably determine the thickness required. These matters must, however, be investigated when firm data on heat flows and neutron and gamma fluxes become available. If the core and blanket fluids are kept at low pressures by appro- priate flow passage geometries and appropriate placement of the system pumps, the thickness of the core shell can be kept as low as 5/16 inch. At this thickness, the highest stress under normal operating condi- tions for the Reference Design Reactor described in Section TII would be a compressive stress of from 500 to 850 psi, depending on core shape. This is to be compared with estimated long-time creep strengths of several thousand psi at 1200°F. (Inconel, a weaker material than INOR-8 at these temperatures, has a 1000-hour stress-rupture strength at 1200°F of about 14,000 psi.) Accidental 45/ Personal communication from H. F. Poppendiek, formerly Chief, Heat Transfer and Physical Properties Section, ORNL - 37 -~ core or shell drainage could prodfice short-time stresses of from TOO to 2000 psi, depending on shape and condition. These stresses are to be compared with yield strengths of the order of 30,000 psi at 1300°F. Against collapsing tendencies, a 6-foot diemeter core with a 5/16-inch thick wall has a factor of safety of from 7 to 100, depending on whether the shape is.cylindrical or spherical. In two region reactors, it would probably be impractical to maintain the core shell and the blanket shell at the same temperature level at all times. Therefore, to prevent stress from differential thermel expansion of the two shells, they must be designed to be free to expand and contract independently of each other. This can be done by (1) employing straight-through flow with one or more flexible connections between the two shells, or (2) by connecting the two shells at only one end and directing the core fluid both in and out of the same end. Since flexible connections such as bellows or diaphragms are necessarily thin to allow flexibility, they are unavoidably subject to corrosion or mechanical damage. It is therefore preferable to avoid using them by direct- ing the core fluid into and out of the same end of the core, with no rigid restraint on thermsal expansion and contraétion of the core shell by the blanket shell. For thermal symmetry and resultant minimum stress concentration; the core fluld should enter and leave the core via concentric passages. Studies based on experience with aqueous homogeneous reactor flow tests, and considera- tions of design of the expansion chamber, pumps, and flow passages will determine whether the fuel should enter the core through the central tube and exit via the annulus, or vice versa. Blanket shells are not restricted in thickness by nuclear considera- tions and may therefore be designed to any thickness and geometry found desirable from stress, fluid dynamics, and support considerations. Because of the very low - 38 - power densities, restrictions on flow pattern can be relaxed considerably from those required in the core. Somewhat more freedom is allowable in the use of flow directing vanes near the outside of the blanket, where neutron flux and, consequently, poisoning effects are low. Appropriate blanket geometry can materially lessen shielding requirements. Twc region reactors may be supported from either top or bottom. Since core and blanket fluid pumps will probably be above the top of the reactor, it is anticipated that the main supports should be near the top of the reactor to minimize both vibration stresses and problems of thermal expansion. k., Other Vessels Vessels other than the reactor will be required, including: (1) special storage vessels from which each of the salt and liquid metal systems will be filled and to which they will drain; (2) storage tanks in which new and used salts and liquid metals will be stored; and (3) one or more tanks for the boiler make-up water, which will probably be of special purity and/or composition. Fill-and-drain tanks - Vessels in which fused salt for the fuel cir- cuit or blanket circuit will be stored must be designed for both heating and cooling. To minimize thermal shock, they should be preheated before receiving molten salt; and it will be desirable to maintain them at temperatures above the melting point of the salt whenever salt is in them. It must also be possi- ble to prevent undesirably high temperatures from rising in them from "after- heat” generated by beta and gamma decay of fission products. Preheating and cooling may be provided in a number of different ways as described later in Section II-C-L4. - 39 - Tanks for filling and draining non-radiocactive salts and liquid metal systems will require provisions for heating, but not for cooling, and may therefore be compact (right circular cylinders). Transfer of molten salt or metal between containers through pipes or tubes is usually accomplished by pressurization of the vessel to be emptied with inert gas. Pumps are almost never used for this purpose. Simple gravity drains may also be used, of course, where appropriate. 5. Jolnts and Valves Joints for contalning liquid metals, molten salts, or radiocactive gas require welding or brazing for satisfactory security against leakage. For long life in high temperature joints, no brazing meterial has yet been proven adequately safe; however, basic investigation of brazing is continuing and shows some promise for these applications. Welding has proven to be very satisfactory for both leak- tightness and long life. The high degree of competence that has been developed in welding and the adequacy of the welded joints of Inconel are evident in the lack of weld failures in the ARE and in the large number of circulating molten salt and liquid metal systems operated by ANP. Design and procedure specifica- tions for welds are spelled out in detail in the ORNL Welding Code (ORNL Metallurgy Division). A few vendors have already been trained in, and have met the require- ments of the code. For containment of radiocactive gases at low temperatures, both welding and brazing have been found to be satisfactory (zero leakage over long life). No other methods of fabricating joints have proven to be thoroughly reliable over long periods. Although many tests of API ring gasketed flanged joints and of compression fitting joints such as the Swagelok have shown leak-tightness over fairly long periods of non-cyclic operation, they are not considered to be depend- able enough for this type of service. - 40 - Valves for handling radioactive gases in small (less than 1/2-inch) lines do not present any unusual problems in attaining the required freedom from leaks to or from the surrounding atmosphere. This requirement is met satis- factorily by employing only bellows-sealed, welded or brazed connection valves. Fully adequate valves for these small sizes for gases with limited radiocactivity are available for "off-the-shelf" purchase. Valves for controlling gases with high beta decay activity may require redesign to reduce internal free volume, thereby reducing the amount of heat removal required. Flow control valves which are required only to control flow rates without entirely stopping flow in circuits of molten salt or liquid metal are ifems of semi-routine design and fabrication, except for the requirement of extreme leak-tightness to the atmosphere. Because of the requirement for leak- tightness, bellows-sealed, welded connection type valves are specified. Maximum inspection and acceptance testing are required, but no particular difficulty is encountered in meeting the specifications. Good mechanical design and the care- ful following of the weld design and procedure code are required. Valves for stopping flow of hot salts or liquid metals, with zero or almost zero through-flow required, are more difficult because of the requirement for seat materials which will neither stick nor be attacked by the fluid at high temperatures. This problem has been essentially solved by the use of cermet valve seats and the accurate alignment of parts, but the solution is still too new to be regarded as fully proven for temperastures of 12000F or higher until the results of the additional tests now in progress are known. Stop valves can be avoided in the hot molten salt and liquid metal circuits without serious difficulty (for example, barometric legs or "freeze valves" can be employed in- stead of mechanical valves in the sodium and salt drain lines). There are no significant difficulties involved in the use of stop valves at temperatures - 41 - lower than about llOOoF; therefore, such valves can be freely used in low temperature molten salt and in liquid metal drain lines. ORNL has developed and is successfully using stop valves in molten salt lines as large as 2-inch iron pipe size. Work toward further improvement is continuing, with emphasis on optimizing seat materials and design. 6. Instrument and Control Components Assessment of the adequacy of an existing instrument or control com- ponent for a reactor starts from consideration of the function of the system for which the component need be a part. Instruments are here evaluated in terms of thelr usefulness in systems which fall into four broad categories, according to their functions: those required for safe start-up, those vital for safe control of the reactor once it is in operation, those needed for monitoring and control of reactor auxiliary systems, and those installed for purposes of evaluating performance and gathering data for experimental or test purposes. Attention is focused here, in the main, only on sensory devices which are needed to operate at temperatures up to 1300°F in fused salt or liquid metal gystems. These include sensory devices for the measurement of temperature, liquid level, fluid flow, rotation speeds, neutron flux, pressure, and for the detection of lesks. a. Temperature sensors For sensing of high temperatures of interest, the thermocouple is the device upon which principal reliance has thus far been placed. Present ANP designs include a rugged chromel-alumel thermocouple of wire size B and S No. 8 or larger, coupled with magnetic amplifiers of reliability equal to that of A.C. transformers. Using fabrication and installation techniques developed in the ANP program, these thermocouples have demonstrated no - 4o . detectable drift in a number of 3000-hour tests conducted at 1200°F, Enough experience has been gained to Justify confidence that these temperature sensing systems will operate in the 1000°F range with a plus or minus SOF accuracy reliably over a 20-year period, in the absence of nuclear radiation. The calibration stability of these instruments over extended times in nuclear radiation fields has not been adequately tested. Consequently, at the present state of the art, there is no assurance that thermocouples installed in the core or blanket system would remain calibrated. Provision can be made in reactor design for insertion of a calibrated thermocouple to recalibrate installed thermocouples during periods when the reactor is isothermal, as, for example, during shutdown. Work has been done on & constant volume sodium or rubidium vapor thermometer operating in the form of a bulb or thimble which is inserted into the hot fluid. This device appears to be promising for measurement of the temperatures of molten salts subject to a radiation field. b. Iiquid level devices At least two satisfactory liquid level sensors are presently avail- able. These two have been tested for use in the ANP program. One is a resis- tance probe for use with liquid metals, in which the resistance varies with the level around the probe. The other is a float type &é/o Difficulties in installation of these sensors have been experienced in highly compact reactors due to access limitation. It should be possible to avoid this problem where space is not at a premium. 46/ Southern, A. L., "Closed-Trap Ievel Indicator for Corrosive ILiquids Operating at High Tbmperatures", Instrumentation, ORNL-2093 - L3 - These instruments have performed satisfactorily and reliably in 3000- hour duration tests, and there was no indication that they would not perform for a much longer period. They are not vulnerable to nuclear rediation damage. The equipment is sound in principle and promises to be satisfactory for many years of service. ¢. Fluid flow measuring devices Tests indicate the electromagnetic flow meter to be a dependable device for measurement of high temperature liquid metal flow. Care in design, fabrication, installation and calibration is needed, but the device is rugged and durable and accuracy of 5 percent should be maintained. The art of measurement of flow of molten salt at elevated temperatures is not as well developed. Iimited tests have been made on venturi and rotating vane types with varying degrees of success. Such devices, if needed in a power reactor; will require further testing. The ANP program has under development a rotating vane type which shows promise of reliable operation, and is similar to a8 type used successfully in the ARE. The application of venturi type flow meters depends only on successful application of pressure sensors. d. Pump speed indicators Pump speed measuring devices or tachometers which are durable and reliable are presently available. Access for 1nstallati$n on rotating machinery has proven the only difficulty encountered in the ART project. With restricted access, the more elaborate, but nevertheless reliable, pulse type tachometer must sometimes be used in preference to the simpler D.C. type tachometer. e. Nuclear sensors Nuclear sensors in molten salt reactors pose no problems not shared with other reactors. Existing and well tested fission, lonization, and boron - 4 - trifluoride chambers are available for installation at all points essential to the reactor. Their disadvantages of limited 1life can be countered only by duplication or replacement, and provision can be made for this; however, for circulating fuel reactors,; these instruments are not essential to the routine operation of the reactor. Nuclear sensors to withstand high temperatures are not needed. f. Pressure sensors and pressure controlling devices There exists a variety of pressure sensors of the diaphragm or mano- meter type which will almost certainly operate wherever needed in a reactor system. Among these, devices are known that will operate reliably over long periods of time in the absence of radiation; there are no tests which have demonstrated that they will operate reliably over long times when subjected to. radiation. There is no reason to believe that all of these pressure sensors will fail to operate satisfactorily in a radiation field. % Gas pressure controllers are required in molten salt reactor systems. Except for extended use in radicactive gas systems, adequate, well tested pres- sure controllers are available commercially for all applications. Under strong irradiation, valve seats or seals made of organic materials are suspect. The requirement tfiat the valve make a perfect seal is not essential in a design which incorporates pressure relief valves; metallic seats will probably give valves which will serve the purpose. The construction of such a valve seems straightforward, but a long-life test under radiation conditions will be necessary. g. leak detectors Leaks between a fluid system and its external surroundings, or between two fluid systems, tend to grow in size. To minimize serious damage which might result in a reactor, simple leak detectors capable of detecting very small leaks - 45 - are desirable so that action may be initiated in time to 1limit the size of the leak. ILittle progress has thus far been made in the development of leak detectors which will reliably meet these exacting requirements. The proper approach in design is to avoid relying on sensitive leak detectors, to design for maximum reliability against leaks and for minimum or no damage should leaks occur. Means of detecting larger leaks are of course available. C. Component Systems l. BSalt and Iiquid Metal Charging and Storage Systems Methods of handling molten salts and liquid metals in atomic energy installations have been developed from methods used by manufacturers of sodium, gasoline, and other chemically active or dangerous liquids. It has become almost universal practice to make batch transfers by either inert gas pressuri- zation or by gravity. Both sodium and salts are at present purchased less pure than required. Commercial sodium purity is such that, as & rule, only oxide removal is needed. This is accomplished by filtering the sodium at temperatures as little above its melting point as is convenient, and by "cold trapping" the oxide in operating systems (precipitating the Naeo on a cold wall at a selected point in the system). Salt purification is also accomplished in the molten state. After the desired proportions of salts have been mixed and melted under a protective atmos- phere, successive purifying and purging gases (helium, HF, hydrogen) are bubbled through the melt EZ/. The melt is then transferred into storage containers under inert gas and stored either molten or frozen until needed. 47/ ORNL-57-5k-6-126 - 46 - The liquid salt or sodium may be charged into the preheated and pre- cleaned systems either directly or, as is nearly always done, via fill-and-drsin tanks provided in each liquid system. The mechanics of such transfers is routine " at ORNL, KAPL, MSA, and a number of other installations. Treatments, transfers, and storage should, of course, be carried out in container materials compatible with the liquid used. A closed circuit of especially purified water, extremely low in chlorine and oxygen content, should be used in the water-steam system Eg/’ &2/. One or more storage tanks should be provided for make-up water. Considerations in locating storage or fill-and-drain vessels are: (1) radiocactivity, and required shielding and off-gas handling facilities; (2) accountability for 11255 and other accountable materials; (3) convenience in storing and handling, including ease of access, mobility, length of transfer lines, cost of heating, etc.; (4) hazards other than radiocactivity, such as biological poison; high temperature, etc. 2. O0Off-Gas Handling A major advantage of circulating fuel reactors is the ability to keep the concentration of xenon and krypton at a low level by continuously removing them from the fuel system zg/o Their removal greatly reduces: nuclear poisoning 48/ Williams, W. Lee and Eckel, John F., "Stress-Corrosion of Austenitic Stainless Steels in High Temperature Waters", Journal of The American Society of Naval Engineers, Inc. (February 1956) 49/ Wilson, R. M. and Burchfield, W. F., "Nickel and High Nickel Alloys for Pressure Vessels", No. 24, Welding Research Council Bulletin Series (Janvary 1956) € ORNL-CF-57-1-8, "ANP Chemical Section Progress Report", p. 8; ORNL~192k; ORNIL-2116; ORNL-2095 - 47 - of the reactor by these gases (and by thelir descendents which would have appeared bad the gases not been removed); the difficulties and hazards of re-starting after shutdown; and the biologlcal hazard of any leak or other accident which might open the system and allow leakage of fission gases to the surrounding atmosphere. Removal of gases can be accomplished by by-passing part of the fuel flow through a compartment having a liquid-gas interface, accompanied by aglta- tlon of the liquid in the compartment, by bubbling helium through the liquid, or both. There must be at least one such compartment in any liquid fuel circuit to allow for thermsl expansion of the liquid. The gases so removed will in general be too radiocactive to discharge directly into the air, and must be "stored" until their activity has decayed to a level acceptable for discharge into the air. Decay holdup time may be provided by directing the gases through a large volume of appropriate geometry (e.g., a very long pipe), or by directing them through appropriately designed charcoal packing, in which the xenon and krypton will be held up by absorption on the charcoal. Experiments and designs El/ to date indicate that initial holdup in a gas volume to allow decay of short half-life nuclides, followed by holdup in charcoal, is the preferable method. These studies, together with engineering anslyses 22/, have spelled out the required parameter relations and design criteria. It may prove to be feasible to recirculate the helium instead of dis- charging it to the atmosphere, if the gas is found in operation to be clean and pure enough after the holdup operation. This is, however, only a future possi- bility, as yet unproved as to feasibility. 51/ HRE, HRT and ART 52/ ORNL-2116; ORNL-192k 3. Inert Gas System The molten salts and liquid metals under consideration require com- plete protection from contact with oxygen and water vapor. All liquid-gas interfaces must therefore be protected by inert gas blankets. Systéfis of high pressure gas stdrage containers; pressure-reducing valves, check valves, flow control valves, instruments, and tubings and fittings are required to direct the gas to the required locations at the desired pressures and flow rates. A very considerable amount of experience in deéigfiing and operating experiments with precisely the same requirements has been accumulated at ORNL 22/ and BMI. Extremely high purity EE/ helium, argon and nitrogen are the most frequently used gases for these purposes. At ANP, helium is the gas nearly always used, primarily because it is the cheapest inert gas of the required purity. 4, Heating and Cooling of Components Since the melting points of the salts and sodium are above ambient temperatures, it will be necessary to provide heating for all parts of the system which will contain either sodium or salt. It will also be necessary to provide heating for the water-steam system. Components subject to after- heat (beta and gamma decay heat) will in general require provision for heat removal. Heating and cooling design should, of course, be such as to minimize rapld transients or large gradients in temperature where high thermal stress 53/ ORNL-2095 54/ <10 ppm oxygen and < minus T0° dew point - 49 < would result. This restriction is common to all conventional high temperature component design. In the few locations where beta and gamms heating may be large enough to have significant effect on temperature gradients, it should be taken into account. Several methods of preheating have been used successfully, including: (1) use of electrical strip heaters, heat tubes, ceramic-protected hot wire type heaters, etc., in which the heaters are clamped or otherwise fastened to the parts to be heated; (2) direct resistance heat, in which the components to be heated are made a part of an electrical circuit; (3) gas heating; (4) induction heating 22/; and (5) steam heating (used in England for sodium system heating). In most high temperature salt and sodium systems at ORNL, adequate design and fabrication for preheating using Methods (1) and (2) are now semi-routine. No significant difficulties in preheating or maintaining temperature have been encountered in systems in which the entire system was carefully engineered (e.g., the scores of large and small circulating salt and liquid metal systems success- fully operated at ORNL in the past three or four years). The most common difficulty has been failure to take account of local heat sinks, such as support connections, flanges partly exposed to ambient atmosphere, etc. Good engineering of thermal insulation is clearly required. Cooling radioactive molten salt is not routine because a volume heat source is involved. The use of small diameter vessels such as tubes eases the problem because it provides a short distance for heat to go from the center of the liquid to the cooled surface, and because it increases the surface area of 4 the vessel. When appreciable after-heat is being generated, the vessels should 55/ Used to a considerable extent in the SRE (NAA - Atomics International) ¢ - 50 = be cooled with forced circulation, and for reliability, auxiliary power sources . should be available for this purpose. It is desirable to design the system so that maximum use can be made of naturel air convection, and thus minimize the time required for use of the forced circulation and cooling system. Simple means of biological protection are also desirable. Both of the latter aims could be implemented by use of an underground tunnel with a stack at the end of the tunnel. D. Nuclear Considerations 1. Previous Work and Early Consideration Burners - Molten salt 0235 burner resctors for mobile power have been extensively investigated at Oask Ridge on the ANP program, and these studies pro- vide the foundation for the present investigation. In 1953 a'group of students under the leadership of T. Jarvis §§/ at ORSORT investigated the applicability . of molten salts to package reactors. More recently (1956), another ORSORT group ¢ led by R. W. Davies has prepared a valuable study of the feasibility of molten salt U255 burners for central station power production 21/, Fast Breeders - Fast reactors based on the U258-Pu cycle were studied by J. N. Addoms et al of MIT §§/’ and, more recently, by yet another ORSORT group led by J. Bulmer 22/. Both groups concluded that it would be preferable to use Jarvis, T., et al, "Fused Salt Package', ORNL-CF-53-10-26, Secret Davies, R. W., et al, "A 600 MW Fused Salt Homogeneous Reactor Power Plant", ORNL-CF-56-8-208, Secret Addoms, J. N., "Engineering Analysis of Non-Aqueous Fluid Fuel Reactors"”, MIT-5002 (1953) € I8 Bulmer, J., et al, "Fused Salt Fueled Breeder Reactor", ORNL-CF-56-8-20k, Secret . - 51 - molten chlorides rather than the fluorides on account of the relatively high moderating power of the fluorine nucleus, although it was recognized that the chlorides are probably inferior in respect to corrosion and radiation stability. Furthermore, Bulmer pointed out that it would be necessary to use purified Cl57 on account of the (n,p) reaction exhibited by c1”°, In view of these disadvantages of the chloride systems, and, further, in view of the fact that the technology of handling and utilizing Np and Pu bearing salts is largely unknown, it was decided to postpone consideration of fast chloride salt reactors. Epithermal Breeders - In 1953 an ORSORT group led by D. B. Wehmeyer §9/ analyzed many of the problems presently under study. Many of the proposals set forth in that report have been adopted in the present program. A study by J. K. Davidson and W. L. Robb of KAPL él/ has been most helpful, also. Both this and the Wehmeyer study concerned the possibility of using thorium in a U253 conversion-breeding cycle at thermal or near thermal energles. A consideration of molten fluoride reactors based on the Th--U255 cycle points up the fact that U255 is not available in sufficient quantity to provide the initial charge for a breeder reactor. While conceivably the U233 required could be made in a production reactor, the uncertainties in cost and time of availability militate against designing a reactor to be started up with U253 in the near future. It would appear that, whatever system was selected, the initial charge would be composed of 93 percent enriched U255. 60/ Wehmeyer, D. B., et al, "A Study of a Fused Salt Breeder Reactor for Power Production", ORNL-CF-53-10-25, Secret 61/ Davidson, J. K. and Robb, W. L., "A Molten Salt Thorium Converter for Power Production", KAPL-M-JKD-10 (1956), Confidential - 52 - It is well known that the variation of 73 for U235 with energy impairs ¢ the éonversion ratio of a reactor utilizing U235 and Th, and operating in‘the epithermal neutron energy range. On the other hand, the low moderating power of the fluoride salts (including IiF and BeF2) makes it impossible to design a high performance, homogeneous converter in which, say, 90 percent of the fissions in U235 are caused by thermal neutrons. Parasitic capture in the fuel carrier, etc., would be excessive. It was concluded that heterégeneous cores would be required in thermal converters or breeders to obtain breeding ratios approaching 1.0. None of the container materials under consideration for use with molten fluorides would be satisfactory for the canning of moderators for a thermal breeder reactor. The parasitic absorptions would be intolerably high. At present; graphite is the only suitable moderator that éhows promise of being compatible with the salt, and considerable development work will be required to establish its usefulness. Therefore, consideration of heterogeneous,; thermal, molten salt reactors has been postponed. It was decided to investigate the nuclear properties of homogeneous, 233 epithermal, one and two region, molten salt, U converter-breeders. The in- vestigation to date has been exploratory in nature and most of the work has been centered on two region systems. The calculations were handicapped by a lack of data on nuclear cross sections in the epithe¥mal range and lack of a computational method that would take into account inelastic scattering, resonance saturation, and Doppler broadening. The first calculations were performed on one region reactors by hand,; using cross sections then available in the Univac program Eyewash ég/, which was originally written for the analysis of aqueous homogeneous - 62/ Alexander, J. H. and Given, N. D., "A Machine Multigroup Calculation-- The Eyewash Program for Univac", ORNL-1925 (1955) ¢ - 53 - reactors. The first calculations using the computer were for simple, optimistic cases; i.e., presence of U238, fission products, protactinium, etc., were neg- lected, as well as the presence of 116 in the carrier. Iater, the cross sections were revised on the basis of latest information, including the effects of resonance saturation and Doppler broadening. The lethargy intervals were modi- fied, and other changes were made to increase the amount of information to be obtained from the computer (Ocusol-A program) éé/. These facts should be kept in mind when comparisons of Univac results for the various cases are made. Tfie advantages of a U235 burner versus breeder-converters were con- sidered briefly. Devies et al éfl/ proposed to operate the burner without reprocessing of the irradiated fuel. Builld-up of fission products and other 35 parasitic materials was to be overcome by the addition of U2 in excess of that rzquired to replace the U235 consumed in the reaction. Depending on the assumptions made regarding the cross sections of poisons in the intermediate range, it was found that the reactor would operate from 5 to 20 years before it would be economically advantageous to dump the spent fuel and recharge. A study of the effects of adding fertile material to the Davies' system disclosed that the performance would be improved by the addition of even moderate amounts of thorium to the core. The amount of U235 required to replace U235 destroyed and to override nuclear poisons is reduced because the 0233 formed not only replaces U235, but also has a higher fission cross section and lower absorption cross section in the intermediate range. It was concluded 63/ Alexander, L. G., Roberts, J. T. and Carrison, D. A., "The Univac Program Ocusol-A, A Modification of Eyewash", ORNL-CF- (in preparation) 64/ Davies, R. W., et al, "A 600 MW Fused Salt Homogeneous Reactor Powver Plant" ORNL-CF-56-8-208 (1956), Secret - 54 - that conversion-breeding is desirable. It remained to be determined whether the . ultimate power cost can be further reduced by reprocessing the spent fuel. Some effort was expended in determining whether it is possible to obtain & breeding ratio of 1.0 in a homogeneous reactor employing pure U233. Although the results are not definitive, it appears that it may barely be possible. It seems fairly certain that a breeding ratio of 0.9 can be obtained, and it is felt that it would be economically feasible to compensate for the breeding deficiency by purchasing U253 at a premium price. Since, however, U233 1s not presently available in amounts suffi- cient to fuel a full-scale power reactor, attention has been concentrated mainly on U233--U255 breeder-converters, in which the deficiency in production of fissile material is compensated by the addition of U235. It is clear that the optimum conditions will depend to a large extent on the cross sections of U235, U233, and Th in the epithermal range. The U255 * cross sections are reasonably well known. In the hand calculations, and in the first series of Eyewash calculations, an n of 2.28 for U253 was used ’ uniformly from thermal to 0.2 mev. This was consistent with the then latest published data of Spivak et al éz/’ éé/. More recent data by Magleby et al éZ/, and Moore et al é@/ for energies ranging up to 8 ev disclose a complicated 65/ Spivak, P. E., et al, "Measurement of 7 for U-233, U~235 and Pu-239 with Epithermal Neutrons", Sov. J. of Atomic Energy, 1, 13 (1956) Spivak, P. E., et al, "Measurement of 7 for U-233, U-235 and Pu-239 with Neutrons in the Energy Range from 30 to 900 kev", Sov. J, of Atomic Energy, 1, 18 (1956) Magleby et al, "Energy Dependence of n for U-233 in the Region 0.1 to 8.0 ev", PTR-142 (1956), Unclassified & S Moore, M. S., et al, "Uranium-233 Resonance Parameters for Neutron Energies Below 4.0 ev", PTR-141 (1956), Unclassified & - 55 - resonance structure in the range above 2 ev, and it appears that the avérage n in this range cannot exceed 2.1. Recent KAPI data §2/ indicate that even at 30 ev, n does not exceed 2.21. On the basis of conservative estimates for the ranges where data are lacking, new values of the absorption cross section of U253 were computed. Table VII VAIUES OF n FOR U>> USED IN THE UNIVAC CALCULATIONS Neutron Energy, ev Eyewash Ocusol-A 0.076-0.125 2.28 2.28 0.125-0,.186 2.28 2.08 0.186-0.92 2.28 2.28 0.921-10.2 2.28 2.08 10.2-33.7 2,28 | 2.13 33.7-67,000 . 2.28 2.21 67,000-183,000 2.28 2,25 183,000-820,000 2,49 | 2.4k 820,000-10" 2.46-2.52 2.52 —. S Sem SEm e G eemm ame In Figure 1 are graphed the fission cross sections and n's used in the Ocusol-A program for U‘Q53 and U235. It 1s seen that, with two exceptions, the average fission cross section of U253 is substantially greater than that of U235 in all lethargy intervals. Also n is substantially greater at nearly all lethargies. It seems clear that reactors fueled largely with U253 will give the best neutron economy when the majority of the fissions are caused by neutrons in the epithermal range, where parasitic absorptions in fuel carrier and other materials are relatively less important. 69/ Knolls Atomic Power laboratory, "Report of the Physics Section for June, July end August 1956", KAPL-1611 (1956), Confidential 56 UNCLASSIFIED ORNL~LR-~DWG 20325 233 2.5 = —H .} BEs=————"" =TT ! & 2.0 — L n U235/|__ r s ] | [ ] 200 [; u233\ £ 100 nl_ 1 235 < i ~ i L O - - 0 {0 20 LETHARGY FIGURE 1 Fission Cross Sections and Eta for U233 and U235 in the Ocusol- A Program. - 57 - The average energy of neutrons causing fission can be increased by 35 increasing the U2 concentration, which increases the probability for a neu- tron to cause fission before it gets slowed down very much. The resulting increase in reactivity can be compensated for by adding thorium to the core or by using smaller cores. The improvement in breeding ratio must be bvais.ced against increased inventory charges. 2, One Region Reactors The critical mass of a bare reactor fiasses through a minimum with increasing diameter at the point where the effect of diminished leakage on critical concentration is just compensated by the volume increase. The minimum inventory for the entire system occurs at a somewhat larger diameter which depends on the ratio of total fuel volume to core volume. By means of preliminary hand calculations, a reactor having a diameter of 10 feet was selected as provid- ing a reasonable compromise between the demands for low inventory and reascnable size. The power density in the four cases studied was set nominally at 100 w/ce 4 of core volume. The results are summarized in Table VIII, and discussed in detail by Carrison and Alexander 19/. Pure U233 was used and neutron losses to fission fragments, protactinium, uranium isotopes, and Lié were ignored. Further, the optimistic U233 cross sec- tions of the Eyewash program were used. Thus the results are only qualitative. However, it does not seem probable that a breeding ratio of 1.0 can be obtained in a practical, bare, one region system. On the basis of the relative cross gsections and from the comparison of certain two region reactors discussed below, — IQ/ Carrison, D. A. and Alexander, L. G., "Nuclear Characteristics of Fused I1-Be Fluoride Reactors", ORNL-CF-57-1-141 (1957), Secret - 58 - Table VIII U255 ONE REGION, MOLTEN SALT, BREEDER REACTORS o Fllel Carrier Ose LB O BB SLERBTES Id.F"BeFE Fertile Material ...cccceee ThFh Nominal Power Density ..... 100 w/cc Temperature ..ccecssececnes 1200°F Computational Program ..... Eyewash Case S1%* S2 S3 Sk Core Diameter, ft 9.34 10 10 10 Nominal power, Mw 1150 1400 1400 1400 IiF, mole % 51.0 50.8 63.0 63.0 BeF,, mole % 47.8 5.0 26.0 15.0 ThF) , mole % 1.0 4.0 10.0 20.0 Fuel U-233 U-233 U-233 U-233 Results . Critical concentration, mole % UF,, 0.0323 0.243 1.02 2.16 Critical mass, kg, U-233 K 53.7 388 2010 37%0 . Specific power 21,400 3,600 700 375 Transmutation ratio ** 0.585 0.848 0.986 1.05 Mean energy of neutrons causing fission, ev ~0.1 ~n2 AJMOB mlol‘L ¥ 1l-inch reactor vessel of Ni-Mo alloy *¥* Combined breeding and conversion it is estimated that if U255 were substituted for U233 in these reactors, the critical masses would be increased by a factor of 2 or 3 and the conversion ratios would decrease markedly for s given thorium concentration. The use of reflectors should improve the performance of the one region - reactors somewhat, particularly in smaller sizes. This matter hag riot been explored - 59 - fully because it was felt that a reflector containing fertile material (a blanket) would in evéry case give better neutron economy for a given fuel inventory than a non-fertile reflector. As expected the mean energy of the neutrons causing fission (bottom line, Table VIII) increased sharply with thorium loading, and a reasonably fast spectrum (th ev) was obtained in Case Sk. This hardvspectrum was obtained, however, at the cost of a very large critical mass of U233. The critical masses of U235 or Pu would be even larger. However, these results do offer some hope that a truly fast, homogeneous, molten fluoride reactor based on the plutonium cycle can be achieved. 5. Two Region Reactors Emphasis has been placed on the study of two region rather than o:o region reactors for severai reasons. First, higher specific powers, and thus lower inventory costs at a given breeding ratio, can be obtained with smaller reactors. Second, in single region reactors there is no convenient way of separating the U255 or protactinium from the U235 énd U258. Third, at power densities of 250 w/cc or below, all the power that can be utilized by one turbogenerator can be obtained from 4 to 6-foot cores. To save neutrons, one places a blanket of fertile material around the core. | Preliminary calculations were performed by means of the Eyewash pro- gram using the nuclear cross sections then available on the Eyewash "sigma” tape. It was felt that the calculations would at least disclose the area where the most favorable combination of specifications would likely be found. Spherical reactors having core-diameters of 4, 5 and 6 feet were studied. The basic fuel carrier was a mixture of IiF and BeF2 with zero (four cases) and 4 mole percent ThF) . The blanket (28 inches thick) consisted of the - 60 - compound Ii_ThF._, and the core vessel was 1/3-inch of INOR-8. The core and 57T lanket densities were computed for a temperature of lQOOOF, but the nearest thermal neutron temperature available in the Eyewash program was 12800F. Cases 21, 22 and 23 of Table IX comprise a series in which the effect of varying the core radius on conversion ratio and critical mass was investigated 235 for reactors employing U as fuel and having thorium in the blanket only. The conversion ratios (line 17) are of the order of 0.6 and are relatively insensi- tive to changes in core size. The critical mass (line 14) increases sharply with decreasing core diameter in the range of 6 to 4 feet, as does the mean energy of the neutrons causing fission (line 35). This hardening of the spec- trum results in a decrease in parasitic absorptions in Ii, Be and F (lines 28 and 29) from about 0.15 to 0.08 neutrons, and a decrease in the absorptions in the core vessel (line 31) from 0.08 to 0.03 neutrons. However, this decrease 235 is offset in great part by an increase in the parasitic captures in U from 0.10 to 0.16, i.e., the mean value of 1 (line 34) decreased from2.0to 1.80 (n at 0.08 ev = 2.08). 2355 In the Eyewash program, mn of U is taken uniformly as 2.28 (v = 2.52) in the energy range from 0.025 ev to 0.18 mev. The result of using this 7 is 233 shown in Case 25, which is similar to Case 23, except that the fuel is U and the Ni-Mo core vessel is 1 inch instead of 1/3-inch thick. The breeding ratio (line 16) is only 0.53%0; however, it is observed that the absorptions in the core vessel amounted to 0.206 neutrons. For a comparison with Case 22, the excess absorption (0.206 - 0.056 = 0.150) was prorated among the blanket com- ponents, and a breeding ratio of 0.78 was estimated for a 1/3-inch core vessel thickness on the basis of the increased absorptions in the thorium. The critical 233 mass for Case 25 was gratifying low, amounting to only ~ 13 kg of U . - 61 - The principal avoidable losses of neutrons in Case 23 were those due to absorptions in fuel carrier and core vessel, the sum amounting to about 0.15 neutrons. It was conjectured that these losses could be decreased by adding thorium to the core. Case 26 shows the result of adding ~/4 mole percent ThFh to the core of the reactor of Case.22. The thermal flux was entirely suppressed (1ine 36). There was a substantial increase in the absorptions in thorium, over half of these taking place in the core, and a corresponding increase in the con- version ratio (line 16). Absorptions in carrier salt were substantially reduced and the absorptions in the core vessel were sharply reduced from v 0.06 to~s0.01 neutrons. This last effect was due in part to the increased transparency of the core vessel at higher neutron energies, and in part to the fact that fewer neu- trons leaked from the core in Case 26 (0.210 neutrons vs. 0.385 in Case 22). It was noticed that n (line 34) for Case 26 was slightly greater than for Case 23. even though the mean energy of the neutrons causing fissions was higher in Case 26. This circumstance results from the fact that n for U235 passes through a minimum with increasing neutron energy. There are a number of small discrepancies between Cases 22 and 26 which resulted from the premature termination of the calculation in Case 26 and a print- out for a subcritical condition. This shows up principally in the neutrons absorbed in fission reactions (line 24), where the absorptions in Case 26 corre- spond to a v of 2.56. The breeding ratios and the critical masses for all cases were corrected by linear extrapolations for these departures from criticality, but the neutron balances were not corrected. Case 27 shows the édvantage of using thorium in the core with U233 as the fuel, provided the capture cross sections of U253 in the epithermal range are as low as those used in the Eyewash program. Parasitic absorptions are re- duced to a very low value and the breeding ratio (line 16) is quite favorable; - 62 - Table IX EYEWASH STUDIES OF TWO REGION MOLTEN SALT REACTORS Fertile Carrier .c.oeececeess LiF-BeFo Core Vessel ..... 1/3-inch Ni-Mo alloy Fertile Material cccce. ss00ce THF) Reactor Shell ... l-inch Ni-Mo alloy Nominal Power Density ....... 200 w/cc Temperature ..... 1280°F Iine Case 21 22 23 25% 26 27 1 Core 2 Diameter, ft 6 5 Y 5 5 5 3 Nominal power, Mw 638 268 188 268 268 368 4 IiF, mole % 65.4 65.3 64.6 65 .4 50.6 52,8 5 BeFp, mole % 34,5 34,3 34,7 34,5 b, 3 Lh.7 6 ThFY, mole % - - -- -- 3.9 3.9 7 Fuel U-235 U-235 U-235 U-233% U-235 U-233% 8 Blanket 9 Thickness, inches 28 28 28 28 28 28 10 IiF, mole % 4.5 4.5 4.5 4.5 4.5 4.5 11 ThF, mole % 25.5 25.5 25.5 25.5 25.5 25.5 12 Results 13 Crit. Conc., mole % 0.063% 0,160 0.728 0,056 1.30 0,397 1% Crit. Mass, kg 7.7 4.2 95.6 12.7 303 95.4 15 Sp. Power, kw/kg 11,000 4,300 950 14,000 580 1,850 16 Transmutation Ratio¥* 0.550 0.610 0.618 0.530% 0.734 1.13 17 Core Loading, tons 18 UFY, 0.05 0.07 0.1% 0.02 0.4k 0,139 19 ThF), -- -- - - 1.36 1.36 20 BeFo 3.50 2.0k 1.01 2.05 2,38 2.40 21 IiF 3,60 2.01 1.0k 2.03 1.50 1.52 22 Total 7.15 h,12 2.19 4.10 5.68 5.42 23 Neutron Balance 2L U-fissions 0.397 0.398 0.397 0.393 0.390 0.392 25 U-captures 0.101 0.123 0.160 0.039 0.141 0.037 26 Th in core - -- -- —- 0.211 0,276 27 Th in blanket 0.276 0.321 0.347 0.230 0.193 0.218 28 F 0.078 o 102) 0.075 0.069 0.026 0.030 29 11 and Be 0.068 *T) 0 0.063 0.027 0,029 30 Total Absorptions 0.920 0.943 0.972 0.794 0.988 0.982 31 Core Vessel 0.080 0.056 0.029 0.206% 0.012 0.017 32 Leakage from blanket 0,000 0.000 0.000 0.000 0.000 0.000 33 Balance 1.000 0.999 1.000 1.000 1.000 0.999 3k n 2,00 1.91 1.80 2.29 1.84 2.30 35 E}, mean energy of ~ 0,09 ~n0.6 ~ 50 ~0.12 ~200 ~s150 neutrons causing fission 36 Thermal Flux 39x100° 1x10%7 0.6x10%7 28x10%7 o ~0 * l-inch Ni-Mo core vessel. For l/B-inch shell, T.R. was estimated to be ~ 0.78. **% Combined breeding and conversion * - 63 - however, no account was taken of fission product poisoning or captures in Pa, uranium isotopes, 116, etc. Thus refinements in the calculations can be expected only to reduce the estimate of the breeding ratio. Compared to Case 25, Case 27 exhibits a rather large increase in critical mass. The optimum concentration of thorium in the core can only be determined by an analysis of the sum of fuel costs and inventory charges as a function of thorium concentration. Detailed neutron balances for Cases 22, 26 and 27 are shown i: Tables X, XI, and XII. The thermal and epithermal absorptions in the various components of the core and the blanket and absorptions in the core vesscl are given. From an examination of the flux plots, it was estimated that a blanket only 12 inches thick wofild have captured substantially all of the neutrons leak- ing from the core vessel. The results listed in Table IX are, of course, quite optimistic and are of the nature of an upper limit on performance. rUse of corrected cross sections and inclusion of the various poisons is expected to reduce the trans- mutation ratios sharply and to increase the critical masses stbstantially. However, 1t is felt that thesg systems show promise of being able to produce povwer at an attractively low cost, mainly through reduction of fuel costs, and it was decided to refine the calculations and to extend the investigation. Accordingly, new and additional cross sections were computed and the program itself was modified in several respects ll/ and renamed Ocusol-A. 71/ Alexander, L. G., Roberts, J. T. and Carrison, D. A., "The Univac Program Ocusol-~A, A Modification of Eyewash", ORNL-CF- (in preparation) - 64 - Table X . NEUTRON- BALANCE Case No. 22 . Epithermal Thermal Core Blanket Core Blanket Total 25-Fissions 0.2617 —— 0.1365 ——- 0.3982 25-Captures 0.0953 - 0.0273 - 0.1226 Thorium -—-- 0.3019 --- 0.0187 0.3206 Fluorine ) ) 0.0041 0.0003 Be 0.0848) 0.0084) 0.0005 -—- 0.1019) Td ) ) 0.00%3 ).0002 ) Total Absorptions 0.4418 0.3103 0.1720 0,0192 0.9433 Absorptions in 0.0435 0.0128 0.0562 core vessel Ieakage from Blanket 0.0003% 0.0000 0.0003% Balance 0.9998 Table XI. NEUTRON BALANCE " Case No. 26 Epithermal Thermal Core Blanket Core Blanket Total 25-Fissions 0.3897 ——— 0.0002 - 0.3899 25-Captures 0.1408 ——— 0.0000 _—— 0.,1408 Thorium 0,.2105 0.1923 0.,0000 0.0015 0.4043 Fluorine 0.0195 0.0064 00,0000 0.0000 0.0259 Ii, Be 0,0268 0,0000 0,0000 0,0000 0.0268 Total Absorptions 0.7873 0.1987 0.0002 00,0015 0.9877 Absorptions in 0.0122 0.0000 0.0122 core vessel Leakage from blanket 0.0007 0.0004 0.0011 Balance 1.0010 ; - 65 - Table XII NEUTRON BALANCE Case No. 27 Epithermal . Thermal Core Blanket Core Blanket Total 23.Fissions 0.387 ——- 0.006 - 0.392 23.Captures 0.0%6 - 0.0006 —_— . 0,037 Thorium 0.276 0.217 0.0008 0.001 0.495 Fluorine 0.022 0.008 0.000 0.000 0.030 11, Be 0.029 0.000 0,000 0,000 0.029 Total Absorptions 0.751 0.225 . 0.007 0.001 0,984 Absorptions in 0.017 0.000 0,017 core vessel | Leakage from Blanket : 0.000 Balance 1.001 Five reactor cases have heen computed by the Ocusol-A program to date. The results have not heen completely analyzed; however, the critical loadings, breeding ratios, and simplified neutron balances are given in Table XIII. | These reactors have cores 6 feet in diameter. For a total power of 600 Mw, the required average power density in the core is 187 w/cca Thé tempera.- ThF_ as before; and the core 377 vessgel is l/5~inch of Ni-Mo alloy. The core fluid consists of 69 mole percent ture is llSOoF. The blanket is composed of Ii IiF and 31 mole percent BeFe, indicated. The lithium contains 0.0l percent 116. The thorium cross sections together with additions of ThFh and UFh as were modified appropriately to account for saturation of the resonances and Doppler broadening. The U255 was assumed to be contaminated with T mole per- cent U258. - 66 - Case 38 has a thorium concentration in the core of 1.0 mole percent and is fueled with U255. Token amounts (10 ppm) of °> and fission fragments were added in order to determine the mean effective cross sections of these materials in the neutron energy spectrum pertinent to each case. The critical mass was found to be 273 kg of U235, and for an assumed ratio of total fuel volume to core volume of 4, the inventory would be 1100 kg. The conversion ratio is 0.610. The effects of build-up of fission fragments, Uejh, U236, and U238 on this reactor have not been investigated. This reactor is reason- ably attractive, but the inventory is large. A reactor having a critical mass of 125 kg of U235 (500 kg inventory) was then selected for study. The clean reactor (Case 44) was found to be critical at a concentration of thorium in the core of 0.284 mole percent. The conversion ratio was 0.614. This improvement, in the face of decreased thorium loading, was surprising, and was ascribed to the fact that the second reactor was appreciably more thermal than the first and, consequently, profited from an improvement in the mean n for U235° By including token amounts of fission products and U233, data were obtained from which the rate of build-up of 0233 3 U23h, U256 » Tission fragments, and protactinium could be estimated. The concentrqtions of various poisons corresponding to about 2UO days of operation at 600 Mw (30 percent fission burn-up of original charge of 500 kg U235) were computed, and the thorium con- centration required for criticality was determined (Case 418)., It was found necessary to remove nearly all the thorium from the core to maintain criticality; the transmutation ratio decreased to 0.567. Note that in this period nearly 6 kg of U253 accumulated in the core, while about 50 kg was produced in the blanket (some present as Pa). The U253 in the blanket system was not considered to con- tribute to the reactivity of the core. - 67 - Since it would have been necessary to remove nearly all of the thorium from the core during the first 240 days of operation, it was clear that the build- up of U235 had been overestimated. On the other hand, there was still some U233 available from the decay of Pa present in the core. In order to avoid repeating the calculation, it was assumed that the decay of the Pa would approximately com- pensate in the next 240 days of operation for the overestimate of the U255 made for the first 240 days. All thorium was removed from the cdre and the 6 kg of U235 was allowed to burn out. The concentrations of fission fragments, Ugjh, etc., corresponding to another 240 days of operation, were computed, and the critical mass of U255 was found to be 154 kg (Case 50). The breeding ratio decreased to 0.480, and about 100 kg of Pa and U235 had accumulated in the blanket system. At some as yet undetermined point, it would become profitable to begin processing the blanket for the recovery of U253 and thé core for the removal of fission fragments. The processing cycle chosen consists of'recovéring the uraniumfi isotopes by the fluoride volatility process and, in the case of the core fluid, throwing away the contaminated salt. The steady state concentrations of the various materials corresponding to a processing rate of 600 ftj/yr (compared to a total fuel volume of about U450 ft5) were estimated. The system was then tested for eriticality. On the basis of the results, the estimates were corrected; the results are listed under Case 49 in Table XIII. The transmutation ratio for this case was found to be 0.56. The U--” feed rate is approximately 0.5 g/Mw-day of heat produced. The concentration of fission fragments in the fuel will, of course, vary with the rate of processing the core. The transmutation ratio will vary correspondingly, and the eritical loading will change. The conversion ratio can _ 68 - Table XIII OCUSOL-A STUDIES OF MOLTEN SALT REACTORS Case 38 by 48 50 k9 Fission burn-up, % of 0 0 30 60 0 initial inventory Thorium in core, mole % 1.0 0.284 0.010 0.000 0,000 Critical mass kg of U-235 273 125 125 154 99 kg of U-233 0 0 5.9 1.0 37 Neutron balance U-235 0.573% 0.549 0.493 0.540 0.281 U-233 -- - 0.052 0,008 0.23) * U-234 -- -- neg. neg. 0.030 U-236 -- -- 0.018 0.030 0.082 U-238 0.018 0.022 0.026 0.031 0.017 Fission products and -- -- 0.055 0.097 0.039 neptunium Salt, core vessel, 0.075 0.110 0.099 0.09k 0.087 and leakage Thorium in core 0.1L45 0,090 0.0%2 - -- Thorium in blanket 0,189 0.229 0.225 0.200 0.231 - Transmutation ratio (including absorptions in U-238 and U-235) Core 0.28 0.19 0.16 0.11 0.10 Blanket 0.3% 0.ho 0.41 0.37 0.h6 Total 0.61 0.61 0.57 0.48 0.56 Accumulation of U-233 in - 0 ~ 50 ~ 100 ~100 blanket, kg * Includes 0.011 absorptions in U~233 in blanket be increased by adding thorium to the core fluid. This will,of course, increase the inventory of fissionable material. The best combination of thorium loading and processing rates can be determined only by an economic analysis. These matters are discussed in Section II-G. - 69 - These reactors appear to be reasonably attractive. It is planned : to perform a parametric study of the effect of variation of core diameter and thorium loading on the transmutation ratio. Initial, transient, and steady state performance will be investigated. These studies will form the basis for a detailed, parametric study of the economics of power production in reactors of this type. L. Reactivity Effects in Typical Reactor Reactor Case U4 of Table XIII was selected for a study of reactivity effects of changes in temperature, fuel concentration, and thorium concentration. In the Ocusol-A program, it was possible to change the thermal neutron tempera- ture and to change the atom densities of the materials in the cbre, but it was not possible -~ short of modifying the program - to make changes that would . reflect the effect of the shift in the thermodynamic temperature on the Doppler broadening of the resonances in the various cross sections. The contribution of the change in thermel neutron temperature was found to be -0.3 x 10'5/°F, and the change in salt density, (1/k)(0 k/Jp)(dp/dT), was found to contribute -6.x 1077 /°F. The reactivity effects of adding U‘?35 and Th were evaluated independ- ently: (N25/k)(a k/9 N>°) = 0.4585 and (1°2/x) (0 /3 B°2) = -0.116. It follows that the ratio of the U‘?55 addition required for criticality to the thorium addition, d N°°/3 N°2, is 0.26, or one must add one atom of U->? for every 4 atoms of Th added. The reactivities determined for the U235 and Th were used to break the term (1/k)(2k/Jdp)(dp/dT) into its component parts, thus: (1/%)(2 %/ 3 0)(dp/aT) = (1/K)( 2%/ N)(aN"/aT) + (1/k)(3k/IN°) : (ar°2/aT) + (1/x)(2 k/d °)(an®/aT) - 70 - where NS represents the atom density of the salt carrier. The left-hand member is -6.3 X 10-5, as mentioned above. The first two terms on the right are, -5.4 x 1077 and +1.k4 x 10_5, respectively. Thus, by difference, (1/%) (3 k/ 2 ¥°)(an®/aT) = -2.3 x 10~ /°F This reactivity effect associated with the carrier is due to an increase in fast leakage as the density of the salt decreases. On the other hand, it was found, by averaging the macroscopic transport cross sections over the flux gpectrum, that changes in U255 concentration had little effect on the leakage, and that the reactivity effect of changes in 1\125 was due almost entirely to . changes in the macroscopic fission cross section. These preliminary results seem reasonable, and indicate that control- wise the reactors under consideration are well-behaved and can undoubtedly be controlled entirely by changes in heat load and adjustment of the U255 concentration. E. Reacltor Operation, Control and Safety The master-slave relationship which exists between power demand and power production in a circulating fuel reactor has been demonstrated conclusively in both the HRE and the ARE. Once the reactor is adjusted to produce power at the design point; load adjustment to meet fluctuations in power demand is achieved automatically without actuation of control rods or control equipment of any kind. Rupture in the reactor cooling system or malfunctioning of any of its components leading to increased temperature in the reactor core, autometically reduces the nuclear reactivity to a subcritical value. The master-slave feature makes possi- ble, in & practical reactor design, complete elimination of control rods and all attendant equipment and instrumentation for imposing direct and prompt control of nuclear reactivity. Reliability, safety, and great simplification of design - 71 - is achieved. Instrumentation and control equipment is assigned a secondary role: fo function in start-up; to monitor and adjust fuel make-up to main- tain design point power; and to monitor and indicate performance of auxiliary equipment for operational or experimental purposes. This section focuses principally on the stability characteristics of molten salt reactors, and the controls and indicators necessary for start-up and adjustment of power level. 1. The Control Problem of Nuclear Power Reactors Zg/ For nuclear pover reactors, control and safety problems are directly related to the control of temperatures and the possible rapid variations there- of. Temperature excursions can be induced by a number of events. Basically, they occur whenever the power generated is not in balance with the power removed. This relationship is expressed by the following equatipn giving the time rate of change of the temperature: ar = E(P - Pc) (1) at Ig/ The kinetics of circulating fuel reactors have been studied and reported in & number of papers. Among them are: "Current Status of the Theory of Reactor Dynamics”, W. K. Ergen, ORNL-53-7-137 (1953); “Kinetics of the Circulating Fuel Nuclear Reactor”, W. K. Ergen, Phys. Rev. 25, 702 (June 1954); "Stability of Solutions of the Reactor Equations"”, John A. Nohel, ORNL report; "Some Aspects of Non-linear Reactor Dynamics", W. K. Ergen and A. M. Weinberg, Physics XX, 413 (195k4); "A Theorem on Rearrangements and Its Application to Certain Delay Differential Equations", F. H. Brownell and W. K. Ergen, Journal of Rational Mech. and Analysis, 3, 565 (195k) - 72 - Im is the mean core temperature, E is the reciprocal of the heat capacity, P 1is the power generated, and Pc is the power removed. During operation of a reactor, the balance of power can be upset either by an excursion in the power generated or by a change in power removed, or both. Variation in generation may come about by misoperation or failure of a component either in the reactor itself or in the control equipment, if there is any. Change in power demand can occur from two distinct causes: by changes in the electric load or by a breakdown of the heat transfer system between the reactor core and the load. Removal of the electric load as well as pump failure or rupture in the fluid heat transfer system results in stopped, or reduced, heat flow. Should any of these breakdowns occur, prompt and reliable control schemes must be provided in the reactor design unless, as in the molten salt reactor, there is inherent fundamental protection against these dangers. Pro- vision must be made for the removal of after-heat when there is no load. Strictly speaking, this is not, however, a reactor control problem. To see the full consequences of an unbalance between the power generated and the power removed, one must consider the relationship between the rate of change of powef generated, the temperature coefficient of reactivity @, and the mean core temperature Tho It is, d (InP) =-a T, (2) dt T where T 1is the mean neutron generation time of both prompt and delayed neutrons. For a given concentration of fuel and poison, the reactor is critical at only one value of Tm, which is taken as the zero of fhe temperature scale. Molten salt circulating fuel reactors have large and promptly acting negative temperature coefficients of reactivity, and according to Equations (1) and (2), this feature provides great stability for the mean core fuel temperature. - 73 - For example, when the reactor is at full power, a stoppage of the power removal, accomplished either by loss of electrical load or by failure of the heat trans- fer system, will have little effect despite the sharp rate of rise of temperature predicted by Equation (1), for even a slight rise in temperature will shut off the power generation, as indicated in Equation (2). A partial loss of power load does not cause more than a temporary perturbation in the mean core temperature. The only possible difficulty of automatic control of temperature per- turbation occurs at low power levels. For extremely low power densities such as one finds during start-up of the reactor, these temperature perturbations can be large. This comes about because any increase in load appears in the reactor as a decrease in the fuel temperature. Since the power level is so low, this decrease in fuel temperature increases the reactivity without at first raising the temperature appreciably. Consequently, the reactor may go on a shcrter and shorter period until the power rises sufficiently to heat up the fuel and cancel out the excess reactivity which was introduced by cooling the fuél initially. The reactor power level at which the fuel temperature rise cancels the excess reactivity will then be considerably higher than the load. The mean temperature could then continue to rise and overshoot the steady state by a wide margin. In a molten salt circulating fuel reactor, one can expect temperature stability without significant overshoot even at low power densities. A typical value for o is 5 x 1077 pk/k/°F. Based on electronic simulator methods which accurately predicted the ARE performance, E. R. Mann Ié/ has determined for the 73/ Mann, E. R., Private Communication, ORNL - T4 - heat transfer circuit, sketched in Figure 2, that a power density of 4 watts per cubic centimeter is an adequately low limit, above which the reactor will be stable against significant temperature overshoot. This means, for example, that the reference design reactor with a temperature coefficient of reactivity of -5 x lO_S/OF and starting with a power density as low as 4 watts per cubic centimeter can take up load increases at a rate equivalent to placing the reactor on a 10-second period without any appreciable perturbation of the mean fuel temperature. For a design point of 187 w/cc, a reactor of this design can therefore be brought from a power output of about 13 megawatts to 600 megawatts in 43 seconds without serious perturbation in the mean fuel temperature by merely increasing the load demand on the reactor. Thus the reactor will re- spand completely, safely, and automatically to changes in load demand above 15 megawatts without significant changes in mean core temperature and without control equipment of any kind. Of course, this automatic reactivity control occurs as a result of density changes in the core, so that an expansion chamber must be provided, and the passageways to it must be large enough to allow for expansion without pressure shock in the core. At power levels of less than 4 w/cc, the only denger that can arise is that of suddenly injecting into the core fuel which is too cold. In normal operation, this would not occur since the start-up times for turbogenerators far exceed limits below which power increases in the core are not safe. One unusual event that could cause trouble without proper safety precaution is that of stopping momentarily the fuel pumps in all of the circulating fuel circuits, cooling the fuel to a lower than normal temperature, and then restarting all bumps simultaneously. During stoppage, the reactor core would become isothermal above its critical temperature and the reactivity would drop to a low value. UNCLASSIFIED ORNL-LR-DWG 20468 REACTOR CORE FUEL SYSTEM FLINAK SYSTEM SODIUM SYSTEM BOILER o sec 3 sec > O T o~ w9 Mmoo "o noo Mmoo » L‘E'JZ L © S s 3 o =9 S5 9eZ23d o o N O o n © 0 © o~ - F U) @) It Y I o . © ™ o0 waela S < o - N D202 < © o -~ S QL = a © o0 O ’_ 3 sec FIRST HEAT EXCHANGER {0 sec TIME CONSTANT = 0.53 sec SECOND HEAT EXCHANGER TIME CONSTANT = 0.67 sec TEMPERATURE COEFFICIENT = 5 x40™ %/°F Fig. 2. Reference Design Reactor Heat Transfer Circuit Showing Simulator Constants. - 76 - The sudden insertion of cold fuel could then lead to the difficulty previously described. Accidents of this kind can be prevented by suitable interlocks. A more thorough investigation of such unlikely contingencies will be made, but no difficulties that can not be handled by proper design are anticipated. An increased load on the reactor system is met by an increased AT across the core and heat exchangers, with the mean temperature of the core remaining the same. Simulator studies demonstrate that even with sudden failure of coupling, such as might conceivably occur in one or more of ti:e multiple heat transfer systems, the transients of inlet and outlet tempera- tures occur smoothly and in intervals of time which preclude thermsal or pressure shocks. 2., Fueling As fuel is depleted and fission product poisoning builds up, the mean core temperature at which the reactor is critical sinks lower and lower. At constant power load, this depression of temperature is reflected at all roints in the heat exchange system and, in particular, the sodium return tem- perature from the boiler is depressed. This temperature must be kept abave a minimum set by the requirement that the sodium return temperature should be maintained above the melting point of the salts in the salt-to-sodium heat exchangers. This is done by increasing the concentration of fuel in the core as needed. The relation which gives the mass, AM, of fissionable material to be added to the reactor for a given increase in the steady state mean core tempera- ture, A&m, is given by the expression, M=af M Amh where, B = 6Mc -A_k M /k - 77 - Here & is the temperature coefficient of reactivity, k is the effective multi- plication constant, and ME is the critical mess. For epithermal reactors, the ratio B ranges from two to six, usually greater than four, and it can be obtained from criticality experiments or by cowuputation. The graph, Figure 3, shows the reduction in sodium return temperature during the course of time resulting from burn-up when the constant power reneration is 200 w/cc. For example, 1f the fuel inventory is 1000 pounds o U?jfi and B is four,'and if the sodium return temperature can be allowed to drop SOOF, then the reactor must be refveled at intervals no greater than every nine and one-half days. Wheu it is so fu=led, the amount of U255 that is added is 10 pounds. The mean core temperature will rise with addition of fuel. The chance for conceivably dangerous temperature transients due %o too rapid feed can be climinated completely by adopting a design which limi*ts the rate of feed. For examnle, the rate of feed can be limited by designing r:fueling equipment to inject fuel salt in properly sized pellet form only at safe minimum intervals of time. The pellets would be held under the top surface of the fuel in a heavy mesh wire screen until dissolved. The slugging >7fect, caused by tempo- rary inhomogenelty of the concentration of fuel in the e¢irculating salt, is reduced by feeding fuel in solid form due to the time required for complete melting of pellets. 3. Criticality Start-up Before adding fuel to the reactor core system, the core, blanket and all of the fluid heat transfer systems will have been heated to temperature, checked for gas leaks, filled with their respective fluids, all liquid systems checked for full operation, and the systems again checked for leaks. With no heat generation in the reactor and the pumps running, the whole system‘will be >t Na OUTLET TEMPERATURE FROM BOILER (°F) 920 880 840 80O 760 720 ¢t 680 UNCLASSIFIED ORNL —LR-DWG 20467 INVENTORY | {000 Ib TEMPERATURE COEFFICIENT o 85X 10—5/ of 7: e _ 1 | | | | \ .- POWER DENSITY = 200 wotts /em3 — - 25 30 35 40 DAYS Fig. 3. Change in Coolant Inlet Temperature for Intermediate Heat Exchanger due to Fuel Burn —-up, for a Typical Fused Salt Circulating Fuel Reactor at a Power Density of 200 wofls/cmz’. 8L - 79 - isothermal and the temperature can then be controlled in a sensitive way by steam supplied to, or removed from, the boller. With the reactor in this condition, it can be brought to eriticality by the safe procedure outlined below. Nuclear instrumentation will be required for this initial fueiing operation. With no fuel in the reactor, the counting rate from neutrons origina- ting i1n the source is accurately determined and its reciprocal value is used later in the familiar reciprocal counting rate versus fuel mass plot. With fuel pumps operating, concentrated fuel salt is added until the fuel load reachegs 80 to 90 percent of the value which makes the reactor criticsl, as previously determined from hot criticallity experiments. Then a neutron count rate is measured, after which carefully limited amounts of concentrated fuel salt are added step-wise with intervening counting and plotting of points. When the reactor is near critical, as determine? from the plot, the reactor core temperature is slowly lowered until the neutron counter indicates eriti- cality, at which point the temperature of criticality is determined. The zcoler is turned off, the temperature of the core rises to design point, and the adding- fuel, counting, determining-criticality-temperature cycle is repeated until criticality at core design point temperature is reached. Experience from the ARE has demonstrated that a well planned and deliberately executed procedure such as this 1s a simple, safe, and reliable one for achieving criticality in the reactor. Reactor constants, important for future fueling and other operations, can be determined from data recorded in the reactor start-up log. - 80 - F. Build-up of Nuclear Poisons and Chemical Processing l. Fission Product Poisoning A 600 Mw reactor will produce about 190 kg/yr of fission nroducts. Abou* 23 atom percent of the fission fragments have_a decay chain such that tey appear as an inert gas - Xe or Kr - with a half-life greater than one hour and thus are subject to removal by purging with He or N2 gas. These re- movable isotopes contribute about 25 percent of the total fission product poisoning in the 100 ev resonance region. This percentage is higher in a thermal) reactor because of the very large thermal neutron cross section of . ie‘fis, bt burn-out limits the Xe155 poisoning to a maximum of about 5 percent. I the resonance region, however, adjoining nuclei do not have great differ- ences i.. cross section, and burn-out is relatively ineffective in limiting poison. Thus, to a first approximation, poisoning increases almost linearly with time if fission products are not removed. About 26 atom percent of the fission products are rare earths. In a 100 ev resonance reactor, they contribute about 40 percent of the fission pro- duct poisoning. The remaining non-rare-ges and non-rare-earth fission products include a wide variety of elements, fio one of which is outstanding from the poisoning point of view. In a thermal U255 burner reactor operated at constent power and at constant U255 inventory, but with no fission product removal, the fission pro- duet poisoring, szth Zifth’ 135 poisocaing (0-5%) plus equilibrium Sm is approximately equal to the equilibrium Xe 149 poisoning ( ~v1.2%) plus the contribu- tion from "all other fission products”. According to Blomeke and Todd IE/, T4/ ORNL-2127 - 81 - the "all other fission product” poisoning totals about 3 percent when tb~ total amount of U255 burned is equal to the U255 inventory (100% burn-up), increases to about 19 percent at 1000 percent burn-up, and to about 51 percent at 10,000 percent burn-up. Thus it is possible, although not necessarily economical, to run a thermal, fiuid fuel reactor for many years without being forced to vrocess to remove fission product poisons., One would, of course, pay for not processing by higher inventory charges for U255 and by lower breeding-conversion ratios. A 600 Mw thermal reactor burns about 230 kg U255 per year, so that with 460 k- 102 inventory, the fission product poisoning would increase from O-5 percent initially to 20-25 percent after 20 years. Even in thermal reactors, resonance captures in fission products meke the poisoning somewhat worse than the numbers given above. The magnitude of the extra poisoning depends on the ratio of the neutron flux at resonanée energle to the thermal neutron flux, which is determined in part by the effectiveness of the moderator. In resonance reactors, the fission product poisofiing is consider - ably worse due to the higher average fission product absorption cross sections relative to U235. In a 100 ev reactor with a 530 kg U255 inventory, the total fission product poisoning 12/ would increase approximately linearly from zero percent initially to ~ 48 percent after 2 years at 600 Mw. For U255 fueled reactors, the fission product polsoning is about the same as for U255 at thermal neutron energles, but in the resonance region, the higher U255 cross section reduces the poisoning effect of the fission products by a factor of two over U235. Thus a resonance breeder-converter burning half- and-half U253 and U235 would have s fission product poison level of ~s9 percent if processed twice per 100 percent burn-up. 75/ Greebler and Hurwitz, KAPL-1L40 - 80 - 233 237 239 2. Pa and Np Poisoning . 233 , Np Neutron capture in Pa or Np239 has the same result as a non-fission capture in U233 or Pu259, i.e., a fissionable atom is effectively lost as well 257 as a neutron. Although neutron loss to Np does not involve loss of a fission- able atom, the total poison can be worse when U255 is used as make-up fuel. Although neutron capture by any of the three yields a fertile atom, at present relative prices for fertile and fissionable materials, the gain is negligible compared to the loss. 235 The average ratio of neutron captures to B-decays by Pa in a reactor is given to a good approximation by, — Pa 0.046 P (1 + Q) [B'R'é[ca Th - Th M 9y where, P = reactor power level, thermal Mw (1 + @) = ratio of absorptions to 7tssions in fuel : B.R. = breeding or conversion ratio ' M?h = kg of thorium in system . The P and « refer to the whole system. The other parameters can refer either to the whole system or to the core and blanket separately. 1In molten 233 salt power reactor studies to date, neutron captures in Pa have been negli- glble, due primarily to the large thorium inventories considered. In a U233-U235 breeder-converter using highly enriched UE35 make-up, 239 Np poisoning is relatively unimportant, but if the breeding-conversion ratio 2 5T poisoning can become objectionably high if it is allowed to is poor, Np reach its equilibrium value with no removal by chemical processing. This is éspecially true in resonance reactors, in which U236 (and hence Np257 at equilibrium) ylelds may be twice as high as in thermal reactors. Chemical processing to remove Np is discussed in a following section. * - 83 - 3. Corrosion Product Poisoning Corrosion product concentrations in molten fluoride salts in INOR-8 loops at 15000F have not been observed to exceed 100-300 ppm for Fe, Cr and Al, or 100 ppm for Ni, Mo and other alloy constituents (with ~ 20 ppm being typical for Ni and Mo). At lower temperatures, these numbers are smaller, but precise values at 1100—1200°F are not yet available. These corrosion products will be removed from the core system, along with the fission products, by the chenical processing system. They will build up to equilibrium in the blanke: system (where they are relatively less objectionable than in the core since blanket poisons have to compete with such a very high concentration of thorium). The Fe, Cr, Ni and Al are relatively light elements and thus should have lower capture cross sections than typical fission products. Molybdenum is typicai of the lighter group of fission products (about 18 percent of fissione z-entually yield a stable Mo isotope). Since the chemical process- ing will probably be at such a rate that the steady state fission product concentration in core salt will be several thousand ppm, the corrosion product poisoning hias been neglected. 4. Chemical Processing and Fuel Reconstitution The "ideal™ reactor chemical processing scheme would remove £ission Pa255 239 as soon as %“hey were formed. After products, corrosion products, and Np =3 239 . the latter two had decayed to and Pu ”7, it would then return them %o the reactor, along with the U and Pu passing through the process, in the desired form. This ideal chemical plant would have low capital and operating costs, would hold up only small amounts of fissionable and other high-priced materials, and would discharge its waste streams in forms that could be inexpensively disposed of or sold for a profit. Present technology, however, does not offer such an ideal process for any reactor. - 84 - More practical short-term goals for processing a molten salt reactor might be (a) continuous removal of most of the gaseous fission products by purg- ing with §e or N, gas, (b) an "in-line" removal of rare earth, noble metal or other fission products by "freezing out" of part of the salt stream, "plating out” of fission products on metallic surface either naturally or electrolyti- cally, “"salting out", e.g., of rare earths by keeping the salt saturated with Ce, or "slagging out" of & solid carrier phase by adding oxygen or oxides, and (¢) continuous or batch removal of salt fuel from the reactor at an economically optimum rate to separate the U, Pu and salt from the remaining fission products and corrosion products by the least expensive method available. Present tech- nology does not offer all of this for a molten salt reactor which has to be designed now, although there is reason to expect that an adequate development program would enable it to be approached more closely in the "second reactor” design. Operaticn of the ARE and of ANP in-pile loops indicates that gaseous fission product removal can be achieved and that Ru, Rh, and Pd plate out on metal surfaces. The fluoride volatility process achieves one of the most impor- tant goals in separating U from salt and fission products. Scouting experiments in ORNL Chemistry and Chemical TECHnology laboratories have indicated a basis for optimism that further development will yield methods of separating salt from fission products. Adequate technology already exists for the preparation of fresh fuel starting with non-radioactive UF6 and fluoride salts, but further development is required to demonstrate "hot" (and hence remotely operated) reconstitution of fuel from recycled uranium (and salt if possible) if long cooling times are to be avoided. Fluoride Volatility Process - Processes for the decontamination of uranium by fluorination to produce volatile UF6 are under active development at the Argonne and Ozk Ridge National laboratories. Both sites have studied - 85 - the dissolution of solid fuel elements (e.g., STR) in molten fluoride salts as a preparatory step for fluorination, and ORNL has also studied the fluori- nation of molten salt reactor (ARE and ART) fuels as well. The ANL process fluorinates with BrF_, and distills the UFE product to complete the decontami- 5 nation. The ORNL process fluorinates with F, and completes the UF6 decontami - 2 nation with a sorption-desorpti«: cvecie on a NaF pellet bed. The ORM'. volatility rosram is currently running at approximately a Sl,OO0,000-a-year level and is a% an =ariy pilot plant stage. The pilct plant is now being broken in with nor-radioactive feed. This cold stage wili be followed by processing of warm (long-cooled) ARE fuel and then hot STR (10-30 day cocled) fuel, over fiscal 1958 and fiscal 1959 according to present plans. The chemistry of the process has been described by Cathérs Zé/, and the pilot plant by Milford ZZ/. Figure 4 is a biock flow sheet of the proc=ss. For pro- cessing ARE type fuel, the uranium bearing molten salt is transferred to the fluorinator in 1.k ft5 batches. Fluorine, diluted with N2, is bubbled through the salt at 600-650°C until its U content is ~ 10 ppm, and UFg, N, and excess Fé pass out of the fluorinator through a 100°C NaF pellet bed (capacity of 10 kg U) wiiich removes the UF6 from the gas stream. At present, the excess Fé is scrubbed out of the gas stream with a reducing KOH solution. In vola- tility laboratory studies, F, recirculation with a K-25 B-4 pump has been used and may be installed later in the pilo* plant, thus reducing F2 consump- tion and simplifying the disposal problem. The UF6 is desorbed from the NaF by raising the temperature to 400°C and sweeping with F_-N 5~Nye The UFB passes 76/ ORNL-CF-56-9-21, American Nuclear Society Meeting, December 1956 77/ ORNL-CF-57-%-18, American Chemical Society Meeting, April 1957 N(JF-ZrF4 Zr-U FUEL ELEMENTS UNCLASSIFIED ORNL-LR-DWG 19090 F, DISPOSAL = " Ru | F> — COLD | (DESORPTION) L TRAP | UFg ANHYDROUS F, XA coLD |- HF TRAP X . Hy + HF NaF NaF BED BED Y XB UFg UFg +F5 PRODUCT s——— HYDRO- | NoF- 7, -UF, FUSED SALT FLUORINATION | FLUORINATION 650°C 600-650°C — e ABSORBE RS WASTE NaF-ZrF Y WASTE NaoF Fig.4. Fused Salt-Fluoride Volatility Uranium Recovery Process & - 87 - through & second NaF bed and is then enllected in cold traps at -40 to -6000. The plant product is liquid UFB, or;tained by isolating the cold traps from the rest of the system, heating to above the triple point and draining into the product receiver. Most of the decontamination is achieved in the fluorination since mocs® of the fission and corrosion products remain in the salt. The volatile contani - nants (%s, Kr, I, Te, and Mo, most of the Ku, »nd part of the Nb and Zr) ei“her pass through the NeF bed while the UFy is retained (Xe, ur, I, Te, Mo, Ru) or remain ou the bed when the UFg is desorbed (Mo, Zr). The I, Te, Mo, and Ru are most.iy removed by a :cid trap, the remainder being scrubbed out of the gas system aicng with the Fé. The Xe and Kr follow the NQ to the plant off-gas system. The Nb and Zr slowly build up on the laF bed, which is replaced when poisoned. laboratory development indicates tha% a lOOOC micrometallic nickel filter betwszen fluorinator ard NaF bed removes most of the Nb and Zr, and this, too, may be added to the pilot plant. This addition would greatly extend the life of the NaF bed. According to ORNL =xperience, only part »f the Pu follows the U. The behavior of lip and Pa has not beeu ztudied. A molten salt power reactor develop- ment program should include studies of these three elements. Volatility processing of LiF-BeF2 and Na.F-BeFEj salts has not been demonstrated, since the present process is based on NaF-Zth salts, but no difficulties are expected. 1Iu fact, the Ii salt complexes UFh-UF -UF6 less strongly than Na salts and hence should require less 5 excess F.. 2 Reduction of UFE to UFh - The continuous reduction of highly enriched UFE to UFL with hydrogen is well proved as & non-radioactive process. The process - 88 - developed and used at K-25 Ié/ is indicated in Figure 5. The reduction takes s place in a UF|6~F2~H2 flame in a Y-shaped tower reactor. The F2 is added to give a hotter flame. The reaction products are UFu powder and HF-H2 gas. Micrometallic filters recover any UFh which is entrained in the exit gas. A vibrator is used to shake free any UFL vhich clings to the filter or tower walls. A chemical trap using a CaSOh or NaF pellet bed recovers any unreacted UF6 in the exit gases; although the amount so collected is negligibly small in normal operation. The exit gases are scrubbed free of HF with either a KOH solution spray or a NaF bed. This process has Dot been operated at a high level of radioactivity. The UFB from the power reactor volatility process would be somewhat radiocactive, with the major activity probably T, 4 "hgt" pilot plant demonstration should be provided in a fused salt power reactor development program. No unusual diffi- culties are anticipated, however, since the present continuous process is smooth- : running and practically automatic. For molten salt reactors using sodium rather than lithium based fuels, the UFB to UF& recycle may well be greatly simplified by reducing the UF6 with H2 on the NaF bed and using the NaF-UFu pellets for fuel make-up. 5. Build-up of Even-Mass-Number Uranium Isotopes The build-up of U2, 1PoF, 1P, and 1°Pas non-fissionable isotopic diluents in U233 and 0235 plays an important part in fuel cycle economics. Although U252 does not build up enough to affect the neutron balance signifi- cantly, its hard-gamma-emitting daughters grow fast enough to be a biological hazard in the handling of U233, and thus adversely affect the resale value of 78/ Smiley and Brater, TID-7518, Part 2, p. 156-210 SURGE DRUM C.V. i-l E - —D4 ORIFICE a o '.— <{ H, CYLINDERS g:) > Fz'——%—"{_—__“‘—*“—-%—’ .@1_ SINTERED METAL UFs DRAIN FILTERS CYLINDERS UFG - UF4 REACTOR r—(%-——:,l | » 1'% > UFs VAPORIZERS = POWDER REMOVAL SYSTEM i ] e r -~ ——— FLEXIBLE CONNECTION UFs REDUCTION PROCESS FLOW SHEET FIGURE 5 [ CHEMICAL - 9 - the U233o It has been assumed that the molten salt power reactor will process and burn all the U23§ it produces, hence the U252 problem has not been con- sidered in detail. 233 233 Radiative captures in Pa and U lead to isotopic contamination of the U233 with U23h. With no processing to separate these isotopes, and none seems feasible, the U23h builds up until it is being produced and burned at the same rate. At equilibrium, in thermal reactors, there is ~/ 57 percent as much U23h as U253, with U25h capturing ~» 9 percent as many neutrons as U235o In fused salt resonance reactors with higher capture to fission ratios, at equili- 234 233 brium, U and is present in amount equal to ~35 percent of the U255, Nevtron capture in U23h produces captures ~ 13 percent as many neutrons as U fisgionable U255, but a capture in thorium would be preferabl: since U235 is a better fuel than U255° Neutron capture in U256 results in an isotopic poison,; since U257, with a 6.75 day half-1ife, decays to Np257 too fast to permit useful amounts of the fissionable U237 to build up. In completely thermal reacors fiith no 36 builds up until it is ~ 18 times as abundant as U235,' isotopic processing, Hg and captures ~ 16 percent as many neutrons. Normally, in any real thermal reactor, resonance capture by U236 will reduce the steady state ratio to less than 18. Isotopic separation «f U235 and U256 may be feasible because of the large amounts involved an: because it is important in a breeder-converter economy. On its own merits in a separate cascade, it should cost at least 9 times as much as separation of U235 from U258, but by feeding it into exist- ing cascades, either by adding top stages or accepting lower production rates, less expensive processing can probably be achieved. The K-2§ Operations Analysis « division is studying the gaseous diffusion problem, and ORNL is studying the - 91 - over-all problem. At present, AEC buy-back prices for U235 do not penalize 236 any more than they do dilution with U258° isotopic dilution with U23u and U Despite this present buy-back policy, this molten salt reactor study has assumed tha*t, in the 1dng run, equilibrium U256 polsoninz would simply have to be toleratcd. This is pessimistic if the U235 is kept separate from the U235 in a two region converter, or if a resale market to the military or to other reactors is available at prices near the present AEC buy-back level. For a steady state reactor using highly enriched U255 feed (93% 25, 6% 28, 1% 24) and with no iscvtopic processing, the U238 at equilibrium will capture 6/93 as many neutrons as the U fed. In a "e-r:.pletely thermal” reactor, at equilibrium there would be 16 times as much U238 as U255° In fused sait resorance reactors, the U238 would build up to only about 10 percent ~ ¢ the U235 fed. Thus at isotopic equilibrium, U256 is a worse contaminant than U238e Neutron cAapture in U258 produces fissionable Pu239° For this study, it has been assumed that equilibrium concentrations of and captures in U238 235 chain is kept separate from the U233 would have to be accepted. If the U chain, this is a more pessimistic assumption thar is usually made. Even with mixed chains, it is a more pessimistic assumption than is often made. For example, present buy-back prices pefmit the Consolidated Edison reactor to use a mixed U255, thorium fuel and sell the resulting mixture of U«~232-233-234-235- 236-238 for $15 per gram of the amount of U797 and U°2° that it contains. 6. Radioactive Waste Disposal At 600 Mw of heat for TO00 hrs/yr, a reactor produces ~ 190 kg/yr of fission products. In molten salt reactors, perhaps 24 weight percent of the fission products can be removed as Xe-Kr gases, leaving 145 kg to be removed by chemical processing. The proposed chemical processing flow sheet waste streams - 92 - include fused salt, NaF pellets, F2-N2 and HF-H2 gases. Most of the non-gaseous fission products remain in the core salt residue and may be stored in this form. Most of the remaining fission products are removed by periodically flushing out the micrometallic filter between the fluorinator and the NaF bed and the cold trap between the NaF bed and the Fé disposal unit. The NaF bed 1s replaced occasionally when it becomes poisoned with niobium and zirconium. Any remain- ing fission products in the gas streams are serubbed out with the F2 and HF or vented to the reactor off-gas system. For optimum costs, high power molten salt reactors shoul. probably have relatively large inventories of enriched uranium and be processed sbout twice per fuel-inventory burn-up, a compromise between the cost of salt replace- ment and the savings due to improvgd breeding-converzion ratio. The core salt discard rate indicated in the follewing section is 500-1000 ft3/yr. This volume corresponds to a farily high fission product concentration, comparing favorably with any other type of power reactor processing. The disposal problem is not " a small one, however. The high concentration of fission products makes the heat dissipation a problem. The high value which the salt would have if the fission products were removed ( SleO/ft3 for L17Be, perhaps half that much for Na-Be), makes it desirable to store it in an easlly recoverable form until means of re- claiming it economically can be developed. Underground storage tanks cooled by natural circulatioi of air, with the salt kept molten, appear to he an acceptable answer. These might well be built one at a time, each with a one- year storage capacity, until a final decision on ultimate disposal is made. ..95.. G. Fuel Cycle Economics 1. Cost Bases Fissionable isotopes have been valued at $17/g and inventory charges calculated at 4%/yr. Thorium cost in fluoride salts has been taken as $30/kg, an arithmetic average of published prices of $17/kg as nitrate and g43/kg as metal. A price of $91/kg for 99.99 percent 117 in large quantities, an authori- tative estimate from Y-12, has been assumed. A fluoride salt cost of #5/1b, plus the cost of the special materials uranium, thorium, lithium-7, has been used 18/ . This leads to $l1l.1-/1b for 69 IiF-31 BeF, and 315.4%0/1b for 75 LiF- 25 ThFh. About one-ialf of the latter is for thorium. The original salt inventories have been capitalized with a 16 percent annual charge corresponding to a 20-year depreciation period. It is assumed that new core salt will be purchased to replace the amounts rrocessed. The blanket salt is used over the 20-year reactor life without excessive fission product poisoning build-up. Fissionable material consumption cost is based on feeding 93 percent enriched U235 to the core system to compensate for a breeding conversion ratio of less than unity. It is assumed that U233 is not available for purchase at an economic price, although it would be worth more than U235 in a molten salt resctor due to its better nuclear characteristics. It is also assumed that 53 isotopic re-enrichment of U235 or U2 or resale of even isotopic contaminated uranium is not economical and, therefore, that the molten salt power reactor must burn out the non-fissionable isotopes and pay for it in lower breeding- conversion ratio. 79/ Private estimate of W. R. Grimes, ORNL - 94 - 2. "Steady State" Neutron Balances and Comparative Fuel Costs The neutron balances given in this section are not the actual ones given by the Univac; but; rather, modifications or corrections of the original balances as required to make the reactors both "eritical" and "steady state". The Univac results were used to obtain flux-averaged microscopic cross sections for the various elements, and these were assumed to remain constant as con- centrations of the individual absorbers were changed provided the total macro- scopic absorption cross section was held constant. The results are believed to be fairly good for comparison purposes, but no better on an absolute basis than the basic cross sections themselves. The fuel costs compared below include only fissionable material rental, fissionable material purchased for burn-up, and core salt purchased to replace that processed. These are the "rapidly variable" fuel cycle costs, i.e., those which vary sharply in "reference design" type.reactors (Section III), depending on choice of operating conditions. Although the chemical plant capital and operating costs would also vary to some extent with processing rate, their varia- tlon is not fast enough to affect the optimization appreciably. Three "different” reactor cases are compared below, five versions of a "reference" case and one version each of "low inventory" and "high inventory" cases. The first two cases are for 6-foot diameter (113 ft5) cores with 337 ft5 external holdup. The last case is for a 5-foot diameter core (65 ftB) with the same external holdup. The values of v were taken from BNL~325 to be: U-23%3: 2.54 U«235: 2,46 Flux-averaged a's and u's were chosen separately for each case from the Univac calculations on which they were based. o -95_ Reference Case - The columns headed A, B, C, and D in Table XIV describe steady state reactors based on Ocusol-A Case 49 for various chemical processing rates and using 9% percent U255 make-up. The reactor of Column E is like that of D except that U235 make-up is assumed. The flux-averaged values of o for these reactors were: U-233: 0.15 U=-235: O.41 The flux-averaged microscopic absorption cross sections were in the follow- ing ratios: U-233: 1.00 U-234-6-8: Q.37 FP: 0.22 Th (blanket): 0.036 Th {core): 0.125 Column A gives the fissionable material inventory, neutron balances and fuel costs for an "infinite processing rate", i.e., no uranium or fission products in blanket and no fission products in core. In the other columns the blanket is assumed to be processed for uranium removal at a once-per-year rate (with an average-over-thedife-of-the-reactor amount of fission products included in the blanket) and the core processing rate is varied. Corrosion product and Np237 polsoning is not included as such, but it is felt that this is compensa- ted for by taking no credit for separate removal of gaseous products or for the production of, or fissions by, plutonium. The costs in Column E assume that U235 could be bought for $17/g. They may be interpreted to indicate that, by comparison with Column D, the fused salt reactor could pay twice as much for U233 as for U235. - % - Table XIV NEUTRON BALANCES AND FUEL COSTS FOR DIFFERENT INVENTORIES AND PROCESSING CYCIES I Column IT Processing Cycle, yrs Core Blanket IIT Inventory, kg 33, Blanket 23, Core 25, Core IV Neutron Calance 02, o2, 23, 23, 2k, 25, 2 28, FP, FP, Other Totals Effective B.R. v Fuel Costs, 3/yr 23 + 25 rental 23 or 25 make-up Salt make-up QuoaaaaaaaaQbQw A B C D E F G 0 0.25 0.50 0.75 0.75 0.54 0.97 0 1.00 1.00 1.00 1.00 1.00 1.00 ——- 105 105 105 105 100 100 269 228 202 175 318 ok 400 192 258 313 369 69 288 700 471 591 620 649 492 482 1200 0.2446 0.2312 0.2312 0.2312 0.2312 0.2033 0.1854 0.0926 0.066L4 0.0333 —— 0.1099 - 0.0959 ——— 0.0110 0.0110 0.0110 0.0110 0.0211 0.0200 - 0.3372 0.2866 0.253%5 0.2202 0.3993 0.1822 0.261h 0.0433 0.0382 00,0430 0.0297 0.0527 0.0278 0.0373 0.1457 0.1960 0.2382 0.2806 0.0527 0.3126 0.2186 0.0k426 0.05Th 0.0698 0.0821 0.0155 0.0873 0.0641 0.0072 0.0110 0.0143 0.0175 —— 0.0199 0.0127 _— 0.0024 0.0024 0.0024 0.0024 0.004kT 0.0050 _— 0.0129 0.0257 0.03%86 0.03%86 0.0405 0.0271 0.0868 0.0869 0.0866 0.0867 0.0867 0.1006 0.0725 1.0000 1.0000 1.,0000 1.0000 1.0000 1.0000 1.0000 0.82 0.73 0.64 0.56 0.83 0.50 0.68 320,000 402,000 422,000 141,000 535,000 328,000 820,000 815,000 1,260,000 1,630,000 2,030,000 774,000 2,275,000 1,450,000 ® 2,430,000 1,220,000 810,000 810,000 1,116,000 670,000 @ 4,092,000 3,272,000 3,281,000 1,919,000 3,719,000 2,940,000 -97_ Low Inventory Case -~ The above-discussed reference case had its median fission in the 34-61 ev energy group, with U233-U235 inventories of ~ 600 kg. Column F describes a reactor (based on Ocusol-A Case L4i4) with a lower uranium inventory. The medilan fission was in the 10-15 ev group. The flux-averaged a values were: U-233: 0.16 U-235: 0.39 The relative abso~:tinm: cross sections were: U=-233: 1.00 U=-23%5: 0.58 U-234-236-238: 0.31 Th (bleanket): 0.028 Th (core): 0.097 A3 Fuel make-up with i 77 55 assumed. This case does not appear to offer as economical a fuel cycle as the reference case. The biggest difference is in the increased fractios: of neutrons lost to salt, shell and leakage ("other" in the table). High Inventory Case - Since it appeared from the above two cases that an even higher inventory might be still mo¥e economical, another case was examined and is listed in Column G (based on Eyewash Cases 26 and 27, with Ocusol~A a's inserted). Altfiough this case differs from the preceding two in core size as well as inventory, it does appear that the higher inventory "pays off" in increased breeding conversion ratio and reduced salt replacement costs. - 98 - SECTION ITI Reference Design Reactor A, Introduction To allow a realistic estimate of the performancg, safety, economics and cost of construction of a typical molten salt reactor, a specific reactor type and size was chosen for detailed study. The principal items considered in making this choice were: 1. The reactor should be capable of relatively early construction. No component, material, or process is used that is not elther already avallable, or tested or under test in at least pilot plant scale equipment. The assumption is madg{ however, that equipment such as pumps and heat exchangers can be de- signed and built successfully in larger sizeg than those tested to date., Devel- opment and testing charges will be associated with this scale-up, 2. Safety in an early reactor is of paramount importance. Great emphasis is placed on avoiding any reasonable possibility of chemical accident i£volving radioactive comporents, 3. Long life of components is esgential for long-term eeconomy. On this basis the alloy INOR-8 was chosen as a material of construction; however, 1ts reliability has not been demonstrated in long term tests, although short term tests suggest that it should last for many years. L, The réactor should possess the ability to serve as a prototype for central station power generation. >. The possibility that this may be the first power reactor of its kind requires that as.much information as possible be derived from it. Thus the reactor should be as versatile as possible without incorporating experimental facilities that seriously affect the economics or continuity of power production, - 99 - While it was not expected that this "reference design reactor" would necessarily turn out to be the best one possible, it was chosen after considera- tion of a large variety of factors and should be representative of good molten salt power reactors that can be built in the near future. The reference design reactor is a two region homogeneous converter with a core approximately 6 feet in diameter and a blanket 2 feet thick, Moderation is prov;ded by the salt itself so there is no need for a moderator or other structure inside the reactor. The core, with its volume of 113 cubic feet or 3200 liters, is capable of generatin 600 megawatts of heat at a power density of only 187 watts/cc. This rate of power generation is well within safe 1limits, and the total power output of 600 thermal megawatts from the core, plus additional heat from the blanket, allows a net electric power genera- tion of approximately 240 megawatts, The basic core salt is a mixture of about 7O mole percent LiTF and 30 percent BeFe. Additions of thorium fluoride can be made if desired, an? enough UFh will be added to make it critical. The blanket will consist of the eutectic of LiF and ThFh, or mixtures of it with the basic core salt. The melting point of the core salt is 867°F and that of the blanket salt is lOBOoF, or lower, | Both the core fuel and the blanket salt will be circulated to external primary heat exchangers, six in parallel for the core and two in parallel for the blanket. The numbers of parallel circuits were chosen so that useful opera- tion could still be obtained if any one circuit should fail. It also turns out that reasonable sizgs of pumps and heat exchangers result from the 100 Mw heat removal requirement for each of the six core circuits. Flinek, a mixture of alkali fluorides (mixture 12), was chosen as the intermediate coolant because it has reasonable chemical compatibility with - 100 - the core and blanket salts and with sodium, and, of the salts, it has good heat transfer characteristics, It will become only slightly radioactive by exposure to the delayed neutrons from the core fuel and will serve as an isolating link to keep all radioactivity from contact with the steam system. The melting point of the Flinak is 850°F. Secondary heat exchangers will transfer heat from the Flinak to the sodium, which is used directly to heat the bollers, superheaters and reheaters, Details of a reasonaflle heat transfer system have been worked out, so that with a fuel temperature of lQOOOF, a steam temperature of lOQOOF can be achieved. Figure 6 gives a block diagram of the gross features of one of the heat trans- fer systems of the core circuit. The blanket cooling systems are similar. After a careful consideration of the problem of control of the reactor, it has been decided that there is nc need for any control rods. Reactor con- trol is automatically maintained by the negati#e temperature coefficient° Uranium fluoride fuel or thorium fluoride as a poison will be added when the need arises, as indicated by temperature measurements in the non-radioactive sodium system., Thus, this reactor can be properly called a nuclear furnace, requiring fuel additions to maintain temperature. The basic hardware of the system lends itself to a number of different fuels and fuel cycles, For example, the core can be operated with UEBSFh or UzBBFh, or, perhaps later, with PuF i Different inventories of uranium age 50 possible, depending on the emount of fertile material introduced, and the amount of fisslion products and heavy elements that are allowed to build up. One indicated mode of operatlion is to start up with 0235 in the core, together with a little thorium, and to process the core on such a cycle that a complete reprocessing would be carried out appro;imately every year, The processing UNCLASSIFIED ORNL-LR-DWG 19890 1100 F 12.55C 1000F 36.8C @ 1000F ___B890F 5.2C FUEL SALT T 311C < - TO FLINAK FLINAK HEAT TO SODIUM TO 1ROE61-(I)EFAT EXCHANGER HEAT X 6 ac —————— - 060 X EXCHANGER : : 1930F SUPERHEATER— | 12.61C —] H25F 1.4 C \V/ \ 1090F Na TO No BLENDER —— o] - HEAT EXCHANGER i 1200F . 940F 650F 860 F FROM REHEAT REACTOR 800 F 36C 770F 9.1C BLENDER— I 670 F 26.9¢C ec 91¢C BOILER ‘ LEGEND F = TEMPERATURE, °F 515 F C = FLOW, ft¥/sec 1.8C Fig. 6. Schematic Diagram of Heat Transfer System. 101 - 102 - would consist of removing the uranium from the core fluid as UF6 by the volatility process, storing the core fluid with its contaminating fission products for later processing, and making up fresh fuel from the reduced_UF6, with new UFh added to compensate for fuel burn-up and for heavy element build-up. U235 would build up from whatever thorium is included in the core. The 0235 extracted from the blanket by the volatility process would either be stored until enough is cbtained for a pure U233 core loading, or it could be added to the core as soon as it 1is extracted, In either case, the uranium could be recirculated indefinitely and economical operation would be achieved, even after equilibrium concentrations of U25h, U256, and U258 arz built up. | A factor in the selection of the multiple stage heat transfer system . involving two intermediate Tlulds was the desire to have only compatible fluids in adjoining volumes where leaks could involve radioactive materials. This avoids the possibility of liberation of fission products as the result of a chemical accident. As the system is now designed, the pressures developed by the pumps, by difference in density of the fluids, and by over pressure, are such that any failure producing mixing of fluids would tend to produce flow toward the reactor core, rather than from it;, tending furthgr to confine the fission products. An exothermic chemical reaction would result if the sodium and water or steam were mixed, However, this would not involve radioactive materials and would pose only the same danger problems as in any conventional plant han- dling quantities of chemically active materials whgre no biological poisons are involved., The components of the reference design reactor are relatively simple; the apparent complexity stems mostly from the number of components rather than - 103 - their type. The large number of components is required because of the large power output contemplated, The basic simplicity of the components together with the high thermal efficiency and low fuel turnover cost encourages the belief that the future cost of power from reactor plants of this type might be fairly low, B. Heat Generation, Transfer, and Conversion System l. Reactor The heat transfer system for the reference design reactor has six parallel, independent paths for heat flow from the reactor core fuel to the steam- system, and two similer paths for heat from the blanket fluid. Figure 7 shows section and plan views of the reactor vessel, gilving the general arrangement of six parallel core pumps and return lines. The reactor vessel, piping, and pump housings and impellers will be made of INOR-8. As explained in the irntroduction of this section, the multiple heat removal circuit system was selected as offering reliability and safety. The symmetrical positioning of reasonably sized components allows low fuel holdup and low thermal stresses in the reactor structure and its attendant piping. The fuel will be enriched to the extent that the average temperature of the core for criticality will be 1150°F. At the full rated power genera- tion of 600 megawatts of heat in the core, the system is designed so that the fuel salt will enter the core at 1100°F and leave it at 1200°F. With this 100°F temperature range, the total flow of fuel through the core must be 3k ,000 gallons per minute. At this rate, it will take the fuel an average of 1.5 - seconds to pass through the core and six seconds to make a camplete circuit of core plus external system, Each of the six parallel core heat removal cir- cuits will then handle 5,650 gallons per minyte of fuel, over a temperature range of lOOOF, and transfer 100 megawatts of heat, - 10k - The multiple circuits, each independent, make unnecessary any valv- ing or flow regulation in the fuel or primary circuit. Stopping one fuel pump would result in slow back flow through that pump and its exchanger. With all fuel pumps stopped, convection would provide sufficient flow to dispose of after-heat, provided primary coolant flow is maintained. Each blanket cooling circuit will be able to provide full blanket cooling. Normally one circuit would be in full scale operation and the other pumping slowly, on stand-by. The total amount of heat to be removgd from the blanket will vary with the U253 content of the blanket. Though detalled calculations of the blanket heat removal system have not been made, it is estimateq that the total heat generation in thé blanket will not exceed 35 megawatts at any time. Hence, it is expected that heat exchangers in that system will be similar to, but smaller than, those used in the core system. 2. Heat Exchangers All heat exchangers, including those in the steam system, are of shell and tube design, with counter flow of the fluids, U-shaped shells, which minimize tube thermal stress, will be used in every case except in the reheat circuit, where tube and shell temperature differences are low, Table XV gives a summary of the characteristics of the principal heat ;xchangers used in the entire sjstem, and Figure 8 shows diagramatically the temperature conditions of each heat exchanger. The primafy heat exchangers will transfer heat from the fuel (composition T4) to the primary coolant (Flinak). The principal objectives in the design of the primary heat exchangé?s are first, dependability, and second, low fuel holdup and low pressure drop. Both of these objectives are met best by having the fuel flow through the tubes rather than on the shell 105 UNCLASSIFIED ORNL-LR-DWG 19330 PUMP OUTLET - W - = PLAN SECTION Fig. 7. Reference Design Reactor. - » ' . ¥ & - 106 - TABLE XV SUMMARY OF HEAT EXCHANGERS Exchanger Primary Secondary Fluid Fuel Flinak Flinak - Ne Location Tubes Shell Tubes Shell Material Inor Inor Shape U U U U Fluid (Hot End, °F 1200 1125 1125 1090 Temperature (Cold End,°F 1100 1000 1000 890 Temperature change, CF 100 125 125 200 Temperaturs (Hot End, °p 75 35 Difference (Cold End,°F 100 110 Heat Transfer Capacity, Mw 100 100 ' Heat Transfer Surface, ft 3040 3600 Avg. Heat Flux, Btu/hr-ft< 120,000 9%,000 Tube Data: | Length, ft 16 31 Number 2k50 930 0.D., inches .380 .600 Wall Thickness, inches .0L0 .050 Pitch (A), inches .580 .825 Tube Bundle Dia., inches 26 27 Flow, ft2/sec 12.55 12,61 12.61 31.1 Flow, gpm 5650 5680 5680 14,000 Flow, 1b/hr - - - i Fluid Velocity (inlet), ft/sec 10.5 7.4 10 15.3 ' Reynolds No. (Nominal) 6,000~T,000 6,000-8,000 12,000-17,000 2.6--3.0.:;;105 Pressure Drop, psi Lo 19 3l 18 Power Consumed, . Kw Heat Transferred Mw .98 AT Bk 93 Max. Heat Flux, Btu/hr-ft2 120,000 145,000 o v Thermal Stress, (l?vx ATewa.ll ; psi 4,000 6,000 - w - e 4 a - 107 - TABLE XV (Continued) Na-to-Water Na-to=-Steam " Na-to-Na Na-to-Steam Exchanger Boiler Superheater for Reheat Reheater Fluid Water Ne. Steam Na Reheat Na Main Na Steam Na Location Tubes Shell Tubes Shell Tubes Shell Tubes Shell Material 2 1/4 Croloy %16 SS 316 S8 316 SS Shape U L U Straight Straight Fluid (Hot End,°F 650 T70 1000 1090 1060 1090 960 1060 Temperature (Cold End,®F 515 670 650 800 860 9Lo 610 860 Temperature Change, F 135 100 350 290 200 150 350 200 Temperature (Hot End, °F 120 | 90 30 100 Difference '(Cold End, OF 155 150 80 250 Heat Transfer Capacity, Mv 61.7 ol 2 13.5 81.2 Heat Transfer Surface, ft° 2540 2000 450 7550 Avg. Beat Flux, Btylm-fte 82,700 4l ,100 103,000 37,000 Tube Data: Length, ft L5 46 1k 30 Number 28l 220 300 1190 0.D., inches 1,000 1.000 .500 1,000 Wall Thickness, inches 120 .120 .0k9 .095 Pitch (A), inches 1,800 1,800 745 1.800 Tube Bundle Dia., inches 29 28 1k 65 Flow, ft7/sec 1.8 36 20.8 5.20 4,03 5.4 682 2l ,2 Flow, gpm 810 17,300 -- 2340 1810 2430 - 10, 900 Flow, lb/hr 317,000 -- 317,000 - - -- 1,546,800 -- Fluid Velocity (inlet) 2 8.75 30 1.7 15 9,15 160 | 1,46 ft/sec ' Reynolds No. (nominal) 3.5 x 10° 6.9x10° 4.5 x 10° 1.2-1.7x10° 9 x 10° 1.65x10° 3.25 x 10° L2x10 Pressure Drop, psi 5 '; 3 6 1 20 4,5 16 1 Power Consumed, Kw Heat Transferred MW .05 3b 1.02 -Ob 1.17 e 35 26.5 .05 Max. Heat Flux, Btyhr-ft° 95,000 54,000 169,000 56, 000 Thermal Stress, (g E _ AT wa'll)p_si"'{OOO 4200 10,000 £500 (T->" 7 2 ) ) 1200 CUEL SN‘T 75 125 100 “ 100 gl 1000 PRIMARY EXCHANGER {FUEL SALT-FLINAK OT\m =87 SECONDARY EXCHANGER {FLINAK-SCDIUM } 890 AT\ m =655 SODIUM TO WATER BOILER 763 _ 770 OWUN g0 120 670 €650 155 515 A7’,m=}¢105+——— 100 130 ALL TEMPERATURES ARE IN °F 108 UNCLASSIFIED ORNL -LR-DWG 19889 SODIUM TO STEAM SUPERHEATER 1090 » 1000 - 800 150 650 O Tim= U7 SODIUM TO SODIUM FOR REHEAT 1090 1060 940 » 80 B60O A T|m = 51 o SODIUM TO STEAM REHEATER 1060 100 vh 960 200 860 @ Q/V‘ & 250 " 610 ATim= 123 B Fig. 8. Temperature-Heot Diagrams for Heat Exchangers. - 109 - slde of the heat exchangers. The temperature range of the fuel affects these £wo objectives., Increasing thé temperature range decreases the external fuel volume holdup, but at the cost of increasing the pressure drop in the exchgpger. Moderate fuel holdup in the exchanger 1s obtained with O.3-inch ID tubing. Although a wall thickness of 0.039 inches has been selected as offer- ing reasonable life and temperature drop, it is possible that additional data and analyses will make a wall thickness of 0.060 inches appear desirable. The secondary heat exchanger, Flinak-to-sodium, has flow design criteria similar to the primary exchanger, but the volume of salt in the ex- changer is a minor consideration. Consequently, larger tubes than are used in the primary can be used, which will reduce the number of tubes in a unit and simplify fabrication., One-half inch 09 x 0.050-inch wall tubing has been selected. The Na-to-water boiler and Na-to-steam superheater and reheater are designed with single wall tubes but double tube sheets., Easier contalnment of high pressure steam is achieved by putting it inside the tubes, as is standard practicé‘in ordinary boilers, ?he low pressure sodium is easily contained in the shell, As the sodium tube sheet and the steam tube sheet are at different temperatures, sufficient length of tubing must be provided between them to glve low bending stresses in the tubes, To keep this parasitic length low, the tubes should be as small in dilameter as possible, but reliable fabrication 1s easier with large tubes and thicker walls. One«inch ID x 0.120-inch wall tub- ing has been selected as a_practical solution for these conflicting demands. The boiler is of the "once-through" variety and is designed £o deliver steam superheated by about 50°F to the superheater. This assures that no moisture will enter the superheater, The superheater is designed to deliver - 110 - steam to the turbine at 1000°F, and an attemperator will be placed in the steam line ahead of the turbine to insure control of the maximum temperature. Throttle valves in the feed water supply line will be used to main- tain balance between the eight separate circuits, while throttle valves in the sodium lines of each separate circuit will maintain the balance of the heat supply among the boiler, superheater, and heat exchanger.“ A separate sodium circuit will be used for reheat, with the reheater located near the turbines. A sodium-to-sodium heat exchanger in each of the six main sodium circuits will supply heat to the reheat sodium circuit. The insertion of this extra heat exchanger link, together with the lower heat trans- fer properties of the lower-pressure reheat :team (350 psi) means thgfi reheat- ing to 960°F 1s more economically accomplished than reheating to 1000°F, This does pot-appfeciably reduce the turbine efficiency, but only requires that the first expansion be carried to a lower pressure than for reheating tp_lOOOoF° With the reheater located near the turbines, 5‘percent pressure dr;p is predicted as compared to 10 percent when the steam is returned to a furnacgeég/ The reheater, and probably the Na-to-Na exchangers in the reheat cir- cuit, can be of the straight tube design without causing undue thermal stress., As the reheat boller has, of necessity, a large number of short tubes, the U- shaped configura@ion would be difficult to fabricate., Thus, it is fortunate that low stress permits use of the straight tube dgsign; ‘ As shown in Table XV, the primary and secondary heat exchangers will be made of INOR-8, the boilers of 2 1/4 Croloy, and the superheaters, reheaters, and sodium-to-sodium heat exchangers of 316 stainless steel, Although these §Q/ Shannon, R, H., and Selby, J. B., "Double Reheat Cycles--The Next Step?" .Power, February 1953 - 111 =- are suitable choices at present, it is indicated in Section II that future ex- perimental work may lead to the liberal use of high nickel alloys such as Inconel in the sodium, water and steam systems, 3. Steam Cycle The steam cycle selected for design consideraticn uses 1800 psia, 1000°F steam, with one reheat to 960°F in 3600-1800 rpm cross-compound tur- bines, Figure 9 gives a diagram of the steam cycle, showing pressures, tem- peratures, and mass flow rates of the steam. Seven bleed-offs are used for heating feed water to 515°F, With a condenser pressure of 1 1/2 inches Hg (920F), a turbine heat rate of 8070 Btu/kwh is achieved. This corresponds to a steam cycle efficiency of 42.3 percent. Coal-fired central stations use six percent of the gross power out- put in.auxiliaries; seven percent appears to be a reasomable figure for this reactor power plant. This latter figure is reached by subtracting one per- cent for draft fans, coal pulverizers, ash handling, etc., connected with a fossil fuel furnace, and adding two percent for pumping fuel and coolants associated with the nuclear furnace. Net station efficlency is 39 percent or 8675 Btu/kwh delivered at the bus-bars. For comparison, an efficient coal-fired unit, the Commonwéalth Edison, Chicago, State Line Plant, reports 8550 Btu/kwh sent out from the 191 Mw Unit No. 3. At this preliminary state, no attempt has been made to optimize the steam cycle on a cost basis, or to select an extremely efficient cycle. The conditions selected--1800 psi, 1000°F, reheat to 960°F--are in line with proven industrial practice, and require that the.design of the reactor system face up to problems of pressure, temperature, and han- dling of reheat. The steam cycle proposed is based on the cycle used in the Astoria Station of Consolidated Edison Company, New York, which UNCLASSIFIED 340A 960F 1,546,746W ORNL-LR-DWG 19879 ,! 1800A 1000F 1,851,060W ‘ REEAT 90A 648F 1,281,674W INTERMEDIATE PRESSURE TURBINE SUPER HEATERS [ T A o LOW LOW NER R TURBINE GENERATOR [j GENERATO PRESSURE PRESSURE B Pl w Y = o z 3 o oJ H')P 2 0 w - L ) g 1%-inHg 92F 1t o 4 7 7 3 1,132,479 W O @ I o o m m a T 59,789 W 55,632 W - T Y | ] | ] d BOILERS S50F 58,308 W z ® = 5 [ S ? © 0 o 3 \EVAPORATOR o w o - r~ = g o~ ) w o - L g = " 0 3 L e L - o = < = - < @O @< =z P~ oJ a W N I 5 ~ N ~ Q N~ |8 & o s - ©w =z 0, - O 0 - <1 8 8 | - w|o L o M= I ek |0 n ® - 0|0 n ¥ 7 o ' ' 502F 304F 1 ! ‘ i 395,386 W 429F 373F 260F 210F 160 F 100 F NN\ NN\ AN\ - NN AVAVAV: L AN N AN 449F 393F \DEAERATOR 250F 200F 140F a7F 304,314 W 129,019 W 190,104 W 263776 W LEGEND A = PRESSURE, psia 312F F = TEMPERATURE, °F w = FLOW, ib/hr - # - | ] < . ¢l - 11% - was the result of an extensive study éi/’ Qg/, Modifications were made to meet the special conditions of the heat supply and cooling water temperature available. A feed water temperature about lOOQF below the saturation témpera- ture of the boiler seems to be as low as is desirable for a sodium boiler, based on thermal stress considerations. This 1s higher than is normally used with & fossil fuel boiler, where lower tefiperatures allow more heat extraction from the flue gases, In compensation for the additional feed-water heaters, however, the higher feed water temperature increases the steam cycle efficiency slightly. C. Compgpents‘and Component System l. Pumps The molten salt and liquid metal pumps will be of the same general design as those now in use in ANP, except for differences of size and capacity, and refinements of details which may develop before the time for procurement arrives., One pump of each new size (capacity and head) will be given thorough tests before additional pumps of the same size are procured, and each pump procured will be proof-tested before use in a power plant. The pump sizes re- quired are indlcated as to flow and head by the data in Table XV describing the hea@ exchanger, Pumps, compressors, fans, etc.,, for all purposes, except for the movement of molten salts and liquid metals, will be conventional, 2. Valves Salt and sodium mechanical stop valves will be bellows sealed, Ql/ Milne, G. R.,, "Basic Study for a Generating Station", ASME Transactions 82/ Milne, G. R., "Cost Amalysis for 160 Mw Units Favor 1800.ps1--1000°F. Reheat", Electrical World, February 26, 1951 - 114 - cermet-seat type. No salt or sodium stop valves will be used on other than fill-and-drain lines, and these will be 2-inch IPS. No flow control valves are required in molten salt lines, The flow control valves in sodium lines will be bellows sealed., They will not be required to stop flow, but dnly to control it. All molten salts and sodium valves will follow the most recent ANP design for such valves, except for size and possible modifications for diminish- ing fluid head loss. 3., Pipes and Tubes Pipes and tubes must be designed to absorb all temperature transients and differences without exceeding specified stress levels. 1In general, they will be anchored at walls and at heavy pleces of equipment, and appropriately shaped bends between anchor points will be provided to absorb all expansions, contractions, and twists. The main piping in the fuel salt circuits leading to the primary heat exchangers will be 12 inches in diameter., The blankep cir- cuits will use lO-inch . pipe. The coolant salt systems will use lk-inch . pipe in the core heat.removal system and 10-inch . pipe 1n the blanket heat removal system. All these pipes will be made of INOR-8., L, Fill-and-Drain Tanks a, Fuel salt fill-and-drain tanks Two tanks will be required, each with sufficient capacity to contain all the fuel salt in the circulating fuel system (approximetely 500 ft5)° One tank will be available for temporary storage of used fuel salt while the othepy tank is being used to serve the continued dperation of the power plant. Each of the tanks will be designed so that it can be heated to 1200°F, and so that maximum after heat (approximately seven Mw total) can be removed , - 115 - with the hottest part of the shell cooler than 1500°F and the axial salt tem- perature considerably less than the fuel boiling point. It is estimated that this criterion can be met by fabridating the tanks from a number of 12-inch I1.D. pipe, aggregating to a total length of 600 feet. Shell temperatures under after-heat conditions will be maintained by recirculating air, in turn cooled by water-cooled radiators, Provision will be made for powering the cooling gystem from emergency power (diesel or battery) if the regular power supply should fail. Heating will be done by electrical resistance, Tank fittings required will include the following: (1) salt charge lines; (2) salt drain line; (3) line for transferring molten salt to and from the circulating-fuel system; (4) inert-gas supply line; and (5) inert-gas discharge (off-gas) 1line, b. Blanket salt fill-and-drain tanks Two tanks will be required, each with sufficient capacity to contain all the blanket salt in the circulating blanket system (approximately 600 ft5). One tank will be available for storage of used blanket salt while the cther is being used to serve the_cpntinued operation of the power plant. Each of the two tanks will be designed with provisions for heating to l}OOoF. No special provision need be made for removal of decay heat other than including capacity in the storage room atmosphere cooling equipment to remove this additional heat from the room. Tank fittings requipred will be the same as for the fuel salt tanks, ¢, Intermediate hHeat transfer salt fill-and-drain tanks Four tanks are required: (1) two duplicate tanks, each capable of containing all the salt in all the Intermediate heat transfer circuits; and (2) two duplicate tanks, each capable of containing the salt for the largest - 116 - intermediate heat transfer circuit. Each of the four tanks is to be connected for filling and draining each intermediate transfer circult separately. Each of the four tanks is to be provided for heating to lEOOOF, Tank fittings required are the same as for the fuel salt fill-and- drain tanks, d. Sodium fill-and-drain tanks Four tanks are required: (1) two duplicate tanks, each capable of containing all the sodium in all the circulating systems; and (2) two dupli- cate tanks, each capable of containing the sodium for the largest single sodium circuit. Each of the four tanks is to be connected for filling or draining each of the sodium circuits separately. Each of the four tanks is to be provided for heating to 1200°F. Tank fittings required are the same as for the fuel salt fill-and- drain tanks. 5. Gas Supply Systems (Helium, Nitrogen, and Compressed Air) Helium gas, contalning less than 10 ppm.oxygen and having 4 dew point less thanm minus TOOF, wlll be the only gas allowed to come in contact with finy salt or sodium, except during chemical processing. Supply lines from the helium storasge banks will contain pressure reducing valves, shutoff valves, flow and pressure instrumentation, and safety devices as required for safety and dependability. The consumption of helium will depend on & number of design details, including: (1) dilution of off-gas; (2) use of helium in instrumentation; and (3) pump design for pumps serving barren salts and liquid metal systems. Nitrogen may be used as an atmosphere around some assemblies afid*aub— assemblies of equipment. If so, the handling equipment will be similar to the - 117 - heligm’handling equ%pment, but with less strict purity requirements. Whether or not nitroggn will be used anywhere in the plant will depend on future deci- sions as to ambient atmosphere requiréments. Compressed alr will probably be requifed for some instruments, but is a comparatively minor item. 6. Off-Gas System The xenon and krypton evolving from the fuel at the fuel-to-helium interface in the fuel expansion chambe; will be (1) diluted with helium in that chamber, (2) removed through a tube to a holdup tank to allow time for short half-life decay, thence (3) passed through a charcoal bed for further holdup and decay, then (4) discharged to the atmosphere through a stack, The bases for design of ;he system will be the safie as for the HRE, tPe HRT, and the ART 92/. Parameter relations are becoming well established QE/ and continued experimentation is refining the quantitative knowledge still further, The present design basis (for later refinement) is the ART off-gas X system, modified to ten times the ART off-gas production, 7. Preheating and Temperature Maintenance Provision will be made for preheating the following items to 1200°F in 48 hours, although a longer tifie will normally be used for preheating: | reactor, salt expansion chamber-pump units, salt and salt-to-sodium heat ex- changers, connecting pipes and tubes, salt and sodium filltgnd-d;ain tanks and fill-and drain lines, salt fill-and-drain tank charging lines, new;éalt storage vessels (from which molten salts are charged into salt fill-and-drain 83/ Reference Quarterly Reports 84/ ORNL-2116 - 118 - tanks), and off-gas lines to off-gas system. Provision will be made for pre- heating the following items to 1000°F in 48 hours: superheaters, boilers, re- heaters, sodium lines, and sodium stérage vessels gé/o The boilers will have provision for additional heat input for start- up, both in the boiler and upstream of the boiler, so that excessive thermal stresses will be avoided during start-up. The general pattern of boller start- up operation will be very similar to that in a conventional power plant. It is anticipated that most of the heating of reactor system com- ronents will be done with heaters consisting of electrical resistance wire embedded in shaped ceramic matripes.. More detailed engineering of components may show that it is more convenient, more dependable, or cheaper té heat some parts of the system with direct electrical resistance, electrical induction §§/, gas heat, or auxiliary steam, | | Preheaters will be so arranged as to allow separate temperature con- trol of individualjsubassembliese D. Plant Layout Figfi;gs 10 and 11 show a workable disposition of the major power plant components. “ Figure 10 is a plan view and shows the reactor surrounded by six primary (Tuel salt tolFlinak) heat exchangers and two blanket-cooling heat excgangers. All of these components should be as closely gfouped as remote maintenance will allow, in order to minimize the volume of fuel salt. Directly above the reactor and ips primary heat exchangers is a shield containing removable plugs which allow access from above. (See 85/ May be designed for slower preheat rate 86/ Used by Atomics International on the SRE - 119 - Figure 11.) This shielded region above the reactor is allocated to tools for remote disassembly &nd reassembly of fuel salt pumps or primary heat exchangers. It may also be used as a temporary storage room for radicactive parts, o Epclosing the reactor primary complex and the shieldedGSPace above is a 1arge”steel shell to contain fission fragment gases should they escape from the primary system. It also will allo# for maintaining a relatively inert atmosphere should repairs to the system become necessary. The lock lead- ing to the vessel makes it péssible to take parts or toq;s fn or out of the vessel without appreciably affecting its atmosphere, Personnel may, if nec- essary, enter the shielded space above the reactor to service or alter. the re- mote handling tools. The primary heat exchangers are showp_in“a vertical position and are connected to the main piping py welded flange Jbints., This arrangement is believed to provide forrthe least difficulty in thegremote operations of removing and replacing heat exchangers. The primary pumps will be of the tap access variety, to make the replacement of the motor-impeller portion less difficult by remote control. The pump units will probably be held in place with a bolted flange. It is recognized that much detailed design and develop- ment work will be required to insure adequate operation of remote maintenance -equipnment. Surrounding the steel shell and behind concrete shielding are eight compartments, each one in turn shielded from the others, These compartments contain the secondary heat exchangers (Flinak-to-Na) and Flinak pumps. Remov- able shield plugs above each compartment will allow access to, and replacement of, the components containeq therein, The sodium is conducted from the above-mentioned compartments to four separate layers of boilers, superheaters, pumps, and blenders; The top layer is to serve the blanket cooling system. UNCLASSIFIED ORNL-LR-DWG 19337 __—SECONDARY HEAT EXCHANGER CONDENSER SERVICE WELL -4 HEAT EXCHANGER TO REHEAT-- — GENERATOR U ! ) | ~— TN "= | Ow PRESSURE TURBINES AT —_ {0 O ACCESS REGION FRIMARY HEAT EXCHANGER ‘o T v CY - ) \ ‘ ‘ ‘ | i R — . ! . i\ mREHEATER CCONTROL ROCM { b Fig. 10. Plan View of Power Plant. 0¢l UNCLASSIFIED ORNL-LR-DWG {9373 _-Na BOILER CIRCUIT PUMP SECONDARY HEAT EXCHANGER SPACE LOW PRESSURE TURBINE LOW PRESSURE TURBINE HIGH PRESSURE TURBINE GENERATOR GENERATOR SHIELD REMOVABLE PLUG — NeTONa .~ * ~ o - HEAT EXCHANGER FOR REHEAT 100 10 20 30 SCALE IN FEET Fig. 11. Section Through Reactor and Power Plant. 12t - 122 - From the boilers and superheaters on, the system and building arrange- ment closely resemble a conventional steam-electric power plant. However, in place éf coal and ash handling equipment; space not shown im the layouts must be provided for fuel, Flinak, sodium, and blanket salt sforage vessels,'and chemical processing eQuipment, It is visualized that these facilities would be located below and at the side of the reactor complex opposite the steam plant. Space for the auxiliary features of a conventional power plant, such as water processing, machine shop, instrument shop, offices, etc., mflst, of course, be provided. E, Chemical Processing and Fuel Cycle Economics 1., Core Processing The core salt chemical processing system is a combination of the ORNL fluoride volatility and the K-25 UF6 reduction processes discussed in Secfion, Part F. The core salt is transferred by gravity or inert gas pressure . from the core loop to a holdup vessel where short-half-lived fission products are allowed to decay before the salt is transferred to the fluorination vessel. The reactor fill-and-drain tank is felt to be the ideal holdup vessel since it is already designed for cooling hot sal£ and since its volume is large com- pared to the heldup required by the chemical'plant, If thirty days proves to be an adequate cooling period, about sixty cublc feet of core salt will be stored at one time, The core salt will be fluorinated in 1.l ft” batches (ORNL pilot plant size) at one=-two batches per day (600 ft3/yr), The barren salt, stripped of uranium but containing most of the Pu, fission products, and corrosion products, is transferred to tank storage where it is held for future salt - 12% - recovery., The 10-kgU-capacity NaF pellet beds (ORNL pilot plant size) will be’operated on a once-or-twice-per week cycle (i,e,, severalifluorination cycles per sorption cycle). The volatility part of the chemical plant produces liquid UF6 in cylinders, The UF6 reduction tower will operate simi-continuocusly, using UF6 from the volatility plant as feed. It will discharge its UFh product directly to a fuel salt mixing pot, to which is also fed fresh salt and make- up UFb{° Uranium losseg in chemical processing are quite low, apout 10 ppm in waste salt and 2 ppm in waste gases. This is approximately 1 kg/yr uranium ioss, 2. Blanket Processing Chemical processing of the blanket salt 1s physically much the same as that of the core salt, except that after fluorination,the blanket salt with the Pa and fission products that it contains is returned to the blanket system. Because of the much lower power density in the blanket, holdup for the fission product cooling is not a problem, Separate fluorinators for core and blanket salts will prevemnt pqssible cross-contamination. A continuous fluérinator for the blanket is assumed since complete uranium recovery is not necessary, and although continuous fluorina- tion has not been demonstrated, no new basic development is required., (Con- tinuous fluorination of the core salt would require further development to demonstrate complete uranium recovery.) A conservative blanket processing rate of one blanket loop volume per year is assumed (achievable even with batch fluorination) though a much faster rate is probably possible. Separate NaF beds and cold traps are provided to enable withdrawal of pure U235F6 from the system if desired, It is assumed that the same UF, reduction tower will serve both core and blanket, - 124 - 3. Chemical Processing Costs s a, Investment Costs - » The ORNL volatility pilot plant'capital costs can be broken down as follows: Replacements and Design Construction Modifications Through FY-56 $ 130,000 $350,000 $ --- Budget FY-57-58-59 210,000 296,000 328,000 Subtotals $ 340,000 $646,000 $328,000 Grand Total $1,314,000 These figures include replacements and modifications which should not be required in a second plant. They also include solid fuel element han- dling and dissolution facilities which would not be fequired in the reference design reactor plant., These costs do not include building and service facil- » ities for reducing UF6 to UFh and reconstituting fuel. The over«ali plant cost estimate in Appendix I includes the following capital costs assignablg‘to the " chemical plant: v Building and site $ 250,000 VI Equipment and installation | 1,500,000 VII Design 262, 500 VIII Prime Contractor 402,500 IX Contingency __ho2,500 $2,817,500 | While the breakdo#n may not be exactly accurate, the total is felt to be an adequate estiméte of the cost. These investment costs are charged v at sixteen percent per year. - 125 - b, Operating costs The ORNL pilot plant operating budget for three fiscal years--1957, 1958, and 1959--totals $1,368,000, The reference design reactor chemical plant would have lower "unusual costs" but higher "production-proportional costs" and is estimated to cost $500,000/yr to operate, not including core salt replace- ment cost which adds $810,000/yr (600 ft5/yr at $1350/ft5). These costs are included with other fuel cycle costs in the discussion cf the over-all power costs that follows, Other fuel cycle costs are explained in Section II, Part G, of this report. 4, Fuel Cycle Economics The nuclear characterlstics of two region molten salt reactors have been discussed in Secticn II, Part D, General aspects of fuel cycles for molten salt power reactors have been discussed in Section II, Parts F and G, together with the interdependence of dollar and néutron economics, For the purpose of estimating fuel cycle costs for the reference design reactor, the reactor of Column D, Table XIV, was chosen. (Reactor "E" had a much lower fuel cost but assumed U235 make-up at $17/g.. If U255 were available at $34/g such a reactor would have fuel costs about like "D" but would "look better" because of the higher breeding ratio. Reactor "G" had a fuei cost ~~ ten per- cent lower than "D" but had only a five foot core., Reactor "C" had very nearly the same fuel cost as "D", with better breeding ratio but higher salt make-up charge ) In Section III, Part F, the chemical plant construction cost has been caMbined with the other reactor complex capital costs The fuel cycle "operating" costs are shown in Section III, Part F, as 2, 3 mills/kwh, which results from Reactor "D" as follows: - 126 - $/yr Mills/kwh » 23 + 25 rental’ 441,000 0.3 25 make-up 2,030,000 1,2 & Salt make~up 810,000 0.5 - Chemical plant operation 5002000 0.3 ) 3,781,000 2.3 The first three $/yr figures come from Table XIV., The chemical plant operation estimate is given above in this section. The conversion to mills/kvh 1s based on 240 Mv electricity for 7000 hrs/yr. The cost difference between having the reactor on stand-by or on the line at full power is 1.7 mills/kwh, the sum of "25 mgke-up" and "salt make-up," F. Cost,Analzsis 1. Introduction ' Any cost analysis of a system such as this natura;;y breaks into three categories: (l) materials and component development costs. necessary be- fore construction; (2) design and constructiéncost; and (3) operating costs, if is on this basis £hat_costs are estimated for the reference design reactor. For a detaile& analysis, certain assumptions and decisions had to be made, It is assumed for this cost analysis that the reference design reactor will be the next molten salt power reactor constructed, This means qpecifically that it is assumed its construction wouid not be preceded by the construction of a smaller reactor and that most of the development work undertaken would be pointed specifically at this reactor. Implicit in the cost analysis are all of the decisions outlined in the above description of . the reference design reactor. It is further assumed that for both the o -127- development and cogstruction the timing will be such that this work can proceed in an orderly and businesslike manner without either undue delay or extreme urgency. 2. Materials and Components Development Costs The present state of knowledge of this system is such that it is reasonable to expect that a large part 6f the costs to be incurred prior to consfiructing a power station will be for extrapolating and improving present designs and testing particular components. | Section IT of this report gives a review of the present-state of technology as applied to this reactor, and any development program is necessar- ily based on the background presented. Specific points of importance are: 1. An alloy (INOR-8) exists which has satisfaptory mechanical and fabrica@ion properties and fqr which there is reason to expect adequate corro- seion resistance to the sgits used., However, long-term corrosion tests under the specific condition imposed by the referénce design reactor are yet to be conducted. | 2. DSatisfactory %iquid metal and moltén salt handling techniques are known. 3« Batisfactory equigmept for pump%ng liquid metals and molten salts has been developed and adquateky proven, but will have to be redesigned and reproyen in sizes and for operating periods appropriate to the needs of thls system. L, A satisfactory chemical process has been devised and is in pilot plant operation for recovery of uranium from molten salt.. | 5. The temperature coefficient of reactivity of the syétem is such that no mechanical reactivity control devices are needed other than equipment for fuel addition, - 128 - O this basis the items listed below would constitute the develop- ¢ ment effort required. Items one and two are required to prove that the y materials proposed are compatible and suitafile for long-term reactor use. (1) Out-of-pille pumped loops and natural convection loops $1,000,000 (2) 1In-pile loops, at least one each for the core an& blanket fluilds 125002000 | Subtotal $2,500,000 These two items constitute the initial investment in déveiopment work, and further expenditures for development and construction Qf the reference design reactor would be spent only 1f the expected favorable results were real- ized. The $2,500,000 listed for items one and two represent this optimistic expectation., If disappointing results were obtained, or if a great number of experiments were performed, the amount could be considerably greater, - Before construction of a reference design reactor, the folldwing . items should have development attention: (3) Pump for.circulation of reactor core salt, blanket salt, and intermediate coolant salt () TFuel-to-salt heat exchangers (5) Coolant salt-to-sodium heat exchangers (6) Piping and vessels (7) Instrumentation (8) Chemical processing (9) Critical experiments (10) Remote maintenance equipment » (ll) Sodium pumps » (12) Sodium heat exchangers, boilers, superheaters, blenders, and valves - 129 - For the last two items, only acceptance tests are.included in the estimate. It is thus assumed that any basic problems of sodium-to-water boilers and superheqters which are not already solved will be solved elsewhere, Although rough estimates for individual items 3-12 have been attempted, it is not believed that their accuracy warrants listing them individually. Estimates of the costs of development for all the items 3-12 vary from $12,000,000 to $19,000,000, A very large uncertainty is associated with item ten--remote maintenance equipment--this item canndt be assessed completely without further study. | If a molten salt reactor program is to be taken seriously, there must also be supporting research carried out in metallurgy9 chemistry, and solid state physics. A portion of this work would be aimed at'future better modifications of the molten salt system. It is important that this research bé started early because cf fhe natural time lag between early research re- sults and practical operating systems, Hence, the following item has to be included in the total research and development costs: (13) Supporting research and development --$600,000 per year Estimates of the total research and development costs (items 1-13) range from $18,000,000 to $27,000,000. 3. Deslgn and Construction Costs The best available system flow diagrams for the reference design resctor have been broken down in@o individual components insofar as possible, and costs of purchasing, inspecting, and installiné“these components have been estimated on the basis of standard engineering cost estimating procedure as modified by experience in the nuclear power field., These modifications are extensive, as & result of the higher standards required and the necessity for multiple inspection of every piece that goes into a reactor. - 130 - Wherever individual components were too small or numerous to be properly J isolated, an attempt has been made to assign a cost figure to an entire subsystem, such as in the cases of the inert gas systems of helium and nitrogen. In this manner, an estimated construction cost that is felt to be on the conservative gide was reached. This includes 15 percent for engineering design; a sum of $1,000,000 for a period of start-up operations before the plant is "on line"; a contingeqcy fartor of 20 percent of all reactor costs; and a factor of 23 percent which has been found normal for prime contractor fees in reactor con- struction: The conventionai portibn of the plant, turbine; generator, etc., has not been treated in this manner as costs here on an installed basis can be obtained with a high degree of accuracy §Z/o This preliminary, but detailed, cost analysis shows an anticipated cost of $238 per installed kilowatt of generating capacity. A detailed break- down of the cost estimate is shown in Appendix T. 4y, Cost of Power from the Reference Design Reactor , Power costs fall naturally into three categories: (1) fixed costs, (?) operation and maintenance, and (3) fuel cycle costs, Fixed costs are those charges resulting from capital investment in the plant. In this study, this investment 1s calculated as shown in Appendix I to be roughly $57,000,000 or $238/kw of installed capacity. Of this invest- ment, $99/kw is for the conventional portion of the plant and $139/kw for the reactor complex portion, incluaing_chemical processing equipment, As pointed out by Mr, W, K. Davis §§/, a charge of roughly 12 percent per year is applicable 87/ See, for example, "Ninth Steam Station Cost Survey", Electrical World, October 1955 88/ Progress in Nuclear Energy, Vol., VIII, McGraw Hill (1957), p 215 O - 131 - to both these investment figures due to financing and taxes., If, for pur- poses of computing amortization costs, one assumes & 4O-year life for the conventional portion of the plant and a 20-year life for the remainder, these add two and four percent, respectively, to these charges, giving a fixed cost of 1k percent per year on the $99/kw of conventional plant and 16 per- cent per year on the $l§9/kw of reactor portion. Using the accepted load factors of 80 percent for a base load plant such as this results in a fixed charge of 5.1 mills per kwh, Operation and malntenance expérience in conventional coal-fired plants since the last war has been that these charges in plants of this size have amounted to O.h mills/kwh, and that about 0.3 mills of this charge was for thg coal burning and heat transfer equipment., Assuming reactor experi- ence three times as bad leads to a cost (0.3 x 3 + 0.1) for operation and maintenance of 1.0 mills/kwh, Fuel and fuel processing costs, as discussed in Section III, Part E, of this report, amount to 2.3 mills/kwh. These costs thus add up as follows: Fixed costs 5.1 mills/kwh Operation and maintenance 1.0 Fuel and fuel processing 2.3 Total power cost 8.4 mills/kwh It is probable that a reactor of this type could be built and would operate safely for some time. The principal uncertainties concern the life of the components and the ability to replace them in case of failure. These un- certain points should be investigated as recommended in Section I before the construction of a large reactor is undertaken. - 132 - The costs outlined above for the construction and operation of the reference design reactor are predicated on obtaining favorable results from a development program aimed at removing these uncertainties; with this pro- vision they are believed to be as realistic as is feasible at this time, I, - 133 - APPENDIX COST ESTIMATE OF REFERENCE DESIGN REACTOR Fuel Circult A, B. D. Reactor core Heat transfer equipment 1, six 6000 gpm pumps ) 2, s8ix 2-speed motors, 300 HP ) 3, six 100 mw fuel-to-coolant salt exchangers, 16 £t L x 29 inches D, 8000 1b, 3040 sq ft at $55/ft2 4. 100 ft of 12-inch INOR pipes, 5700 1lb at $4/1b 5. 2b welds of 12-inch INOR pipes at $600 6, insulation T. heating equipment Fuel handling equipment 1. 2 full volume fill-and-drain tanks (INOR pipes), 450 ft3 each for radioactive core salt o, five 2-inch shutoff valves (INOR) (remote control) 3., 50 ft 2-inch INOR pipe Ik, heating equipment 5. fifteen 2-inch welds, etc. 6., insulation Auxiliary equipment l, one 4000 £t water tank plus pipes, pumps, etc, 2, one 1000 £t° S8 30k pipe to be put inside H,0 tank to hold gases temporarily 3, one 1300 £t~ SS 304 pipe full of charcoal L, one 150-ft exhaust stack, 4-ft D,, metal 5. fuel enrichment and sampling equipment (1n addition to chemical plant) Subtotal $ 650,000 660,000 640,000 22,800 14,400 9,600 9,900 240,000 8,000 800 2l 000 3,000 1,500 12,000 9,000 13,000 5,000 50,000 $2,373,000 - 134 - II. Blanket Circuit 1. 2. 5. L, De A, Container vessel, pump inlet header, expansion tank heating equipment insulation supporting structure and foundation biological shield, reactor room B. ‘Heat transfer equipment 1, 20 3. L, Do 6. To two 4000 gpm pumps ) two 2-speed motors, 150 HP ) two 35 mw salt-to-salt exchangers, 1500 fte, 4000 1b, $h5/ft2 30 £t of 10-inch pipe (INOR-8) 8 welds 10-inch pipe insulation heating equipment C. Blanket salt handling equipment 2 full volume fill-and-drain tanks for non-radicactive 40,000 1. 2. 5. b, 50 60 blanket salt, 500 ft° each five 2-inch shutoff valves (remote control) 50 f£ 2-inch INOR-8 pipe heating equipment insulation fifteen 2-inch welds D, Auxiliary equipment 1, 2. 30k SS connections to off-gas system sampling system (in addition to chemical plant) Subtotal III. Coolant Salt System (6-core, 2-blanket) A. Core system 1. 2. 900 ft li-inch INOR pipe six 100 mw salt-to~-Na exchangers, 31 ft L x 30 inches D, 1100 1b, 3600 t2 at $5o/ft2 $ 250,000 4,000 4,000 100,000 250,000 140,000 135,000 5,500 - 4,000 3,000 2,000 8,000 800 7,000 6,000 3,000 10,000 30,000 $1,043,000 25,000 650,000 ‘- - 135 - IIT-A (continued) 3, six 6000 gpm pumps ) $ 540,000 L, 6 motors, .300 HP ) 5. 2L shutoff valves (remote control) 38,000 6., 2 INOR-8 fill-and-drain tanks (for all 8 systems), 64,000 6450 Tt 7. 2 INOR-8 fill-and-drain tanks, capacity 1 system each, 15,000 200 £t° 8. heating equipment 35,000 9, insulation 40,000 10, welding 100,000 B. Blanket system 1., 210 ft 10-inch pipe 36,000 2, two 25 mw salt-to-Na exchangers, 31 ft L x 20 inches 96,000 D, S 304 shell, 4500 1b, 1200 ft° at $40/rt° 3, two 3000 gpm pumps ) 100,000 4, 2 motors, 125 HP ) 5. 6 shutoff valves (remote control) 12,800 6. heating equipment 11,400 7. insulation 9,600 8. welding | 18,000 Subtotal $2,011,000 IV. Sodium Clrcuits A, From core loops 1, 23%0 ft 18-inch pipe SS 38,800 2. 1450 ft 1l6~inch pipe SS 191,500 3. 3%0 ft 12-inch pipe SS 28,200 L, 360 ft 10-inch pipe SS 23,200 5. 690 ft 8-inch pipe S8 31,600 6. 540 ft 6-inch pipe SS 16,500 7. six 14,000 gpm pumps ) 660,000 8. six 2-speed motors, 375 HP ) 9. 6 Na-toE0 boilers, 45 ft L x 32 inches D, 381,000 Croloy, 24,000 1b, 2540 £t at $25/ft IV-A 10, 11, 12, 13, 14, 15, 16, 17. 18. 19, 20, 21, 22, - 136 - (continued) 6 Na-to-steam superheaters, 46 ft L x 32 inches D, SS, 2000 £t2 at $30/ft2 6 Na-to-Na exchangers, 1k ft L x 16 inches D, 2000 1b, 450 2 at $30/ft2 six 18,000 gpm Na-to-Na blenders, 3 ft Dx 3 ft L, 3/8-inch t six 1L,000 gpm Na-to-Na blenders 6 cold traps and pro rata share of accessory equipment 30 control valves (incl. control mechanism) 36 shutoff valves (Na) 2 full Na system fill-and-drain tanks, 7000 f£t° at $11/£¢° 2 Na fill-and-drain tanks (largest single circuit) 1000 £t° six 18,000 gpm pumps (Na) ) six 2-speed motors, 115 HP ) welding heating equipment and insulation B, From blanket locps l. O i F W o o 10, 11, 12, 490 ft 10-inch diameter pipe, SS 200 ft 6-inch diameter pipe, SS 2%0 ft h-inch diameter pipe, SS two 5000 gpm pumps ) two 2-speed motors, 110 HP ) 2 Na=t0=H20 boilers, 45 ft L x 19 inches D, 9000 1b 850 £t° at $35/ft° 2 Na-to-steam superheaters, U6 ft L x 19 inches D, 700 £t° at $ho/ft2 2 Na-to-Na 6000 gpm blenders 2 Na