LB S gfl‘ ; g 3 4455 031387 3 - ’i“* ORNL-2751 A jleacfo-s—FfoiVer i TID-4500 (14th ed. i ot . .9: /‘ LN | . .hé N R Jie J.. i 4 G, ! A e | 5 . & iy i Lt 4 a 1 1 W v | ", i s NUCLEAR CHARACTERISTICS OF SPHERlCAL, HOMOGENEOUS, TWO-REGION, Lo AN - Tl e MOL TEN-FLUORIDE-SALT REACTORS | . Alexander L.G D. A. Carrison H. G J. T £ otur . MacPherson . Roberts T OAK RIDGE NATIONAL LABOQM%Y operated by ; UNION CARBIDE CORPORATION | e for the | ‘ R U.5. ATOMIC ENERGY COMMISSION' it | R ] T‘o‘ TR N2 you wnsh someone RV _document, send in name with arid the librory will orrange -~ Printed in USA. Price Z?_q_,_.enfs. Available from the Office of Technical Services Department of Commerce Washington 25, D.C. LEGAL NOTICE This repors was prepared as an account of Government sponsored work. Neither the United States, nor the Cammission, nor any person acting on behalf of the Commission: A, Makes ony warranty or representation, expressed or implied, with respect to the accuraey, completeness, or usefulness of the information contairned in this report, or that the use of any informotion, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of eny information, apparatus, method, or process disclesed in this report. As used in the above, ‘'person acting on behalf of the Commission’" includes any employee or contractor of the Commission, or employee of such contractor, 1o the extent thot such employee or contracter of the Commission, or employee of such contractor prepares, disseminctes, or provides access to, any information pursuant to his empleyment or contract with the Commission, or his employment with such contractor, ORNL.-2751 Reactors—Power TID-4500 (14th ed.) Contract No. W-7405-eng-26 REACTOR PROJECTS DIVISION NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS, TWO-REGION, MOLTEN-FLUORIDE-SALT REACTORS L. G. Alexander H. G. MacPherson D. A, Carrison J. T. Roberts DATE ISSUED SEP 161909 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNICN CARBIDE CORPORATION for the e AL 3 445k 03L1387 3 CONTENTS ABSTRACT tuturrerretiiteeereserisriresereesiessersessstessessessrerseessesessesssessasesssesesesseeersseseerenenes 1 PRIOR WORK tvvevttrttvrrersvrietereereressereereeessessresssssssesessssnsssssessssssssssnsssssssssnsnsssssessens 2 METHOD OF CALCULATION tttttetteeeeeviieeiriesiussstrtrnrerereeesesseesesesssssensssseesessessaessaarons 4 HOMOGENEOUS REACTORS FUELED WITH U?%3% (it vninniriiinr e e e e seeessennans 4 It Al ST S Lttt ittt it et it er et ittt aesaenerneeaeneansnaesrnsraarnatnransenreranerns 4 Neutron Balances and Other Reactor Variables cuvirviiiissiieeieiiisieersiostsniensonnseenmacansssncans 7 HOMOGENEQUS REACTORS FUELED WITH U233 1ottt 9 NUCLEAR PERFORMANCE OF A REFERENCE DESIGN REACTOR ceveevviiennnnnnvererrnenes. 14 COMPARATIVE PERFORMANCE +vvvvvttuentttetiniiersersseseessesssesssssssnsssassnsesssnsssssannssesssnnn 18 ACKNOWLEDGMENTS 1vvvereeiiiisierssnunereetestrerresessessssssesossssseesssssesessesssessesnsnnsessessess 18 NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS, TWO-REGION, MOL TEN-FLUORIDE-SALT REACTORS H. G. MacPherson J. T. Roberts L. G. Alexander D. A, Carrison ABSTRACT The use of a molten-salt fuel makes possible the production of high-pressure, superheated steam with a nuclear reactor aperating at low pressure. The corrosion resistance of the INOR-8 series of nickel-molybdenum alloys appears to be sufficient to guarantee reactor component life- times of 10 to 20 years, Proposed continuous fuel-processing methods show promise of reducing fuel-processing costs to negligible levels. With U233 45 the fuel and Th23245 the fertile material in both core and blanket, initial regeneration ratios up to 1.08 can be obtained at critical masses less than 600 kg. initially about 1300 kg. With U in excess of 0.6 con be obtained with critical masses of less than 300 kg. The corresponding The corresponding inventory for a 600-Mw(th) central station power reactor is 235 45 the fuel, U233 is produced, and initial regeneration ratios critical inventories for 600-Mw(th) central station power reactors are 600 kg or less, depending on the thorium loading. It is concluded that homogeneous, molten-salt-fueled reactors are competi- tive in regard to nuclear performance with present solid-fuel reactors, and they may be economi- cally superior because of lower fuel and fuel-processing costs, Molten fluoride salts provide the basis of a new family of liquid-fueled power reactors, The range of solubility of uranium and thorium compounds in the salts makes the system flexible and allows the consideration of a variety of reactors. Svitable salt mixtures have melting points in the range 850 to 950°F and are sufficiently compatible with known alloys to assure long-lived components, if the temperature is kept below 1300°F. As may be seen, molten-salt-fueled reactor systems tend to operate natyrally in o temperature region suitable for modern steam plants; they have the further advantage that they achieve these temperatures without pressurization of the molten salt, The molten-salt system, for purposes other than electric power generation, is not new. Intensive research and development over the past nine years under ORNL sponsorship has provided reasonable answers to a majority of the obvious difficulties. One of the most important advances has been the development of methods for handling salt melts at high temperatures and maintaining them at tem- peratures above the liquidus temperatures. In- formation on the chemical and physical properties of a wide variety of molten salts has been ob- tained, and methods were developed for their manufacture, purification, and handling that are in use on a production scale. It has been found that the simple ionic salts are stable under radiation and that they suffer no deterioration other than the buildup of fission products. The molten-salt system has the usual benefits attributed to fluid-fueled systems., The principal advantages claimed over solid-fuel systems are: (1) the lack of radiation damage that can limit fuel burnup; (2) the avoidance of the expense of fabri- cating new fuel elements; (3) the continuous re- moval of fission products; (4) a high negative temperature coefficient of reactivity; and (5) the avoidance of the need for excess reactivity, since makeup fuel can be added as required. The last two factors make possible a reactor without control rods that automatically adjusts its power in re- sponse to changes of the electrical load. The lack of excess reactivity con lead to a reactor that is safe from nuclear power excursions. In comparison with aqueous systems, the molten- salt system has three outstanding advantages: it allows high-temperature operation at a low pres- sure; explosive radiclytic gases are not formed; and thorium compounds are soluble in the salts. Compensating disadvantages are the high melting points of the salts and poorer neutron economy; the importance of these disadvantages cannot be assessed properly without further experience. Probably the most outstanding characteristic of the molten-salt systems is their chemical flexi- bility, solutions which are available for reactor use. In this respect, the molten-salt systems are practi- cally unique, and this is the essential advantage which they enjoy over the uranium-bismuth sys- tems. 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 the embodiments which comprise the whole of solid-fuel-element technology: Y233 burners, thorium-uranium thermal converters or breeders, and thorium-uranium fast converters or breeders, Of possible short-term that is, the wide variety of molten-sait interest is the U%35 burner. Because of the in- herently high temperatures and because there are no fuel elements, the fuel cost in the salt system is of the order of 1 mill /kwhr in graphite-moderated, molten-salt-fueled reactors. The present technology suggests that homo- geneous converters using a base salt composed of BeF, and either Li’F or NaF and using UF, for fuel and ThF | for a fertile material are more svit- able for early reactors than are graphite-moderated The chief virtues of the homogeneous converter reactor are that it is or plutonium-fueled systems. based on well-explored principles and that the use of a simple fuel cycle should lead to low fuel cycle costs, With further development, the same base salt (using Li’F) can be combined with a graphite moderator in a heterogeneous arrangement to pro- vide a self-sustaining Th-U23? system with a breeding ratio of about 1. The chief advantage of the molten-salt system over other liquid systems in pursuing this objective is, as has been men- tioned, that it is the only system in which a solu- ble thorium compound can be used, and thus the problem of slurry handling is avoided. PRIOR WORK The applicability of molten salts to nuclear reactors has been ably discyssed by Grimes and 1.2 The most promising systems are those the fluorides and chlorides of the These appear to possess the most desirable combination others. comprising alkali metals, zirconium, and beryllium. of low neutron absorption, high sclubility of ura- inertness. In general, the chlorides have lower melting points, but they appear to be less stable and more cor- rosive than the fluorides. The fluoride systems appear to be preferable for use in thermal and epithermal reactors. Many mix- tures have been investigated, mainly at ORNL and at Mound Laboratory. The physical properties of these mixtures, in so far as they are known, have been tabulated, and the results of extensive phase studies have been reported.’® Lithium-7 has an attractively low capture cross section, 0.0189 barn at 0.0759 ev, but Li®, which comprises 7.5% of the natural mixture, has a cap- ture cross section of 542 barns at the same energy. The cross sections at 0.0759 ev and 1150°F for several lithium compositions are compared below with the cross sections of sodium, potassium, rubidium, and cesium. nium and thorium, and chemical Cross Section Element (barns) Lithium 0.1% Li® 0.561 0.01% Li 0.0731 0.001% Li® 0.0243 0.0001% Li® 0.0194 Sodium 0.290 Potassium 1.130 Rubidium 0.401 Cesium 29 The capture cross sections of the lighter ele- ments at higher energies presumably stand in approximately the same relation as at thermal. |t may be seen that purified Li’7 has an attractively low cross section in comparison with the cross sections of other alkali metals and that sodium is the next best alkali metal. The sodium-zirconium fluoride system has been extensively studied at ORNL.®> A eutectic con- taining about 42 mole % ZrF, melts at 910°F. 'W. R. Grimes, ORNL CF-52-4-197, p 320 ff (April 1952) (classified). 2. R. Grimes, D. R. Cuneo, and F. F. Blankenship, in Reactor Handbook, ed. by J. F. Hogerton and R. C. Grass, vol 2, sec 6, p 799, AECD-3646 (May 1955). 3J. A. Lane, H. G. MacPherson, and F. Maslan (eds.), Fluid Fuel Reactors, p 569, Addison-Wesley, Reading, Mass., 1958. Small additions of UF, lower the melting point appreciably. A fuel of this type was successfully used in the Aircraft Reactor Experiment (ARE).? Inconel, a nickel-rich alloy, is reasonably re- sistant to corrosion by this fuel system at 1500°F. Although long-term data are lacking, there is reason to expect the corrosion rate at 1200°F to be sufficiently low that Inconel equipment would last several years. However, with regard to its use in a central- station power reactor, the sodium-zirconium fluo- ride system has several serious disadvantages. The sodium capture cross section is less favorable than the Li’ cross section. In addition, there is the so-called “*snow’’ problem; that is, ZrF , tends to evaporate from the fuel and crystallize on sur- faces exposed to the vapor. In comparison with the lithium-beryllium system discussed below, the sodium-zirconium system has inferior heat transfer and cooling effectiveness. Finally, the expecta- tion at Oak Ridge is that the INOR-8 alloys will prove to be as resistant to the beryllium salts as to the zirconium salts and that there is therefore no compelling reason for selecting the sodium- zirconium system. The capture cross section of beryllium appears to be satisfactorily low at all energies. A phase diagram for the system LiF-BeF, has recently been published.® A mixture containing 31 mole % BeF, reportedly liquefies at approximately 980°F. Substantial concentrations of ThF, in the core fluid may be obtained by blending this mixture with the compound 3LiF-ThF,. A temperature diagram for the ternary system has been published.® The liquidus temperature along the join appears to lie below 930°F for mixtures containing up to 10 mole % ThF,. Small additions of UF, to any of these mixtures should lower the liquidus tempera- ture somewhat. The ARE was operated with a molten-fluoride- salt fuel in November 1954, The reactor had a moderator consisting of beryllium oxide blocks. The fuel, which was a mixture of sodium fluoride, zirconium fluoride, and uranium fluoride, flowed through the moderator in Inconel tubes and was pumped through an external heat exchanger by bid., p 673. 31bid., p 573. Stbid., p 579. means of a centrifugal pump. The reactor op- erated at a peak power of 2.5 Mw. It was dis- mantled after carrying out a scheduled experi- mental program. In 1953 a group of QRSORT students, under the leadership of Jarvis,” investigated the applica- bility of molten salts to package reactors., More recently, another ORSORT group led by Davies prepared a valuable study of the feasibility of molten-salt U232 burners for central-station power production.® Fast reactors based on the U?38.py cycle were studied by Addoms ez al. of MIT and by an ORSORT group ied by Bulmer.® Both groups concluded that it would be preferable to use molten chlorides, rather than the fluorides, because of the relatively high moderating power of the fluorine nucleus, although it was recognized that the chlo- rides are probably inferior with respect to corro- sion and radiation stability. Bulmer et al. also pointed out that it would be necessary to use puri- fied CI®7 on account of the (n,p) reaction ex- hibited by CI33, Because of the disadvantages of the chloride systems and, further, because the technology of handling and utilizing neptunium- and plutonium-bearing salts is largely unknown, it was decided to postpone consideration of chloride salt reactors. In 1953 an ORSORT group led by Wehmeyer!'® analyzed many of the problems presently under study. The proposals set forth in that report have influenced the present program. A study by David- son and Robb of KAPL!! has also been helpful. Both studies considered the possibility of using thorium in a U233 conversion-breeding cycle at thermal or near thermal energies. A recent conceptual design study'? of a 240-Mw (electrical) central-station molten-salt-fueled re- actor was used as a basis for examining the eco- nomics and feasibility of a reactor using molten- salt fuel. An attempt was made to keep the 7T. Jarvis et al., ORNL CF-53-10-26 (August 1953) (classified). 8R. W. Davies et al, ORNL CF-56-8-208 (August 1956) (classified). 9J. Bulmer et al., Fused Salt Fast Breeder, QORNL CF-56-8-204 (August 1956). IOD. B. Wehmeyer et al., Study of a Fused Salt Breeder Reactor for Power Production, ORNL CF-53- 10-25 (September 1953). ”J. K. Davidson and W. L. Robb, A Molten-Salr Thorium Converter for Power Production, KAPL-M-JKD- 10 (Oct. 15, 1956). ]2ane, MacPherson, and Maslan, op. cit., p 681. technology and the processing scheme as simple as possible, METHOD OF CALCULATION Reactor calculations were performed by means of the UNIVAC program Ocusol, '3 a modification of the Eyewash program.'® Ocusol is a 31-group, multiregion, spherically symmetric, age-diffusion code. The group-averaged cross sections for the various elements of interest that were used were based on the latest available data.!® Where data were |acking, reasonable interpolations based on resonance theory were used. The estimated cross sections were made to agree with measured reso- nance integrals where available. Saturations and Doppler broadening of the resonances in thorium as a function of concentration were estimated. The molten salts may be used as homogeneous moderators or simply as fuel carriers in hetero- geneous reactors, Although graphite-moderated heterogeneous reactors have certain potential ad- vantages, their technical feasibility depends upon the compatibility of fuel, graphite, and metal, which has not as yet been established. For this the homogeneous reactors, although in- ferior in nuclear performance, have been given prior attention. A preliminary study indicated that, if the in- tegrity of the core vessel could be guaranteed, reason, the nuclear economy of two-region reactors would probably be superior to that of bare and reflected one-region reactors, The two-region reactors were, accordingly, studied in detail. Although entrance and exit conditions dictate other than a spherical shape, it was necessary, for the calculations, to use a model comprising the following concentric, (1) the core; (2) an INOR-8 reactor vessel, ‘/3 in, thick; (3) a blanket, approxi- mately 2 ft thick; and (4) an INOR-8 reactor vessel, 2/3 in. thick. The diameter of the core and the con- spherical regions: centration of thorium in the core were selected as independent variables, The primary dependent variables were the critical concentration of the 3L, G. Alexander et al., Operating Instructions for the Univac Program Ocusol-A, a Modification of the Eyewash Program, ORNL CF-57-6-4 (June 5, 1957). MJ. H. Alexander and N. D. Given, A Machkine Multi- group Calculation. The Eyewash Program for Univac, ORNL-1925 (Sept. 15, 1955). ISJ. T. Roberts and L. G. Alexander, Cross Sections for Ocusol-A Program, ORNL CF-57-6-5 (June 11, 1957). fuel (U233, U233, or Pu?%%) and the distribution of the neutron absorptions among the various From these, the inventory, regeneration ratio, burnup rate, etc., could be readily calculated. atomic species in the reactor. critical mass, critical HOMOGENEOUS REACTORS FUELED WITH U235 U233 would be a superior fuel While the isotope in molten-fluoride-salt reactors, it is unfortunately not available in quantity, Any realistic appraisal of the immediate capabilities of these reactors must be based on the use of U233, The study of homogeneous reactors was divided into two phases: (1) the mapping of the nuclear characteristics of the initial (i.e., ‘‘clean’’) states as a function of core diameter and thorium con- centration and (2) the analysis of the subsequent performance of selected initial states with various processing schemes and rates. The detailed re- sults of the first phase are given here. Briefly, it was found that regeneration ratios of up to 0.65 could be obtained with moderate investment in U235 (less than 100 kg). Initial States A complete parametric study was made of molten- fluoride-salt reactors having diometers in the range of 4 to 10 ft and thorium concentrations in the fuel ranging from 0 to 1 mole % ThF ,. [n these re- actors the basic fuel salt (fuel salt No. 1) was a mixture of 31 mole % BeF., and 69 mole % LiF, which has a density of about 2.0 g/cm® at 1150°F, The core vessel was composed of INOR-8. The blanket fluid (blanket salt No. 1) was a mixture of 25 mole % ThF , and 75 mole % LiF, which has a density of about 4.3 g/cm® at 1150°F. In order to shorten the calculations in this series, the re- actor vessel was neglected, since the resuitant error would be small. These reactors contained no fission products or nonfissionable isotopes of uranium other than U238, A summary of the results is presented in Table 1, in which the neutron balance is presented in terms of neutrons absorbed in a given element per neu- tron absorbed in U233 [both by fission and the (,y) reaction]. The sum of the absorptions is therefore equal to 7, that is, the number of neutrons pro- duced by fission per neutron absorbed in fuel. Further, the sum of the absorptions in U238 and thorium in the fuel and in thorium in the blanket Table 1. Initial-State Nuclear Characteristics of Two-Region, Homogeneous, Moiten-Fluoride-5alt Reactors Fueled with U Fuel salt No, 1: 31 mole % BeF2 + 69 mole % LiF + UF’4 + ThF4 Blanket salt No. 1: 25 mole % ThF4 + 75 mole % LiF 600 Mw (heat) External fuel volume: 339 ft° Total power: 235 Case number Core diameter, ft ThF4 in fuel salt, mole % U235 in fuel salt, mole % U 235 3tom den sity* Critical mass, kg of y23s Neutron absorption ratios** U233 (fissions) U235 (n,y) Be, Li, and F in fuel saolt Core vessel L.i and F in blanket sait Leakage U238 Th in fuel salt in fuel salt Th in blanket salt Neutron yield, i Median fission energy, ev Thermal fissions, % Regeneration ratio 1 2 3 4 5 6 7 8 9 10 11 4 5 5 5 5 5 6 6 6 6 6 0 0 0.25 0.5 0.75 1 0 0.25 0.5- 0.75 1 0.952 0.318 10.561 0.721 0.845 0.938 0.107 0.229 0.408 - 0.552 0.662 33.8 11.3 20.1 25.6 30.0 33.3 3.80 8.13 14.5 19.6 23.5 124 81.0 144 183 215 239 47.0 101 179 243 291 0.7023 0.7185 0.7004 0.6996 0.7015 0.7041 0.7771 0.7343 0.7082 0.7000 0.7004 0.2977 0.2815 0.2996 0.3004 0.2985 0.2959 0.2229 0.2657 0.2918 0.3000 0.2996 0.0551 0.0871 0.0657 0.0604 0.0581 0.0568 0.1981 0.1082 0.0770 0.0669 0.0631 0.0560 0.0848 0.0577 0.0485 0.0436 0.0402 0.1353 0.0795 0.0542 0.0435 0.0388 0.0128 0.0138 0.0108 0.0098 0.0093 0.00%90 0.0164 0.0116 0.0091 0.0081 0.0074 0.0229 0.0156 0.0147 0.0143 0.0141 0.0140 0.0137 0.0129 0.0122 0.0119 0.0116 0.0430 0.0426 0.0463 0.0451 0.0431 0.0412 0.0245 0.0375 0.0477 0.0467 0.0452 0.0832 0.1289 0.1614 0.1873 0.1321 0.1841 0.2142 0.2438 0.5448 0.5309 0.4516 0.4211 0.4031 0.3905 0.5312 0.4318 0.3683 0.3378 0.3202 1.73 1.77 1.73 1.73 1.73 1.74 1.92 1.82 1.75 1.73 1.73 270 15.7 105 158 270 425 0.18 5.6 38 100 120 0.052 6.2 0.87 0,22 0.87 0.040 35 13 3 0.56 0.48 0.59 0.57 0.58 0.60 0.61 0.62 0.56 0.61 0.60 0.60 0.61 * 1019 cxfoms/cms. **Neutrons absorbed per neutron absorbed in u23 5 Table 1 (continued) Case number Core diameter, ft ThF4 in fuel salt, mole % U2:35 in fuel salt, mole % U235 atom density* Critical mass, kg of U235 Neutron absorption ratios** U235 (fission) U235 (n,y) Be, Li, and F in fuel salt Core vessel Li and F in blonket salt Leakage 238 u in fuel salt Th in fuel salt Th in blanket salt Neutron yield, 7 Median fission energy, ev Thermal fissions, % Regeneration ratio 12 13 14 15 16 17 18 19 20 21 22 7 8 8 8 8 8 10 10 10 10 10 0.25 0 0.25 0.5 0.75 1 0 0.25 0.5 0.75 1 0.114 0.047 0.078 0.132 0.226 0.349 0.033 0.052 0.081 0.127 0.205 4.05 1.66 2.77 4.67 8.03 12.4 1.175 1.86 2.88 4.50 7.28 79.6 48.7 81.3 137 236 364 67.3 107 165 258 417 0.7748 0.8007 0.7930 0.7671 0.7362 0.7146 0.8229 0.7428 0.7902 0.7693 0.7428 0.2252 0.1993 0.2070 0.2329 0.2638 0.2854 0.1771 0.2572 0.2098 0.2307 0.2572 0.1880 0.4130 0.2616 0.1682 0.1107 0.0846 0.5713 0.3726 0.2486 0.1735 0.1206 0.0951 0.149 0.1032 0.0722 0.0500 0.0373 0.129 0.0915 0.0669 0.0497 0.0363 0.0123 0.0143 0.0112 0.0089 0.0071 0.0057 0.0114 0.0089 0.0073 0.0060 0.0049 0.0068 0.0084 0.0082 0.0080 0.0077 0.0074 0.0061 0.0060 0.0059 0.0057 0.0055 0.0254 0.0143 0.0196 0.0272 0.0368 0.0428 0.0120 0.0153 0.0209 0.0266 0.0343 0.1761 0.2045 0.3048 0.3397 0.3515 0.2409 0.3691 0.4324 0.4506 0.4098 0.4073 0.3503 0.3056 0.2664 0.2356 0.3031 0.2617 0.2332 0.2063 0.1825 191 2.00 1.96 1.89 1.82 1.76 2.03 2.00 1.95 1.90 1.83 0.16 Thermal 0.10 0.17 5.3 27 Thermal Thermal 0.100 0.156 1.36 33 59 45 29 13 5 66 56 43 30 16 0.61 0.42 0.57 0.64 0.64 0.63 0.32 0.52 0.62 0.67 0.67 * 1019 cfloms/cms. **Neutrons absorbed per neutron absorbed in U23 5 . salt give directly the regeneration ratio. The losses to other elements are penalties imposed on the regeneration ratio by these poisens. A graph of critical mass plotted as a function of core diameter, with thorium concentration as a parameter, is presented in Fig. 1. The masses vary from about 40 kg of U?3°% in a 7-ft-dia core having no thorium in the fuel to about 450 kg in the 10-ft-dia core having 1 mole % ThF, in the UNCLASSIFIED ORNL—LR—DWG 39521 SOO ¥ T T T l CORE AND BLANKET SALTS NOt mole % Tth ! IN FUEL. SALT ® ° - g ¢ " ND // s @ 300 ra . ; / O.T?___________-u < '/"-__'"—'—-—Ii—-—-"'—"" = ¢ &J 200 28 | i, 9" e 2 2 N 0.50 o E ” ® -..____"—/T—/‘ © I‘r"\ \ 100 " '." 0.25 . ! \ ® | — e / ' NO ThF, IN FUEL SALT 0 | i CORE DIAMETER () Fig. 1. |Initial Critical Masses of U235 iy Two- Region, Homogeneous, Molten-Fluoride-Salt Reactors., tuel. plotted in Fig. 2, range from 0.5 for the minimum mass reactor to 0,63 for the largest mass reactor. It does not seem likely that further increases in diameter or thorium concentration would increase the regeneration above 0.65. The effects of changes in the compositions of the fuel and blanket salts were studied in a series of calculations for salts having more favorable The corresponding regeneration ratios, melting points and viscosities. The BeF, content was raised to 37 mole % in the fuel salt (fuel salt No. 2), and the blanket composition (blanket salt No. 2) was fixed at 13 mole % ThF4, 16 mole % Ber, and 71 mole % LiF. Blanket salt No. 2 is a somewhat better reflector than No. 1, and fuel salt No. 2 is a somewhat better moderator than No. 1. As a result, at a given core diameter and thorium UNCLASSIFIED ORNL-LR—-DWG 39522 1.0 T T T T CORE AND BLANKET SALTS NO.1 0.8 mole % ThF, IN FUEL SALT o 1 AND 0.75 = | e = AV e z 06 =2 + 0.5 o e 0.25 < o > w > * NO ThF, IN FUEL SALT/ 0.2 ' 0 q 5 6 7 8 S 10 CORE DIAMETER (ft) Fig. 2. [Initial Fuel Regeneration in Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U235, concentration in the fuel salt, both the critical concentration and the regeneration ratio were somewhat lower for the No. 2 salts. Reservations concerning the feasibility of con- structing and guaranteeing the integrity of core vessels in large sizes (10 ft and over), together with preliminary consideration of inventory charges for large systems, led to the conclusion that a feasible reactor would probably have a core di- ameter lying in the range between 6 and 8 ft. Ac- cordingly, a parametric study of the No. 2 fuel and blanket salts in reactors with core diameters in the 6- to 8-ft range was made. |n this study the presence of an outer reactor vessel consisting of 2/3 in. of INOR-8 was taken into account. The results are presented in Table 2. In general, the nuclear performance is somewhat better with the No. 2 salts than with the No. 1 salts, Neutron Balances and Qther Reactor Variables The distributions of the neutron captures are given in Tables 1 and 2, where the relative hard- ness of the neutron spectrum is indicated by the median fission energies and the percentages of thermal fissions. It may be seen that losses to lithium, beryllium, and fluorine in the fuel salt and to the core vessel are substantial, especially in the more thermal reactors (e.g., case No. 18). However, in the thermal reactors, losses by radia- tive capture in U239 are relatively low. Increasing Table 2. {initial-State Nuclear Characteristics of Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U23 S Fuel salt No. 2: 37 mole % BeF2 + 63 mole % LiF + UF, + ThFd Blanket salt No. 2: 13 mole % ThF4 + 16 mole % BeF, + 71 mole % LiF Total power: 600 Mw (heot) External fuel volume: 339 f3 Case number Core diameter, ft Thl:4 in fuel salt, mole % U235 in fuel salt, mole % U235 atom density* Critical mass, kg of U235 Neutron absorption ratios** U235 y235 (n,y) Be, Li, and F in fuel salt (fissions) Core vessel Li eand F in blanket salt Quter vessel L eakage U238 in fuel salt Th in fuel salt Th in blanket salt Neutron yield, 77 Median fission energy, ev Thermal fissions, % Regeneration ratio 23 24 25 26 27 28 29 30 31 32 33 34 6 6 6 6 7 7 7 7 8 8 8 8 0.25 0.5 0.75 1 0.25 0.5 0.75 1 0.25 0.5 0.75 1 0.169 0.310 0.423 0.580 0.084 0.155 0.254 0.366 0.064 0.099 0.163 0.254 5.87 10.91 15.95 20.49 3.13 3.38 8.70 13.79 2.24 3.51 5.62 9.09 72.7 135 198 254 61.5 106 171 271 65.7 103 165 267 0.7516 0.7174 0.7044 0.6958 0.7888 0.7572 0.7282 0.7094 0.8014 0.7814 0.7536 0.7288 0.2484 0.2826 0.2956 0.3042 0.2%112 0.2428 0.2718 0.2906 0.1986 0.2186 0.2464 0.2712 0.1307 0.0900 0.0763 0.0692 0.2147 0.1397 0.1010 0.0824 0.2769 0.1945 0.1354 0.1016 0.1098 0.0726 0.0575 0.0473 0.1328 0.0905 0.0644 0.0497 0.1308 0.0967 0.0696 0.0518 0.0214 0.0159 0.0132 0.0117 0.0215 0.0167 0.0131 0.0108 0.0198 0.0162 0.0130 0.0105 0.0024 0.0021 0.0021 0.0019 0.0019 0.0018 0.0016 0.0015 0.0017 0.0016 0.0014 0.0013 0.0070 0.0065 0.0064 0.0061 0.0052 0.0050 0.0048 0.0045 0.0045 0.0043 0.0042 0.0040 0.0325 0.0426 0.0452 0.0477 0.0214 0.0307 0.0392 0.0447 0.0177 0.0233 0.0315 0.0392 0.1360 0.1902 0.2212 0.2387 0.1739 0.2565 0.2880 0.3022 0.1978 0.3043 0.3501 0.3637 0.4165 0.3521 0.3178 0.2962 0.3770 0.3294 0.2866 0.2566 0.3240 0.2892 0.2561 0.2280 1.86 1.77 1.74 1.72 1.95 1.87 1.80 1.75 1.97 1.93 1.86 1.80 0.480 10.47 58.10 76.1 0.1223 0.415 7.61 25.65 51% th 0.136 0.518 7.75 21 7 2.8 0.84 43 24 11 4.3 51 38 23 11 0.59 0.58 0.58 0.58 0.57 0.62 0.61 0.60 0.54 0.62 0.64 0.63 *10]9 cn‘oms/cm3. **Neutrons absorbed per neutron absorbed in u23s, the hardness decreases losses to salt and core vessel sharply (case No. 5) but increases the loss to the (n,y) reaction. The numbers given for capture in the lithium and fluorine in the blanket show that these elements are well shielded by the thorium in the blanket, and the leakage values show that leckage from the reactor is less than 0.01 neutron per neutron absorbed in U235 in re- actors over 6 ft in diameter. The blanket con- tributes substantially to the regeneration of fuel, accounting for not less than one-third of the total, even in the 10-ft-dia core containing 1 mole % ThF . HOMOGENEOUS REACTORS FUELED WITH u233 Uranium-233 is a superior fuel for use in molten- fluoride-salt reactors in almost every respect. The fission cross section in the intermediate range of neutron energies is greater than the fission cross section of U23°. Thus initial critical inventories are less, and less additional fuel is required to override poisons. Also, the parasitic cross sec- tion is substantially less, and fewer neutrons are lost to radiative capture. Further, the radiative captures result in the immediate formation of a fertile isotope, U234, The rate of accumulation of U236 is orders of magnitude smaller than with U233 as a fuel, and the buildup of Np237 and Pu23? is negligible. The mean neutron energy is somewhat nearer thermal in such reactors than it is in the cor- responding U235 cases. Consequently, losses to core vessel and to core salt tend to be higher. Both losses are reduced substantially at higher thorium concentrations because of the hardening of the neutron spectrum, Results from a parametric study of the nuclear characteristics of two-region, homogeneous, molten- fluoride-salt reactors fueled with U233 are given in Tables 3 and 4, The core diameters considered range from 3 to 12 ft, and the thorium concentra- tions range from 0.25 to 7 mole %. The regenera- tion ratios are very good compared with those ob- tained with U235, With 7 mole % ThF, in an 8-ft-dia core, the U233 critical mass was 1500 kg, and the regeneration ratio was 1,09, The data in Table 3 are for reactors using fuel and blanket salts No. 1 with ThF concentrations ranging up to 1 mole %. The critical masses are graphed in Fig. 3 and the regeneration ratios in Fig. 4. The masses range from a minimum of about 20 kg in a 5-ft-dia core, with no thorium present, to 130 kg in a 10-ft-dia core having 1 mole % thorium in the fuel. The corresponding regenera- tion ratios are 0.60 and 0.90. For a given thorium UNCLASSIFIED ORNL-LR-DWG 39523 140 I w 1 CORE AND BLANKET SALTS NO. 4 * o 0 / T/ mole %o ThF, IN FUEL SALT, 4 ! . 5 & 075 o o) 4 = 4 60 . a L = 5 | / / 40 \../ / NO Tth—1 ~— "IN FLEL SALT *— ; / + 20 R ] : | 0 2 4 6 8 10 12 CORE DIAMETER {f Fig. 3. Critical Masses of Two«Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U233, UNCLASSIFIED ORNL-LR~DWG 39524 1.2 | I ! [ CORE AND BLANKET SALTS NO. ‘ mele % ThF, 1.0 IN FUEL SALT Q 5 08 & = 2 o = 06 |59 Z i < w « 04 NO ThFg IN FUEL SALT g2 1 | 0 3 4 5 6 7 8 9 10 CORE DIAMETER (ft} Fig. 4. Initial Fuel Regeneration in Two-Region, Homogeneous, Molten-Fluoride«Salt Reactors Fueled with U233, ] Table 3. Initial-State Nuclear Characteristics of Two-Region, Hemogeneous, Molten-Fluoride-Salt Reactors Fueled with U Fuel salt No. 1: 31 mole % BeF2 + 69 mole % LiF + UF4 + ThF4 Blanket salt: 25 mole % ThF4 + 75 mole % LiF Total power: 600 Mw (heat) External fuel volume: 339 #3 233 Case number Core diameter, ft ThF4 in fuel salt, mole % U233 U Critical mass, kg of U233 in fuel salt, mole % 233 atom density* Neutron absorption ratios™** U233 (fissions) U233 (s Be, Li, and F in fuel soit Core vessel Li and F in blanket salt Leakage Th in fuel salt Th in blanket salt Neutron yield, n Median fission energy, ev Thermal fissions, % Regeneration ratio 35 36 37 38 39 40 41 42 43 44 45 46 3 4 4 5 5 5 5 5 6 6 6 6 0 0 0.25 0 0.25 0.5 0.75 1 0 0.25 0.5 0.75 0.592 0.158 0.233 0.076 0.106 0.141 0.179 0.214 0.048 0.066 0.087 0.113 21.1 5.6 8.26 2.7 3.73 5.0 6.35 7.605 1.7 2.3 3.1 4.0 64.9 22.3 30.3 19.3 26.9 35.8 45.5 54.5 20.4 29.2 38.4 49.5 0.8754 0.8706 0.8665 0.8767 0.8725 0.8684 0.8674 0.8672 0.8814 0.8779 0.8744 0.8665 0.1246 0.1294 0.1335 0.1233 0.1275 0.1316 0.1326 0.1328 0.1186 0.1221 0.1256 0.1335 0.063%9 0.1051 0.0860 0.1994 0.1472 0.1174 0.1010 0.0905 0.3180 0.2297 0.1774 0.1412 0.0902 0.1401 0.1093 0.1808 0.1380 0.1112 0.0944 0.0821 0.1983 0.1508 0.1209 0.0989 0.0233 0.0234 0.0203 0.0232 0.0196 0.0172 0.0157 0.0146 0.0215 0.0179 0.0157 0.0139 0.0477 0.0310 0.0306 0.0197 0.0193 0.0190 0.0189 0.0188 0.0160 0.0157 0.0157 0.0154 0.0000 0.0000 0.1095 0.0000 0.1593 0.2561 0.3219 0.3702 0.0000 0.1973 0.3111 0.3989 0.9722 0.8857 0.8193 0.7777 0.7066 0.5487 0.6255 0.6004 0.6586 0.5922 0.5539 0.5169 2.1973 2.1853 2.1750 2.2007 2.1900 2.1797 2.1773 2.1766 2.2124 2.2035 2.1948 2,185 174 14.2 19.1 1.752 2.87 9.625 16.5 29.35 0.326 1.18 2.175 10.16 0.0527 7.952 2.970 24.80 16.499 10.09 5.99 3.192 37.832 29.37 27.12 14.87 0.9722 0.8856 0.9288 0.7777 0.8659 0.9148 0.9474 0.9706 0.5486 0.7895 0.8651 0.9158 *]0]9 otoms/cm3. **Neutrons absorbed per neutron absorbed in U23 . 3 tH Table 3 {continued) Cose number Core diameter, ft ThF4 in fuel salt, mole % U233 in fuel salt, mole % U233 atom density* Critical mass, kg of y233 Neutron absorption ratios** U233( U233( fissions) n,y) Be, Li, and F in fuel salt Core vessel i and F in blaonket salt L eakage Th in fuel salt Th in blanket salt Neutron yield, 7 Median fission energy, ev Thermal fissions, % Regeneration ratio 47 48 49 50 51 52 53 54 55 56 57 6 8 8 8 8 8 10 10 10 10 10 1 0 0.25 0.5 0.75 1 0 0.25 0.5 0.75 1 0.133 0.028 0.039 0.052 0.066 0.0478 0.022 0.031 0.041 0.051 0.063 4,72 1.01 1.41 1.85 2.33 2,72 0.780 1.09 1.45 1.8 2.25 58.4 29.6 41.1 54,3 68.4 86.6 44.7 63.0 83.1 103.2 131.3 0.8693 0.8876 0.8850 0.8808 0.8779 0.8755 0.8921 0.8881 0.8842 0.8814 0.8781 0.1307 0.1124 0.1150 0.1192 0.1221 0.1245 0.1079 0.1119 0.1158 0.1186 0.1219 0.1216 0.5433 0.3847 0.2896 0.2285 0.1899 0.7166 0.5037 0.3758 0.2952 0.2360 0.0855 0.1866 0.1406 0.1112 0.6915 0.0778 0.1560 0.1168 0.0919 0.0754 0.0629 0.0127 0.0176 0.0141 0.0120 0.0106 0.0095 0.0133 0.0108 0.0091 0.0080 0.0071 0.0152 0.0095 0.0095 0.0093 0.0091 0.0090 0.0068 0.0068 0.0066 0.0065 0.0065 0.4580 0.0000 0.2513 0.4044 0.5055 0.5768 0.0000 0.2852 0.4585 0.5708 0.6507 0.4889 0.4707 0.4211 0.3842 0.3582 0.3344 0.3466 0.3058 0.2774 0.2564 0.2408 2.1820 2.2277 2.2212 2.2108 2.2035 2,1975 2.2392 2,2290 2.2194 2.2133 2.2040 8.51 52% th 0.197 0.4915 1.185 1.12 58% th 50% th 0.1735 0.455 3.25 12.42 51.93 43.398 35.79 29.078 24.36 58.34 50.39 42.8 36.45 29.96 0.9470 0.4707 0.6725 0.7886 0.8638 0.9112 0.3467 0.5910 0.7359 0.8271 0.8915 *10]9 afoms/cm3. **Neutrons absorbed per neutron absorbed in U2:3 3 cl Table 4. Initial-State Nuclear Characteristics of Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U Fuel salt No. 2: 37 mole % BeF, + 63 mole % LiF + UF, + ThF, Blanket salt No. 2: 13 mole % ThFd + 16 mole % BeF, + 71 mole % LiF Total power; 600 Mw (heat) External fuel volume: 339 ft 233 Case number Core diameter, ft Th4 in fuel salt, mole % U233 in fuel salt, mole % U233 atom density* Critical mass, kg of U233 Neutron absorption ratios** U233 U233 (fissions) (n,y) Be, Li, and F in fuel salt Core vessel Li and F in blanket salt L eakage Th in fuel salt Th in blanket salt Neutron yield, 17 Median fission energy, ev Thermal fissions, % Regeneration ratio 58 59 60 61 62 63 64 65 66 67 68 69 70 6 6 6 6 7 7 7 7 8 8 8 8 4 0.25 0.5 0.75 } 0.25 0.5 0.75 1 0.25 0.5 0.75 1 2 0.062 0.081 0.10 0.121 0.047 0.059 0.074 0.091 0.039 0.049 0.062 0.075 0.619 1.98 2,6 3.2 3.88 1.5 1.92 2.38 2.9 1.21 1.58 1.97 2.41 19.8 24.51 32.19 39.62 48.03 29.49 37.75 46,79 57.01 35.51 46.37 57.82 70.73 72.0 0.8805 0.8762 0.8741 0.8722 0.8843 0.8809 0.8784 0.8749 0.8880 0.8828 0.8809 0.8827 0.871 0.1195 0.1238 0.1259 0.1278 0.1157 0.1191 0.1216 0.12517 0.1120 GQ.1172 G.1191 0.1173 0.129 0.2427 0.1915 0.1604 0.1383 0.3209 0.2525 0.2062 0.1735 0.2407 0.3051 0.2458 0.2073 0.070 0.1891 0.1526 0.1288 0.1109 0.1858 0.1505 0.1258 0.1070 0.1756 0.1405 0.1168 0.1003 0.073 0.0313 0.0272 0.0243 0.0221 0.0276 0.0238 0.0211 0.0190 0.0247 0.0212 0.0187 0.0169 0.025 0.0133 0.0111 0.0109 0.0108 0.0094 0.0081 0.0080 0.0078 0.0070 0.0069 0.0068 0.0068 0.031 0.1901 0.3088 0.3902 0.4504 0.2182 0.3531 0.4455 0.5125 0.3952 0.3891 0.4903 0.5678 0.343 0.5454 0.5079 0.4794 0.4566 0.4589 0.4228 0.3983 0.3763 0.3859 0.3533 0.3325 0.3164 0.653 2.2100 2.1992 2.1940 2.1891 2.2197 2.2110 2.2049 2.1960 2,2289 2.2160 2.2110 2.2155 2.195 0.721 1.575 2.475 3.685 0.1875 0.465 0.992 2.025 0.1223 0.230 0.676 1.345 147 33.878 26,269 20.518 15.584 41.997 35.191 28.685 23.051 47.965 40.663 33.87 28.301 0.23 0.7355 0.8167 0.8695 0.9071 0.6770 0.7760 0.8438 0.8887 0.6264 0.7424 0.8228 0.8842 0.996 * ]0]9 a'roms/cmz. **Neutrons absorbed per neutron absorbed in U 233 £l Table 4 (continued) Case number Core diameter, ft ThF4 in fuel salt, mole % U233 in fuel salt, mole % U233 atom density™* Critical mass, kg of y233 Neutron absorption ratios** U233 ( U fissions) 233 (n,y) Be, Li, and F in fuel salt Core vessel Li end F in blanket salt Ledkage Th in fuel salt Th in blanket salt Neutron yield, 1 Median fission energy, ev Thermal fissions, % Regeneration ratio 71 72 73 74 75 76 77 78 79 80 81 82 83 84 4 4 6 6 6 8 8 8 10 10 10 12 12 12 4 7 2 4 7 2 4 7 2 4 7 2 4 7 0.856 1.247 0.236 0.450 0.762 0.152 0.316 0.603 0.121 0.262 0.528 0.101 0.222 0.477 27.4 39.9 7.55 14.4 24.4 4.88 10.1 19.3 3.86 8.3¢9 16.9 3.24 7.39 15.25 100.5 146.5 94.2 177.8 301 143 299 566 221 481 970 320 732 1510 0.874 0.881 0.864 0.868 0.876 0.867 0.865 0.873 0.870 0.864 0.871 0.873 0.864 0.870 0.126 0.119 0.136 0.132 0.124 0.133 0.135 0.127 0.130 0.136 0.129 0.127 0.136 0.130 0.066 0.069 0.093 0.075 0.076 0.120 0.082 0.078 0.142 0.088 0.081 0.164 0.093 0.083 0.059 0.048 0.068 0.049 0.035 0.057 0.037 0.025 0.046 0.030 0.018 0.033 0.022 0.012 0.021 0.019 0.016 0.014 0.011 0.019 0.012 0.009 0.009 0.006 0.006 0.012 0.004 0.004 0.031 6.028 0.017 0.017 0.015 0.010 0.010 0.009 0.007 0.008 0.0067 0.004 0.006 0.004 0.426 0.517 0.581 0.650 0.740 0.716 0.785 0.865 0.800 0.865 0.938 0.872 0.922 0.998 0.600 0.538 0.403 0.382 0.330 0.264 0.254 0.2113 0.189 0.170 0.146 0.115 0.130 0.092 2,203 2,219 2.1770 2.187 2.2072 2.1860 2.180 2.200 2.1933 2.177 2.196 2.1995 2.177 2.1931 503 1085 24.2 64.0 443 8.42 51.3 243 3.64 45.9 193 2.45 41.4 178 0.11 0.084 4.3 0.35 0.076 11.0 1.5 0.091 17 1.8 0.12 21 2.1 0.15 1.026 1.055 0.984 1.032 1.070 0.980 1.039 .1.078 0.989 1.045 1.084 0.987 1.052 1.090 * ]0]9 afoms/cm3 **Neutrons absorbed per neutron absorbed in U23 3 concentration, the regeneration ratio tends to in- crease with decreasing core size, and ratios up to 0.97 were observed in this series of caiculations, as shown in Fig. 4. The data in Table 4 are for reactors using fuel and blanket salts No. 2. in this series of calcula- tions, the diameter ranged up to 12 ft and the thorium concentration in the core up to 7 mole %. [t was necessary to alter progressively the com- position of the base salt as the thorium concentra- tion was increased in order to keep the liquidus temperature below 1000°F. There was a slight increase in concentration of LiF at the expense of BeF, For cores having thorium concentrations in the range from 0.25 to 1 mole %, the results are about the same as those obtained with cores using No. 1 salts. The behavior with No. 2 salts at higher concentrations of ThF, is shown in Figs. 5 and 6. It is seen that an initial regeneration ratio of about 1.0 can be achieved with about 2,5 mole % thorium in the fuel, regardless of the diameter of UNCLASSIFIED ORNL-LR-DWG 39525 2000 | | —e— CALCULATED . ~e == INTERPOLATED Pyl il I T A i s L7 7 + - ” rad 500 |- — " o = 5 200 £ &) oy L=t = 3 100 [ = & o 50 20 f | CORE AND BLANKET SALTS NO. 2 t 10 ‘ | \ 2 4 6 8 10 12 CORE DIAMETER (ff) Fig. 5. Critical Masses of Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U233. 14 UNCLASSIFIED ORNL-LR--DWG 39526 14 : ! | | mole 7o ThE, IN FUEL SALT, 7 o 8= _] .—-"""—_— /i“":6 - - 5 - ::"" - 44/0/“ _"T——‘Ta-——"””_— o g |\¢ ¢ "-_____-0 x 3 2 | 5 ~~-. L ‘ ‘ 1.5 =z i % CORE AND BLANKET SALTS NC. 2 & ‘ ® 09 1 T \'1 . —e— CALCULATED \. — — INTERPOLATED \ ! $0.75 08 | o 2 4 6 8 10 12 CORE DIAMETER (f) Fig. 6. |Initial Regeneration Ratic in Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U233, the core in the range from 4 to 12 ft. The corre- sponding critical masses range from about 80 to 400 kg of U233, NUCLEAR PERFORMANCE OF A REFERENCE DESIGN REACTOR A conceptual design study of a 240-Mw (electri- cal) central-station molten-salt-fueled reactor (MSR) was described by the Molten-Salt Reactor Group at ORNL.'® The system employs a two-region homo- geneous reactor having a core approximately 8 ft in diameter and a blanket 2 ft thick. The core, with its volume of 113 ft3, is capable of generating 600 Mw of heat at a power density in the core of 187 w/em>®. The general arrangement of the core and blanket is shown in Fig, 7. The basic core salt is a mixture of lithium, beryllium, and thorium fluorides, together with sufficient fluoride of UZ3% or U233 to make the system ]6.]. A. l.ane, H. G. MacPherson, and F. Maslan (eds.), Fluid Fuel Reactors, p 569, Addison-Wesley, Reading, Mass., 1958. SECTION A-A SIPHON DRAIN FUEL LINE TO HEAT EXCHANGER 2 4q —=_=C | FEET FUEL RETURN {——- A UNCLASSIFIED ORNL-LR~DWG 286364 ! BLANKET | PUMP FUEL PUMP MOTOR MOTOR ‘ BLANKET EXPANSION ‘ TANK E : LT = \ - % T / ..7, \\_— e FUEL EXPANSION TANK BREEDING BLANKET BLANKET RETURN Fig. 7. General Arrangement of Core and Blanket, 15 critical. The blanket contains ThF ,, either as the eutectic of LiF and ThF4, or mixtures of it with the basic core salt. The liquidus temperature of the fuel salt is about 850°F and that of the blanket is 1080°F or lower, Both the core fuel and the blanket salt are circu- lated to external heat exchangers, six in parallel for the core and two in parallel for the blanket. The heat is transferred by intermediate fluids from these heat exchangers to the boilers, superheaters, and reheaters. The heat transfer system is de- signed so that, with a fuel temperature of 1200°F, a steam temperature of 1000°F ot 1800 psi can be achieved. The volume of fuel salt external to the core in the transfer lines, pumps, and heat exchangers was estimated to be 339 ft3. 1t is this external volume that largely determines the fuel inventory of the system. A parametric study of the regeneration ratio as a function of critical inventory in this system was performed. With U?3% in reactors em- ploying core and blanket salts No. 1, the results are as shown in Fig. 8, where regeneration ratio is plotted vs critical inventory, with thorium con- centration in the fuel as a parameter, The numbers associated with the plotted points are the di- ameters of the cores. The curves are observed to peak rather sharply, and these peaks define a locus of maximum re- generation ratio for a given inventory. It may be UNCLASSIFIED ORNL~LR-DWG 39527 08 | ] | 07 e — s) I ‘ ° | e .0 mole % ThF, INFUEL SALT © g 8 B L 10L” ~—0 5 | 5 2 g6 |- — T p2l i = . . 9 8 ——5.—__1—-— g_r & 6 NO ThF, IN FUEL SALT i : x 10 | ‘ ) : ] e o ' TOTAL POWER: 600 Mw (th) o EXTERNAL VOLUME OF SALT: 339 3 04 A——————t— CORE AND BLANKET SALTS NQ.1 NUMBERS ON CURVE POINTS ARE CORE : DIAME TERS IN FEET 10 | 03 " l \ | L1 0 200 400 600 800 1000 1200 1400 CRITICAL INVENTORY (kg OF UZ3%) Fig. 8. [Initial Fuel Regeneration in Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U235, 16 seen that, with no thorium in the core, a regenera- tion ratio of 0.5 can be obtained with an inventory of 100 kg of U23% in a 7-ft-dia core. The addition of 0.25 mole % thorium to the core salt yields a regeneration of about 0.6 for an inventory of 200 kg in a 7-ft-dia core. Optimum core size increases hereafter. Also the rate of increase of regeneration ratio falls off substantially. With 0.75 mole % thorium, a regeneration of 0.67 is obtained with 400 kg of fuel in a 10-ft-dia core. As mentioned it was felt that it would be difficult to fabricate reliable core vessels having diameters greater than 10 ft, and, accordingly, larger cores were not investigated. However, an examination of the curves indicates that further increases in tho- above, rium loading and core diameter would probably not increase the regeneration ratio above 0.7. Reactors employing core and blanket salts No. 2 (see Table 4) require somewhat lower inventories than the corresponding cores using salts No. 1, as shown in Fig. 9. The 8-ft-dia core, for instance, requires only about 600 kg of U?3° with a loading of 1 mole % thorium, whereas with the No. 1 salts, about 850 kg was required. ratios, however, are about the same, ranging from 0.62 to 0.64 for the 8-ft-dia cores. The regeneration UNCLASSIFIED ORNL-LR-DWG 39528 0.65 oo ® | - l o / \ ; \.? 5 0860 ‘—1—4 - G —— & 1 & \ \ ! . i» g " ‘. oL TNt = ; ! 05 075 1.0 < ® | 3 ’ ‘ | » | 2 s | i ! | - & 055 44 0.25mole %o ThE, IN FUEL SALT — ~—1——- —‘L w 88 L x | | \ CORE AND BLANKET SALTS NG.2 NUMBERS ON DATA POINTS ARE CORE DIAMETERS IN FEET i | { | _ . 050 L C 200 400 600 800 1000 1200 CRITICAL INVENTORY (kg OF UZ39) Fig. 9. Initial Fuel Regeneration in Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U235, With U232 as the fuel, there is a marked im- provement in the performance, and the inventories are much lower. The performance of cores using core and blanket salts No. 1, with thorium con- centrations ranging up to | mole %, is shown in Fig. 10. The regeneration ratios range up to 0.95 at inventories less than 300 kg of U233, The beryllium-rich core and blanket salts (No. 2) gave substantially the same results, as shown in Fig. 11. UNCLASSIFIED ORNL-LR-DWG 39529 1.0 ® — /. . 0.9 |— ///"",‘,:’ | - - G N0 e ThF, %{ N FUEL SALT 08 .:< 0.75 o K // ~ x 07 0.5 z A 5 1 H < & o6 /0'25 = w a ol 05 ! f ® NO ThF, IN FUEL SALT 04 — CORE AND BLANKET SALTS NO. ! 0.3 | 0 100 200 300 400 CRITICAL INVENTORY (kg OF UZ3% Fig. 10. Initial Regeneration of Fuel in Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled UNCLASSIFIED (o ORNL—-LR-DWG 39530 . T T [ NUMBERS ON DATA POINTS ARE CORE DIAMETERS IN FEET | | .0 ———— CORE AND BLANKET SALTS NO. 2 \ t mole % ThF, IN FUEL SALT >& 0.9 — 87 o = 0.75 mole % ThF, (..:’_ g o ‘ 7e” 1 © z 0.50 mole % ThF, 6",'8/ 5 o8 H 2 7 Lfi' @ ; ot L ® £ 0.25 mole % ThF4\7/8 \ 7@ 8@ 0.6 \ 0.5 0 50 100 150 200 CRITICAL INVENTORY (kg OF U233) Fig. 11. Initial Regeneration of Fuel in Two-Region, Homogeneous, Molten-Fluotide-Salt Reactors Fueled . 233 Wlfh U . Regeneration ratios of the order of 0.6 can be ob- tained at inventories of about 100 kg of U233, Increasing the thorium concentration up to 7 mole % gives a monotonically increasing regenera- tion ratio, up to about 1,09, but the fuel inventories become very high. The performance of cores having diameters ranging from 4 to 12 ft and thorium con- centrations of 1, 2, 4, and 7 mole % are shown in Fig. 12. The dashed line is the estimated envelope of the curves shown and represents the locus of maximum regeneration for a given inventory. It is seen that regeneration ratios above 1,0 can be ob- tained from fuel investments of 400 kg or greater. Also, it appears that the 8-ft-dia cores give about the highest regeneration at all thorium concentra- In Fig. 13 the performances of 8-ft-dia U235 tions, cores using fuel and blanket salts No. 2 and and U233 fuel, respectively, are compared. With U233, a maximum regeneration of about 0.65 is obtained at an inventory of about 400 kg. The same amount of U233 gives a regeneration ratio of 1.0, and 1000 kg of U%33 gives a regeneration of 1.07. UNCLASSIFIED ORNL-LR-DWG 39534 1.45 1 \ FUEL SALT: 37 mole % BeF, + 63 mote 7 LiF + UF, + ThF, 110 == e — K | == ///6 R ~ 7mole 7o ThF, 105 7 4 — o 8 Lo | IN FUEL SALT ) Ao | 5 o g // 6 "'-—-—z- 4 mole % ThF, = / e / =100 £ 2 mole T ThF, & Lo y coe L) e~ ] | - & /8 o { = / 095 ! Il i NUMBERS ON DATA POINTS ARE i CORE DIAMETERS iN FEET P 0.90 —~|‘t~——1 moie > Thfy ;97 f 0.85 ’ ' 0 400 800 1200 1600 2000 2400 TOTAL FUEL INVENTORY (kg OF Uu233) Fig. 12, Initial Regeneration Ratio in Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U233, 17 COMPARATIVE PERFORMANCE In comparison with converter reactors, the homo- geneous MSR fueled with U23%is somewhat inferior UNCLASSIFIED ORNL-LR-DWG 39532 14 F : [ [ ] CORE AND BLANKET SALTS NO. 2 CORE DIAMETER: 8t 1.2 |~ EXTERNAL FUEL VOLUME: 339 f13 233 £D u2®3 FUEL ____._..-—----—""-"'—'-'fi 1.0 o + |._ 90 148 2.5 8.8 0.52 Y233 >90 75 239 PWR 60 600 486 26.7 1.81 33.2 0.7 Pu 0.72 96 YER 134 520 Saturated 26.2 2.6 550 6.6 20.4 0.64 py239 CER 140 600 485 34.7 >90 275 4.7 29.8 0.47 u233 MSR 260 1800 1000 40.5 >90 400 6.8 38.2 0.6~0.8%* p233 6l *EFR, Enrico Fermi Reactor, Solid Fuel Reactors, J. R. Dietrich and W. H. Zinn (eds.), Addison-Wesley, Reading, Mass., 1958. GCR, ORNL Gas Cooled Reactor, ibid SGR, Sodium Graphite Reactor, Sodium Graphite Reactors, C. Starr and R. W. Dickinson (eds.), Add.ison-Wesley, 1958. OMR, Organic Moderated Reactor, Solid Fuel Reactors, loc. cit. DPR, Dresden Nuclear Power Reactor, Boiling Water Reactors, A. W. Kramer (ed.), Addison-Wesley, 1958. ERR, Elk River Reactor, ibid PWR, Pressurized Water Reactor, Shippingport Pressurized Water Reactor, R. T. Bayard et al., Addison-Wesley, 1958. YER, Yankee Atomic Electric Co. Reactor, Preliminary Hazards Summary Report, Y AEC-60 (1957). CER, Consolidated Edison Reactor, Nuclear Reactor Data No. 2, Raytheon Mfg. Co., Waltham, Mass., 1956. MSR, Molten Salt Reactor, Fluid Fuel Reactors, J. A. Lone, H. G. MacPherson, and F. Maslan (eds.), Addison-Wesley, 1958. **Ratio depends on processing rate. 1-10. 11. 12. 13. 14. 15. 16. 17. 18. 19. 20. 21. 22. 23. 24. 25. 26. 27. 28. 29. 30. 31. 32. 33. 34. 35. 36. 37. 38. 39. 40, 41. 42. 43. 44. 45. 46-47. ITA-EPNEOQPIEIMEPO-E-MINOUEDEAMEOErmame . Alexander Bettis . Blankenship . Blizard . Boch . Boudreau . Boyd . Bredig Breeding . Briggs . Browning . Campbell . Carr . Carrison Cathers . Charpie . Claiborne . Culler . DeVan . Ergen . Estabrook . Ferguson . Fraas . Grimes uth OO UM< AIrOPr-">ITOMBB-I>TMAIT TTNYLO . W. Hoffman = . Jordan . Kasten . Keilholtz . Keim - 0 =EAOI . Kelley . Kertesz . W. Kinyon . E. Lackey A. Lane . N. Lyon . G. MacPherson INTERNAL DISTRIBUTION 48. 49. 50. S1. 52. 53. 54. 55. 56. 57. 58. 59. 60. 61. 62. 63. 64. 65. 66. 67. 68. 69. 70. 71. 72. 73. 74. 75. 76. 77. 78. 79-82. 83--102. 103. 104--105. EXTERNAL DISTRIBUTION 106. F. C. Moesel, AEC, Washington 107. Division of Research and Development, AEC, ORO 108-695. Given distribution as shown in TID-4500 (14th ed.) under Reactors—Power category (75 copies — OTS) ORNL-2751 Reactors—Power TID-4500 (14th ed.) . Manly . Mann . Mann . McDonald Metz . Milford . Miller . Morgan Nessle Perry . Reyling . Roberts Robinson . Rosenthal . Savage . Savolainen Shor Skinner A. Swartout . Taboada . E. Thoma . Tobias . Trauger . Vonderl.age . Watson . Weinberg . Whatley . Whitman . Willioms . Winters J. Zasler ORNL — Y-12 Technical Library, Document Reference Section Laboratory Records Department Laboratory Records, ORNL R.C. Central Research Library CCEEETAAZTTICENZT TP OT OOOZPONMUIANP-—ZTPPIZZT-TPORCPIEIME mOoomzzx= N W 21