CHAPTER 14 NUCLEAR ASPECTS OF MOLTEN-SALT REACTORS* The ability of certain molten salts to dissolve uranium and thorium salts in quantities of reactor interest made possible the consideration of fluid- fueled reactors with thorium in the fuel, without the danger of nuclear ac- cidents as a result of the settling of a slurry. This additional degree -of freedom has been exploited in the study of molten-salt reactors. Mixtures of the fluorides of alkali metals and zirconium or beryllium, as discussed in Chapter 12, possess the most desirable combination of low neutron absorption, high solubility of uranium and thorium compounds, chemical inertness at high temperatures, and thermal and radiation sta- bility. The following comparison of the capture cross sections of the alkali metals reveals that Li? containing 0.019, Li% has a cross section at 0.0795 ev and 1150°F that is a factor of 4 lower than that of sodium, which also has a relatively low cross section: Element Cross section, barns Li7 (containing 0.019; Li®) 0.073 Sodium 0.290 Potassium 1.13 Rubidium 0.401 Cesium 29 The capture cross section of beryllium is also satisfactorily low at all neutron energies, and therefore mixtures of Lil' and BeFg, which have satisfactory melting points, viscosities, and solubilities for UF4 and ThF4, were selected for investigation in the reactor physics study. Mixtures of NaF, ZrF4, and UF4 were studied previously, and such a fuel was successfully used in the Aircraft Reactor Iixperiment (see Chap- ters 12 and 16). Inconel was shown to be reasonably resistant to corrosion by this mixture at- 1500°F, and there is reason to expect that Inconel equipment would have a life of at least several years at 1200°F. As a fuel for a central-station power reactor, however, the Nal'-ZrF4 system has several serious disadvantages. The sodium capture cross section is less favorable than that of Li?. More important, recent data [1] indicate that the capture cross section of zirconium is quite high in the epithermal and intermediate neutron energy ranges. In comparison with the Lil'-BeFs system, the Nal'-Zrl, system has inferior heat-transfer characteristics. *By L. G. Alexander. 626 NUCLEAR ASPECTS OF MOLTEN-SALT REACTORS 627 Finally, the INOR alloys (see Chapter 13) show promise of being as resistant to the beryllium salts as to the zirconium salts, and therefore there is no compelling reason for selecting the Nal-ZrF4 system. Reactor calculations were performed by means of the Univac* program Ocusol [2], a modification of the Eyewash program [3], and the Oracle?t program Sorghum. Ocusol is a 31-group, multiregion, spherically symmet- ric, age-diffusion code. The greoup-averaged cross sections for the various elements of interest that were used were based on the latest available data [4]. Where data were lacking, reasonable interpolations based on resonance theory were made. The estimated cross sections were made to agree with measured resonance integrals where available. Saturation and Doppler broadening of the resonances in thorium as a function of concentration were estimated. Inelastie scattering in thorium and fluorine was taken into account crudely by adjusting the value of £¢:; however, the Ocusol code does not provide for group skipping or anisotropy of scattering, Sorghum is a 31-group, two-region, zero-dimensional, burnout code. The group-diffusion equations were integrated over the core to remove the spatial dependency. The spectrum was computed, in terms of a space-averaged group flux, from group scattering and leakage parameters taken from an Ocusol caleulation. A critical calculation requires about 1 min on the Oracle; changes in concentration of 14 elements during a specified time can then be computed in about 1 sec. The major assumption imvolved 18 that the group scattering and leakage probabilities do not change appreciably with changes in core composition as burnup progresses. This assumption has been verified to a satisfactory degree of approximation. The molten salts may be used as homogeneous moderators or simply as fuel carriers in heterogeneous reactors. Although, as discussed below, graphite-moderated heterogeneous reactors have certain potential advan- tages, their technical feasibility depends upon the compatibility of fuel, graphite, and metal, which has not as yet been established. For this rea- son, the homogeneous reactors, although inferior in nuclear performance, have been given greatest attention. A preliminary study indicated that if the integrity of the core vessel could be guaranteed, 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 *Universal Automatic Computer at New York University, Institute of Mathe- matics. 70ak Ridge Automatic Computer and Logical Engine at Qak Ridge National Laboratory. 628 NUCLEAR ASPECTS OF MOLTEN-SALT REACTORS [cHAP. 14 spherical regions: (1) the core, (2) an INOR-8 core vessel 1/3 in. thick, (3) a blanket approximately 2 ft thick, and (4) an INOR-8 reactor vessel 2/3 in, thick. The diameter of the core and the concentration of thorium in the core were selected as independent variables. The primary dependent variables were the critical concentration of the fuel (U235 U233 or Pu?39), and the distribution of the neutron absorptions among the various atomic species in the reactor. Irom these, the critical mass, critical inventory, regeneration ratio, burnup rate, ete. can be readily calculated, as described in the following section. 14-1. HomogENEOUS REACTORS FUBLED wiTH U235 While the isotope U232 would be a superior fuel in molten fluoride-salt reactors (see Section 14-2), it is unfortunately not available in quantity. Any realistic appraisal of the immediate capabilities of these reactors must be based on the use of U235, The study of homogeneous reactors was divided into two phases: (1) the mapping of the nuclear characteristies of the initial (i.e., “clean’) states as a function of core diameter and thorium concentration, and (2) the analysis of the subsequent performance of selected initial states with various processing schemes and rates. The detailed results of these studies are given in the following paragraphs. Briefly, it was found that regenera- tion ratios of up to 0.65 can be obtained with moderate investment in U235 (less than 1000 kg) and that, if the fission products are removed (Article 14-1.2) at a rate such that the equilibrium inventory is equal to one year’s production, the regeneration ratio can be maintained above 0.5 for at least 20 years. 14-1.1 Initial states. A complete parametric study of molten fluoride- salt reactors having diameters in the range of 4 to 10 ft and thorium con- centrations in the fuel ranging from 0 to 1 mole 9, ThF4 was performed. In these reactors, the basic fuel salt (fuel salt No. 1) was a mixture of 31 mole 9, Bel's and 69 mole 9 LiI", which has a density of about 2.0 g/cc at 1150°F. The core vessel was composed of INOR-8. The blanket fluid (blanket salt No. 1) was a mixture of 25 mole 9, ThI'4 and 75 mole 95 L1k, which has a density of about 4.3 g/ce at 1150°F. In order to shorten the calculations in this series, the reactor vessel was neglected, since the re- sultant error was small. These reactors contained no fission products or nonfissionable isotopes of uranium other than U#38, A summary of the results is presented in Table 14-1, in which the neutron balance is presented in terms of neutrons absorbed in a given element per neutron absorbed in U235 (both by fission and the n—y reaction). The sum of the absorptions is therefore equal to n, the number of neutrons produced by fission per neutron absorbed in fuel. Further, the sum of the 14-1] HOMOGENEOUS REACTORS FUELED WITH U235 629 (X10'9) 40 ‘ 30 |- AN 4 ® Calculated Values mole % ThF4 In uel Salt 7 & » U235 concentration atomsfem3 o | T T » 2 — Interpolations of Data No ThFA in Fuel Salt \ 9 1 | l | 2 4 6 8 10 Core Diameter, f Fic. 14-1. Initial critical concentration of U?3% in two-region, homogeneous, molten fluoride-salt reactors. absorptions in U and thorium in the fuel, and in thorium in the blanket salt gives directly the regeneration ratio. The losses to other elements are penalties imposed on the regeneration ratio by these poisons; i.e., if the core vessel could be constructed of some material with a negligible cross seec- tion, the regeneration ratio could be increased by the amount listed for capture in the core vessel. The Inventories in these reactors depend in part on the volume of the fuel in the pipes, pumps, and heat exchangers in the external portion of the fuel circuit. The inventories listed in Table 14-1 are for systems having a volume of 339 ft? external to the core, which corresponds approximately to a power level of 600 Mw of heat. In these calculations it was assumed that the heat was transferred to an intermediate coolant composed of the fluorides of Li, Be, and Na before being transferred to sodium metal. In more recent designs (see Chapter 17), this intermediate salt loop has been replaced by a sodium loop, and the external volumes are somewhat less because of the improved equipment design and layout. Critical concentration, mass, inventory, and regeneration ratio. The data in Table 14-1 are more easily comprehended in the form of graphs, such as Fig. 14-1, which presents the critical concentration in these reactors as a function of core diameter and thorium concentration in the fuel salt. The data points represent caleulated values, and the lines are reasonable interpolations. The maximum concentration caleulated, about 35 x 101° Total power: 600 Mw (heat). TasLE 14-1 INtrIaL-STATE NUCLEAR CHARACTERISTICS OF Two-REcioN, HoMoGENEOUS, MowvteEN FLuoripE-SaLT REAcTORS FUELED wiTH U235 Fuel salt No. 1: 31 mole 9, BeFs + 69 mole 9 1aF + UF4+ ThF4. Blanket salt No. 1: 25 mole 9, ThF4+ 75 mole 9}, LiF. External fuel volume: 339 {t3. Case number 1 2 3 4 5 6 7 Core diameter, ft 4 5 5 5 5 5 6 Th¥4 in fuel salt, mole 9 0 0 0.25 0.5 0.75 1 0 U235 in fuel salt, mole % 0.952 0.318 0.561 0.721 0.845 0.938 0.107 U235 atom density* 33.8 11.3 20.1 25.6 30.0 33.3 3.80 Critical mass, kg of U235 124 81.0 144 183 215 239 47.0 Critical inventory, kg of U235 | 1380 501 891 1130 1330 1480 188 Neutron absorption ratiost U235 (fissions) 0.7023 0.7185 0.7004 0.6996 0.7015 0.7041 0.7771 U235 (n— 0.2977 0.2815 0.2996 0.3004 0.2985 0.2959 0.2229 Be-Li-F in fuel salt 0.0551 0.0871 0.0657 0.0604 0.0581 0.0568 0.1981 Core vessel 0.0560 0.0848 0.0577 0.0485 0.0436 0.0402 0.1353 Li-F in blanket salt 0.0128 0.0138 0.0108 0.0098 0.0093 0.0090 0.0164 Leakage 0.0229 0.0156 0.0147 0.0143 0.0141 0.0140 0.0137 U238 in fuel salt 0.0430 0.0426 0.0463 0.0451 0.0431 0.0412 0.0245 Th in fuel salt 0.0832 0.1289 0.1614 0.1873 Th in blanket salt 0.5448 0.5309 0.4516 0.4211 0.4031 0.3905 0.5312 Neutron yield, 7 1.73 1.77 1.73 1.73 1.73 1.74 1.92 Median fission energy, ev 270 15.7 105 158 270 425 0.18 Thermal fissions, 9, 0.052 6.2 0.87 0.22 0.87 0.040 35 n—y capture-to-fission ratio, a 0.42 0.39 0.43 0.43 0.43 0.4203 0.28 Regeneration ratio 0.59 0.57 0.58 0.60 0.61 0.62 0.56 continued 0€9 SLOUJISV UVHATIOAN Jd0 SHOLOVHY LIVS-NAHLTON F1 "dVHD] Tapre 11 1 (continued) ("ase number ~ 9 [0 11 12 13 14 Core diameter, ft {h 6 (s 6 7 8 8 ThI4 in fuel salt, mole 9, 0.25 0.5 0.75 1 0.25 ) 0.25 U235 in fuel salt, mole 9 0.229 0.408 (.552 0. 662 0.114 0.047 0.078 U235 atom density* 8.13 14.5 19.6 23.5 4.0 1.66 2.77 Critical mass, kg of U235 101 179 243 291 79.6 48.7 &1.3 Critical inventory, kg of U233 404 716 972 1160 230 110 184 Neutron absorption ratiost U235 (figgions) 0.7343 0.7082 0.7000 0.7004 0. 7748 0. 8007 0.7930 U233 (n—y) 0.2657 0.2918 0. 3000 0.2996 0.2252 (. 1993 0.2070 Be--Li-F in fuel salt 0.1082 00770 0.0669 0.0631 0. 1880 0.4130 0.2616 Core vessel 0.0795 0.0542 0 0435 0.0388 0.0951 (. 1491 0.1032 Li—F in blanket salt 0.0116 0.0091 0.0081 0.0074 0.0123 (.0143 0.0112 Leakage ‘ 0.0129 0.0122 0.0119 0.0116 0.0068 0.0084 0.0082 U238 in fuel salt 0.0375 0.0477 0.0467 (. 0452 0.0254 0.0143 00196 Th in fuel salt 0.1321 0. 1841 0.2142 0.2438 0.1761 0.2045 Th in blanket salt 0 4318 0.3683 | 0.3378 03202 | 0.4008 0.4073 03503 Neutron yield, 7 1.82 1.75 1.73 1.73 1.91 2.00 1.96 Median fission energy, ev 5.6 38 100 120 0.16 Thermal 0.10 Thermal fissions, 9% 13 3 0.56 0.48 33 59 45 n -y capture-to-fission ratio, « 0.36 0.41 0.42 0. 42 0.29 0.25 0 26 Regencration ratio .61 .60 0.60 0.61 0.61 0.42 0.57 ¥Atoms (X 10719) /ec. $Neutrons absorbed per neutron absorped in U233, continued [1-51 ezl HLLAW QETEOI SHOLOVAM SOOMNEDOKOIL 1€ TABLE 14-1 (continued) Case number 15 16 17 18 19 20 21 22 Core diameter, ft 8 8 8 10 10 10 10 10 ThF4 in fuel salt, mole 0.5 0.75 1 0 0.25 0.5 0.75 1 U235 in fuel salt, mole 9 0.132 0.226 0.349 0.033 0.052 0.081 0.127 0.205 U235 atom density* 4 .67 8.03 12.4 1.175 1.86 2.88 4 .50 7.28 Critical mass, kg of U235 137 236 364 67.3 107 165 258 417 Critical inventory, kg of U235 310 535 824 111 176 272 425 687 Neutron absorption ratiost U235 (fissions) 0.7671 0.7362 0.7146 0.8229 0.7428 0.7902 0.7693 0.7428 U235 (n 0.2329 0.2638 0.2854 0.1771 0.2572 0.2098 0.2307 0.2572 Be-Li-F in fuel salt 0.1682 0.1107 0.0846 0.5713 0.3726 0.2486 0.1735 0.1206 Core vessel 0.0722 0.0500 0.0373 0.1291 0.0915 0.0669 0.0497 0.0363 Li-F in blanket salt 0.0089 0.0071 0.0057 0.0114 0.0089 0.0073 0.0060 0.0049 Leakage 0.0080 0.0077 0.0074 0.0061 0.0060 0.0059 0.0057 0.0055 U288 in fuel salt 0.0272 0.0368 0.0428 0.0120 0.0153 0.0209 0.0266 0.0343 Th in fuel salt (0.3048 0.3397 0.3515 0.2409 0.3691 0.4324 0.4506 Th in blanket salt 0.3056 0.2664 0.2356 0.3031 0.2617 0.2332 0.2063 0.1825 Neutron yield, n 1.89 1.82 1.76 2.03 2.00 1.95 1.90 1.83 Median fission energy, ev 0.17 5.3 27 Thermal | Thermal 0.100 0.156 1.36 Thermal fissions, 9 29 13 5 66 56 43 30 16 n—y capture-to-fission ratio, « 0.30 0.36 0.40 0.21 0.24 0.26 0.30 0.35 Regeneration ratio 0.64 0.64 0.63 0.32 0.52 0.62 0.67 0.67 *Atoms (X 10719)/ce. tNeutrons absorbed per neutron absorbed in U235, 43y SHOLOVHAY LIVS-NALTOW 40 SLOAASV dVATIAN ¥1 'dVHD] 14-1} HOMOGENEOUS REACTORS FUELED WITH U235 633 T T T ] T mole % ThF4 In Fuel Salt 1 . 400 — e Calculated volues / — Interpolations of . o Data o 4 =] ® 300+ — > ° - 2 / . 075 {v a o -—-—-—..____._____.—-—-—‘—"'—_— = /./ o 200 I/ -] = o S ... ~———No ThF 4 In Fuel Sait 04 5 6 7 8 g 10 Core Diameter, Fic. 14-2. Initial critical masses of U235 in two-region, homogeneous, molten fluoride-salt reactors. atoms of U35 per cubic centimeter of fuel salt, or about 1 mole 9, UF,, is an order of magnitude smaller than the maximum permissible concentra- tion (about 10 mole 7). The corresponding eritical masses are graphed in IFig. 14-2. As may be =een, the eritical mass 1s a rather complex function of the diameter and the thorium concentration. The calculated points are shown here also, and the sulid lines represent, 1t is felt, reliable interpolations. The dashed lines were drawn where insufficient numbers of points were calculated to define the curves precisely; however, they are thought to be qualitatively correct. Sinee reactors having diameters less than 6 ft are not economically attrac- tive, only one case with a 4-ft-diameter core was computed. The critical masses obtained in this study ranged from 40 to 400 kg of U= However, the ecritical inventory in the entire fuel circuit is of more interest to the reactor designer than is the critical mass. The critical in- ventories corresponding to an external fuel volume of 339 ft3 are therefore 2 233 £ 200 e ——— v = {Accumulated) J ! | i | | | [ i G 10 20 Time of Operation, years F1g. 14-10. Long-term nuclear performance of typical two-region, homogeneous, molten fluoride-salt reactors fueled with U235, Core diameter, 8 ft; total power, 600 Mw (heat); load factor, 0.80. 1/4300 per day. The U2% inventory rises to 826 kg and then falls, at the end of eight months, to 587 kg. At this time, the processing rate is in- creased to 1/240 per day (eight-month cycle), but the thorium is returned to the core and the thorium concentration falls thereafter only by burnup. It may be seen that the U235 inventory creeps up slowly and that the re- generation ratio falls slowly. The increase in U235 inventory could have been prevented by withdrawing thorium at a small rate; however, the re- generation ratio would have fallen somewhat more rapidly, and more U235 feed would have been required to compensate for burnup. 14-2. HomoGENEQOUs REACTORS FUELED wrTH [J233 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 sections of U235 and Pu23?, Thus initial critical inventories are less, and less additional fuel is required to override poisons. Also, the parasitic cross section is sub- stantially less, and fewer neutrons are lost to radiative capture. Further, the radiative captures result in the immediate formation of a fertile iso- C'ore diameter: 8 ft. TanLe 11-4 Fxternal fuel volume: 339 ft3. NucLeak Perrormasce or A Two-Redion, HoMocENEOUS, MovrreN FLuoribe-Saur Reactor Frenep wita U232 AND Coxrtaining 1 mour 9 Tuly 1x THE Furn Sart Total power: 600 Mw (heat). Load factor: 0.8. Initial state After 1 year Inventory, Absorptions, Fissions, Inventory, Absorptions, Fissions, kg % o kg % % Core elements Th232 2,100 20.3 2,100 16.7 [*a233 8.2 0.3 17233 61.0 5.9 12.5 234 1.9 0.0 235 604 55.4 100 &893 49 .3 86.3 U236 62.2 1.8 Np237 4.2 0.2 1238 45.3 2.2 57.9 2.0 Pu2ss 6.8 0.8 1.2 Fission fragments 181 3.8 L7 3,920 1.9 3,920 0.9 Be? 3,008 0.6 3,008 0.5 Ir9 24,000 3.2 24,000 3.0 Rlanket element 238 8 7 Total fuel 604 963 1225 burnup rate, kg/day 0.69 0.62 U235 feed rate, kg/day 2.22 1.28-0.73 Regeneration ratio 0.64 0.53 continued [g-%1 pez(l HLIM UETINA SUOLOVAY SNOANTADOWOH L¥9 TaBLE 14-4 (continued) After 2 years After 5 years Inventory, Absorptions, Fissions, Inventory, Absorptions, Fissions, kg % %o kg %% % Core elements Th232 2,100 16 .3 2,100 15.4 Pg233 7.9 02 7.5 0.2 Uzss 110 9 7 20.8 201 15.3 33.0 234 6.5 0.1 27.1 04 235 863 44 .3 77.4 818 36.9 64.1 U236 115 31 222 5.2 Np237 0.8 0.4 1.8 0.8 [j238 69.7 23 9.0 2.7 Py239 12.0 13 1.8 24 .3 2.0 2.9 Fission fragments 181 36 181 3.1 Li? 3,920 08 3,920 0.6 Be? 3,008 0.5 3,008 0.5 Fo 24,000 3.0 24,000 3.0 Blanket element 233 16 24 Total fuel 990 1,045 U235 burnup rate, kg/day 0.58 0.47 U235 feed rate, kg/day 0.50 0.45 Regeneration ratio 0.53 0.54 continued 8+9 SIOUdSY UVHTOON SHOLOVHY LIVS-NEHLIOK J0 F1 dVHD) TaBLE 14-4 (continued) After 10 years After 20 years Inventory, Absorptions, Fissions, Inventory, Absorptions, Fissions, kg Yo % kg %o % Core elements Th232 2,100 14.6 2,100 13.7 Pa?33 7.1 0.2 6.7 0.2 233 266 17.6 38.38 322 18.8 41.0 U234 64.4 0.8 124 1.4 U235 831 33.5 58.2 R72 31.7 54.9 U236 328 6.7 450 7.9 Np237 2.6 0.9 3.2 1.0 U238 10.8 2.9 12.9 3.0 Pu?239 37.3 2.4 3.5 52.6 2.8 4.1 Fission fragments 181 2.7 181 2.4 Li7 3,920 0.5 3,920 0.4 Be? 3,008 0.5 3,008 0.5 Ie 24,000 3.0 24,000 3.0 Blanket element 233 28 33 Total fuel 1,129 1,232 U235 burnup rate, kg/day 0.41 0.38 U235 feed rate, kg/day 0.44 0.39 Regeneration ratio 0.533 0.530 [Z-%1 cpz{l HIIM JATINA SHOLOVHY SNO0ANITDONOH 6%9 650 NUCLEAR ASPECTS OF MOLTEN-SALT REACTORS [cHAP. 14 tope, U234, The rate of accumulation of U?3% is orders of magnitude smaller than with U235 as a fuel, and buildup of Np?37 and Pu?3? is negligible. The mean neutron energy is rather nearer to thermal m these reactors than it is in the corresponding U235 cases. Consequently, losses to core vessel and to core salt tend to be higher. Both losses will be reduced sub- stantially at higher thorium concentrations. 14-2.1 Initial states. Results from a parametric study of the nuclear characteristics of two-region, homogeneous, molten fluoride-salt reactors fueled with U23? are given in Table 14-5. The core diameters considered range from 3 to 10 ft, and the thorium concentrations range from 0.25 to 1 mole 9. Although the regeneration ratios are less than unity, they are very good compared with those obtained with U235, With 1 mole 9, ThF4 in an 8-ft-diameter core, the U23? inventory was only 196 kg, and the re- generation ratio was 0.91. The regeneration ratios and fuel inventories of reactors of various diam- eters containing 0.25 mole 9} thorium and fueled with U235 or U233 gre compared in Fig. 14-11. The superiority of U233 is obvious. 1.0 | | U233 .0 5 08[ — = 9 3 g U235 o \ 0.4 | | 0 200 400 600 Critical Inventory, kg of U F1a. 14-11. Comparison of regeneration ratios in molten-salt reactors containing 0.25 mole 9, ThF, and U?35- or U?33.enriched fuel. 14-2.2 Intermediate states. Calculaticns of the long-term performance of one reactor (Case 51, Table 14-5) with U233 as the fuel are described below. The core diameter used was 8 ft and the thorium concentration was 0.75 mole 9. The changes in inventory of U233 and regeneration ratio are listed in Table 14-6. During the first year of operation, the inventory rises from 129 to 199 kg, and the regeneration ratio falls from 0.82 to 0.71. If the reprocessing required to hold the concentration of fission products TasLE 14-5 NucrLear CnaracrerisTics oF Two-Reaion, HoMoGENEOUS, MoLTEN FLUORIDE-SALT REACTORS FUKLED wiTh U233 Core diameter: 8 ft. External fuel volume: 339 ft3. Total power: 600 Mw (heat). Load factor: 0.8, Case number 41 42 43 44 45 46 Fuel and blanket salts* 1 1 1 1 1 1 Core diameter, ft 3 4 4 5 6 6 Th¥, in fuel salt, mole 9 0 0 0.25 0 0.25 0.25 U233 in fuel salt, mole 9 0.592 0.158 0.233 0.106 0.048 0.066 U233 atom density 21.0 6.09 8.26 3.75 1.66 2.36 Critical mass, kg of U233 64.9 22.3 30.3 26.9 20.5 29.2 Critical inventory, kg of U233 1620 248 337 166 82.0 117 Neutron absorption ratios? U232 (figsions) 0.8754 0.8706 0.8665 0.8725 0.8814 0.8779 U283 (n—y) 0.1246 0.1294 0.1335 0.1275 0.1186 0.1221 Be-Li-F in fuel salt 0.0639 0.1061 0.0860 0.1472 0.3180 0.2297 Core vessel 0.0902 0.1401 0.1093 0.1380 0.1983 0.1508 Li-Be-F in blanket salt 0.0233 0.0234 0.0203 0.0196 0.0215 0.0179 Leakage 0.0477 0.0310 0.0306 0.0193 0.0160 0.0157 Th in fuel salt 0.1095 0.1593 0.1973 Th in blanket salt 0.9722 0.8857 0.8193 0.7066 0.6586 0.5922 Neutron yield, 7 2.20 2.19 2.18 2.19 2.21 2.20 Median fission energy, ev 174 14 19 2.9 0.33 1.2 Thermal fissions, % 0.053 8.0 2.3 16 38 29 n—y capture-to-fission ratio, « 0.14 0.15 0.15 0.15 0.13 0.14 Regeneration ratio 0.97 0.89 0.93 0.87 0.66 0.79 continued [z¥1 ezl HLIM QHTENd SHOLOVAYM SNOINADOWOH 169 TaBLE 14-5 (continued) Case number 47 48 49 50 51 Fuel and blanket salts* 1 1 1 1 2 Core diameter, ft 8 8 10 10 8 ThF, in fuel salt, mole 9 0.25 1 0.25 1 0.75 U233 in fuel salt, mole %, 0.039 0.078 0.031 0.063 0.0597 U233 atom densityt 1.40 2.95 1.10 2.29 1.97 Critical mass, kg of U233 41.1 86.6 63.0 131 58.8 Critical inventory, kg of U233 93.1 196 104 216 129 Neutron absorption ratiosi U233 (figsions) 0.8850 0.8755 0.8881 0.8781 0.8809 U233 (n—y) 0.1150 0.1245 0.1119 0.1219 0.1191 Be-Li-F in fuel salt 0.3847 0.1899 0.5037 0.2360 0.2458 Core vessel 0.1406 0.0778 0.1168 0.0629 0.1168 Li-Be-F in blanket salt 0.0141 0.0095 0.0108 0.0071 0.0187 Leakage 0.0095 0.0090 0.0068 0.0065 0.0050 Th in fuel salt 0.2513 0.5768 0.2852 0.6507 0.4903 Th in blanket salt 0.4211 0.3344 0.3058 0.2408 0.3325 Neutron yield, n 2.22 2.20 2.23 2.20 2.21 Median fission energy, ev 0.20 1.1 509, Th 3.2 0.68 Thermal fissions, 9 43 24 50 30 34 n—y ecapture-to-fission ratio, « 0.13 0.14 0.13 0.14 0.14 Regeneration ratio 0.67 0.91 0.59 0.89 0.82 649 SLOHISY UVHATIAN SUOLOVHEY LIVS-NALIOW 40 *Fuel salt No. 1: 31 mole 9, BeFs + 69 mole 9, LiF + UF4+ ThF. Blanket salt No. 1: 25 mole 9%, ThF4+ 75 mole 9, LiF Fuel salt No. 2: 37 mole 9, BeFs + 63 mole 9, LiF' + UF4+4 ThF, Blanket salt No. 2: 13 mole 9, ThF4 + 16 mole 9, BeF3 4+ 71 mole 9, LiF TAtoms (X 10719) /ce. tNeutrons absorbed per absorption in U233, $1 "dVHD] NucLear PerrormaNck or A Two-RecioN, HoMoGuNEOUS, Moty Frroripe-Sart Reacror Fuernep wita U233 Anxp TaBLE 14-6 Conrtaining 0.75 moLE 97 TuF4 IN THE FUEL SavT Total power: 600 Mw (heat). Load factor: 0.8. Core diameter: 8 ft. External fuel volume: 339 ft3. Initial state After 1 year Inventory, Absorptions, Fissions, Inventory, Absorptions, Fissions, kg % % kg % o Core elements Th2s2 1,572 22.2 1,572 19.1 Pa23 9. 0.5 1233 129 45.2 100 199 45.3 99 .5 234 23.3 0.9 235 1.9 0.3 0.5 ;236 01 0.1 Np237 1238 P39 IFission fragments 181 7.9 Li6 3,920 6.5 3,920 3.4 Bef 3,004 0.8 3,008 6.7 e 24,000 40 24,000 3.5 Blanket element {233 8.6 Total fuel 129 210 1233 feed rate, kg/day 0.790 (. 370-0.189 Regeneration ratio 0.82 0.71 continued [z-¥1 ezl HIIM QATIN SHOLOVIY SNOINTDOWOH €29 TaBLE 14-6 (continued) After 2 years After 5 years Inventory, Absorptions, Fissions, Inventory, Absorptions, Fissions, kg Yo Yo kg % T Core elements Th232 1,572 18.9 1,572 18.3 Pa233 9.0 0.5 8.9 0.4 U233 204 44 .9 98.5 216 43.7 95.6 234 44 0 1.7 89 3.1 =235 5.4 0.8 1.5 17.7 2.3 4.4 236 0.6 0.3 4.2 0.2 Np27 0.1 0.1 0.5 0.1 =38 0.3 Pu239 Fission fragments 1381 7.7 181 7.2 L1 3,920 3.3 3,920 2.8 Be? 3,008 0.6 3,008 0.6 Ir19 24,000 3.4 24,000 3.3 Blanket clement, 1J233 10.7 16.2 Total fuel 220 250 U233 feed rate, kg/day 0. 188 0.181 Regeneration ratio 0.72 0.73 continued $8Y SHOLOVHY LIVS-NALIONW JO SLOAdSY HVHTOAN $1 "dVHD] TasrLe L1 6 (continued) After 10 years After 20 vears Inventory, Absorptions, Fissions, Inventory, Absorptions, Fissions, kg Yo To kg T %o Core elements Th232 1,572 17.8 1,572 17.2 Pa233 8.6 0.4 8.4 0.4 17233 231 425 92.8 247 41.5 90.5 1234 132 4.2 172 5.0 U235 32.5 3.7 7.1 47 4.8 9.0 236 12.5 0.6 24 1.1 Np237 1.7 0.2 3.4 0.3 238 1.7 0.1 5.1 0.3 Py 239 0.2 0.1 0.1 0.8 0.3 0.5 Figsion fragments 181 6.7 181 6.3 1.8 3,920 2.5 3,920 2.1 Be? 3,008 0.6 3,008 0.6 F1o 24,000 3.3 24,000 3.3 Blanket element y23s 22 .2 31.6 Total fuel 282 295 U233 feed rate, kg/day 0.171 0.168 Regeneration ratio 0.73 0.73 (5 ¥ epgll LI ETENT SHOLOVIN $10dNADONOI ¢a9 656 NUCLEAR ASPECTS OF MOLTEN-SALT REACTORS fchar. 14 and Np237 constant is begun at this time, the inventory of U?33 increases slowly to 247 kg and the regeneration ratio rises slightly to 0.73 during the next 19 years. This constitutes a substantial improvement over the per- formance with U235, 14-3. HomoGENEOUS REACTORS FUELED WiTH PLUTONIUM It may be feasible to burn plutonium in molten fluoride-salt reactors. The solubility of PuF3 in mixtures of LiF and BeFs is considerably less than that of UF4, but is reported to be over 0.2 mole % [8], which may be sufficient for criticality even in the presence of fission fragments and non- fissionable isotopes of plutonium but probably limits severely the amount of ThF4 that can be added to the fuel salt. This limitation, coupled with the condition that Pu?3? is an inferior fuel in intermediate reactors, will result in a poor neutron economy in comparison with that of U233-fueled reactors. However, the advantages of handling plutonium in a fluid fuel system may make the plutonium-fueled molten-salt reactor more desirable than other possible plutonium-burning systems. 14-3.1 Initial states. Critical concentralion, mass, tnventory, and regen- eration ratto. The results of calculations of a plutonium-fueled reactor having a core diameter of 8 ft and no thorium in the fuel salt are described below. The critical concentration was 0.013 mole % PuF3, which is an order of magnitude smaller than the solubility limits in the fluoride salts of interest. The critical mass was 13.7 kg and the critical inventory in a 600-Mw system (339 ft3 of external fuel volume) was only 31.2 kg. The core was surrounded by the Li-Be-Th fluoride blanket mixture No. 2 (13% ThF4). Slightly more than 19% of all neutrons were captured in the thorium to give a regeneration ratio of 0.35. By employing smaller cores and larger investments in Pu?3® however, it should be possible to increase the regeneration ratio substantially. Neutron balance and miscellaneous details. Details of the neutron economy of a reactor fueled with plutonium are given in Table 14-7. Parasitic cap- tures in Pu?39 are relatively high; n is 1.84, compared with a » of 2.9. The neutron spectrum is relatively soft; almost 60% of all fissions are caused by thermal neutrons and, as a result, absorptions in lithium are high. 14-3.2 Intermediate states. On the basis of the average value of @ of Pu239 it is estimated that Pu?40 will accumulate in the system until it cap- tures, at equilibrium, about half as many neutrons as Pu?®. While these captures are not wholly parasitic, inasmuch as the product, Pu?%!, is fissionable, the added competition for neutrons will necessitate an increase in the concentration of the Pu?3?. Likewise, the ingrowth of fission products 14--4] HETEROGENEOUS GRAPHITE-MODERATED REACTORS 657 will necessitate the addition of more Pu?3”. Further, the rare earths among the fission products may exert a common-ion influence on the plutonium and reduce its solubility. On the credit side, however, is the U233 produced in the blanket. If this is added to the core it may compensate for the in- growth of Pu?® and reduce the Pu?*" requirement to below the solubility limit, and it may be possible to operate indefinitely, as with the U23>- fueled reactors. 14—4. HETEROGENEOUS GRAPHITE-MODERATED RREACTORS The use of a moderator in a heterogencous lattice with molten-salt fuels is potentially advantageous. First, the approach to a thermal neutron spectrum 1mproves the neutron yield, 5, attainable, especially with U235 TaBLE 14-7 INITIAL-STATE NUCLEAR CHARACTERISTICS OF A TypicaL MoLTEN FLUoriDE-SarT REACTOR FueLenp wrtn Pu239 Core diameter: 8 ft. External fuel volume: 339 ft3. Total power: 600 Mw (heat). Load factor: 0.8. Critical inventory: 31.2 kg of Pu?3Y, Critical concentration: 0.013 mole 9 Pu239, Neutrons absorbed per neutrons absorbed in Pu??*® Neutron absorbers Pu?*9 (fissions) 0.630 Pu# (n—y) 0.372 Li% and Li7 in fuel salt 0.202 Be® in fuel salt 0.022 F191n fuel salt 0. 086 Core vessel 0.145 Th in blanket salt - 0.352 Li-Be-F 1n blanket salt 0.024 Reactor vessel 0.004 Leakage 0.003 Neutron yield, 1.84 Thermal fissions, % 59 Regeneration ratio 0.352 (58 NUCLEAR ASPECTS TaBLE 14-8 OF MOLTEN-SALT REACTORS [cHAP. 14 CoMPARISON OF GRAPHITE-MODERATED MOLTEN-SALT AND Liguip-MEeTAL-FUELED *REACTORS LMFR MSFR-1 MSFR-2 Total power, Mw (heat) 580 600 600 Over-all radius, in. 75 75 72 Critical mass, kg of U233 9.9 9.6 27.7 Critical inventory, kg of U233* 467 77.8 213 Regeneration ratio 1.107 0.83 1.07 Core Radius, 1n. 33 33 34.8 Graphite, vol 9 45 45 45 Tuel fluid, vol % 55 DD HH TFuel components, mole % Bi ~100 Lil 69 61 Bels 31 36.5 T}1F4 2 . 5 Unmoderated blanket Thickness, in. 6 6 13.2 Composition, mole 9 Bi 90 Th 10 (Th) | 10 (ThF4) 13 (ThFy) LiF 70 71 Bel's 20 16 233 0.015 0.014 Moderated blanket Thickness, in. 36 36 24 Composition, vol 9% Graphite 66.6 66.6 100 Bianket fluidt 33.4 33.4 Neutron absorption ratiol Th in fuel fluid 0.566 U233 in fuel fluld 0.918 0.925 1.000 Other components of fuel fluid 0.081 0.324 0.106 Th in blanket fluid 1.110 0.825 0.490 1233 in blanket fluid 0.083 0.071 Other compenents of blanket fluid 0.040 0.092 0.038 Leakage 0.012 0.004 0.014 Neutron yield, 9 2.24 2.24 2.21 *With bismuth, the external volume indicated in Ref. 10 was used. The molten- salt systems are calculated for 339 ft* external volumes. tSame as unmoderated blanket fluid. tNeutrons absorbed per neutron absorbed in 17233, 14-4] HETEROGENEOUS GRAPHITE-MODERATED REACTORS 659 and Pu®?”. Second, in a heterogeneous system, the fuel i partially shielded from neutrons of intermediate energy, and a further improvement in ef- fective neutron yield, n, results. Further, the optimum systems may prove to have smaller volumes of fuel in the core than the corresponding fluorine- moderated, homogeneous reactors and, consequently, higher concentrations of fuel and thorium in the melt. This may substantially reduce parasitic losses to components of the carrier salt. On the other hand, these higher concentrations tend to increase the inventory in the circulating-fucl system external to the core. The same considerations apply to fission prod- ucts and to nonfissionable isotopes of uranium. Possible moderators for molten-salt reactors include beryllium, BeO), and graphite. The design and performauce of the Aircraft Reactor Experi- ment, a beryllium-oxide moderated, sodium-zirconium fuoride salt, one- region, U*7-fueled burner reactor has been reported (see Chapter 16). Since beryvllium and BeO and molten salts are not chemically compatible, 1t wax necessary to line the fuel circuit with Inconel. Tt is easily estimated that the presence of Inconel, or any other prospective containment metal in a heterogeneous thermal reactor would seriously impair the regeneration ratio of a converter-breeder. Consequently, beryllium and BeO are elimi- nated from consideration. Preliminary evidence indicates that uranium-bearing molten salts may be compatible with some grades of graphite and that the presence of the graphite will not carburize metallic portions of the fuel circuit seriously [9]. It therefore becomes of interest to explore the capabilities of the graphite- moderated systems. The principal independent variables of interest are the core diameter, fuel channel diameter, lattice spacing, and thorium concentration, 14—4.1 Initial states. Two cases of graphite-moderated molten-salt re- actors have been caleulated for the same geometry and graphite-to-fluid volume ratio as those for the reference-design LMFR [10]. The results for these two cases, together with those for the Hquid bismuth case, are sum- marized in Table 14-8. Only the initial states arc considered, and a metallie shell to separate core and blanket fluids has not been included. With no thorium in the core fluid, the molten-salt-fucled reactor has a significantly lower regeneration ratio than that of the liquid-metal-fueled reactor, with only a slightly lower eritical mass, Adding 2.5 mole ¢ ThF, to the core fluid increases the initial regeneration ratio to about 1.07, with a critical mass and a corresponding total fuel inventory that are acceptably low. 660 NUCLEAR ASPECTS OF MOLTEN-SALT REACTORS [cuap. 14 REFERENCES 1. R. L. MackuiN, Neutron Activation Cross Sections with Sb-Be Neutrons, Phys. Rev. 107, 504-508 (1957). 2. L. G. Avexaxper et al.,, Operating Instructions for the Univac Program Ocusol-1, A Modification of the Eyewash Program, USAEC Report CF-57-6-4, Oak Ridge National Laboratory, 1957. 3. J. H. Avexanper and N. D. Given, 4 Machine Multigroup Calculation. The Eyewash Program for Univac, USAKC Report ORNIL-1925, Oak Ridge National Laboratory, 1955. 4, J. T. Roserts and L. G. AvLeExanprr, Cross Sections for the Ocusol-A Program, USAEC Report CF-57-6-5, Oak Ridge National Laboratory, 1957. 5. B. W. Kinvon, Oak Ridge National Laboratory, 1958, personal communi- cation. 6. W. D. Powgrs, Oak Ridge National Laboratory, 1958, personal communi- cation. 7. L. G. ALExanDpER and L. A. MaNN, First Estimate of the Gamma Heating in the Core Vessel of a Molten Fluoride Converter, USAEC Report CF-57-12-77, Oak Ridge National Laboratory, 1957. 8. C. I. Barton, Solubtility and Stability of PuFg tn Fused Alkali Fluoride- Beryllium Fluoride, USAEC Report ORNL-2530, Oak Ridge National Labora- tory, 1958. 9. F. Kurresz, Oak Ridge National Laboratory, 1958, personal communica- tion. 10. Bascock anp Wincox Co., Liguid Metal Fuel Reactor, Technical Feast- bility Report, USAEC Report BAW-2(Del.), 1955.