CHAPTER 12 CHEMICAL ASPECTS OF MOLTEN-FLUORIDE-SALT REACTOR FUELS* The search for a liquid for use at high temperatures and low pressures in a fluid-fueled reactor led to the choice of either fluorides or chlorides because of the requirements of radiation stability and solubility of appre- clable quantities of uranium and thorium. The chlorides (based on the CI37 isotope) are most suitable for fast reactor use, but the low thermal-neutron absorption cross section of fluorine makes the fluorides a uniquely desirable cholce for a high-temperature fluid-fueled reactor in the thermal or epi- thermal neutron region. Since for most molten-salt reactors considered to date the required con- centrations of UFy and ThEFy have been moderately low, the molten-salt mixtures can be considered, to a first approximation, as base or solvent salt mixtures, to which the fissionable or fertile fluorides are added. For the fuel, the relatively small amounts of Uly required make the correspond- ing binary or ternary mixtures of the diluents nearly controlling with regard to physical properties such as the melting point. 12-1. CHoicE oF BASE OR SOLVENT SALTS The temperature dependence of the corrosion of nickel-base alloys by Huoride salts is deseribed in Chapter 13. From the data given there, 1300°F (704°C 18 taken as an upper limit for the molten-salt-to-metal interface temperature. To provide some leeway for radiation heating of the metal wills and to provide a safety margin, the maximum bulk temperature of the molten-salt fuel at the design condition will probably not exceed 1225°F., Iv « eireulating-fuel reactor, in which heat is extracted from the fuel in an external heat exchanger, the temperature difference between the inlet and outlet of the reactor will be at least {00°I7. The provision of a margin of wifety of 100°1° between minimum operating temperature and melting poitt makes salts with melting points above 1025°F of little interest at present. and therefore this discussion is limited largely to salt mixtures huving melting points no higher than 1022°F (550°C). One of the basic features desired 1In the molten-salt reactor is a low pressure in the fuel svstent, =0 only fluorides with o low vapor pressure at the peak operating temperature ( ~700°C) are considered. *By W. R. Grimes, D. R. Cuneo, F. F. Blankenship, G. W. Keilholtz, H. F. Poppendiek, and M. T. Robinson. 569 570 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHap. 12 Frg. 12-1. The system LiF-NaF-KF [A. G. Bergman and E. . Dergunov, Compt. rend. acad. sci. U.R.S.S., 31, 754 (1941)]. Of the pure fluorides of molten-salt reactor interest, only Bel’s meets the melting-point requirement, and it is too viscous for use in the pure state. Thus only mixtures of two or more fluoride salts provide useful melting points and physical properties. The alkali-metal fluorides and the fluorides of beryllium and zirconium have been given the most serious attention for reactor use. Lead and bis- muth fluorides, which might otherwise be useful because of their low neutron absorption, have been eliminated because they are readily reduced to the metallic state by structural metals such as iron and chromium. Binary mixtures of alkali fluorides that have sufficiently low melting points are an equimolar mixture of KF and Lil", which has a melting point of 490°C, and a mixture of 60 mole 9 RbI with 40 mole 9} Lil’, which has a melting point of 470°C. Up to 10 mole 9, UKy can be added to these alkall fluoride systems without increasing the melting point above the 550°C Iimit. A melting-point diagram for the ternary system Lil'-Nal'-KI, Fig. 12-1, indicates a eutectic with a lower melting point than the melting points of the simple binary Lil'—KF system. This eutectic has interesting properties as a heat-transfer fluid for molten-salt reactor systems, and data on its physical properties are given in Tables 12-1 and 12-2. The KF-LiF and RbF-LiI binaries and their ternary systems with Nal® are the only available systems of the alkali-metal fluorides alone which have TapLe 12-1 MprTivGg Points, HeaT CapAaciTIES, AND EQUATIONS FOR DENSITY AND Viscosity oF Typrcar MoLTEN FLUoORIDES Liquid density, Viscosity, centipoise ¢ it Melting g/ce Heat capacity ‘ "”‘”)l““(;}”“’ point, p=A— BT(°C) at 700°C, n= AeB/TCK) | mole e °C — cal/gram — At 600°C A B A b X 1077 LiF-BeF s (69-31) 05 2.16 40 0.65 0.118 3624 7.5 LiF-Bel's (50-50) 350 2.46 40 0.67 0.0189 6174 22 .2 NaF-Bel’; (H7-43) 3060 2.27 37 052 (.0346 5164 12.8 Nak-7rF, (50-30) 510 3.79 93 (.28 (0.0709 4168 8.4 Lik-NaF-KF (46.5-11.5-42) 454 2.53 73 0.45 0.0400 4170 4.75 LiF-NaF-Bekq (35-27-38) 338 2.22 41 0.59 0.0338 4738 7.8 (121 A0 TOTOTID Asvd SLIIVS LNHATOR HO [LS 572 CHEMICAL ASPECTS. MOLTEN-FLUORIDE-SALT FUELS [CHAP. 12 1000 900 800 Temperatuyre, °C ~ o S o o o 500 — ~r 3 ~ ~ s A ' | SIS 2 ]S 3 400 el e [ e e o o o O o zll |12 |Z z z © vy | ~ o 300 NaF 10 20 30 40 50 60 70 80 %0 zFy ZrF4 . mole % F1a. 12-2. The system NaF-ZrF,. low melting points at low uranium concentrations. They would have utility as special purpose reactor fuel solvents if no mixtures with better properties were available. TaBLE 12-2 TuermaL ConpucTiviTY oF TYPicaL FLUORIDE MIXTURES Thermal conduetivity, Composition, Btu/ (hr) (ft) (°F) mole 9, Solid Liquid LiF-NaF-KF (46.5-11.5-42) 2.7 2.6 NaF-BeF, (57-43) 2.4 Mixtures with melting points in the range of interest may be obtained over relatively wide limits of concentration if ZrF4 or BeF2 is a component of the system. Phase relationships in the NaF-ZrFy system are shown in Iig. 12-2. There is a broad region of low-melting-point compositions that have between 40 and 55 mole 9 ZrF,. 12-1] CHOICE OF BASE OR SOLVENT SALTS 573 900 I | 1 T T 800 . 1 R : - | i | | 700 | 1= T —— | | nU . o 600 Lif + Liquid o ] 2 i 5 ! g & £ 500 e - ‘ BeFp + Liguid 400 = . LipBeF4 e . + Liguid } LiF + LigBeFy 1 T ! f Pt . LioBeF4 + BeF 300 e b 22 TR L — 2 ! L1 LigBeF4 o . : ! 4 E L|iB&F3 + B§F2 | 200 l 1 | LiBeF3 = I [ 1 | LiF 10 20 30 40 50 60 70 80 90 BeF) BeF7 mole % LiBeF3 + Befp Fi1c. 12-3. The system LiF-BeF. i | | 800 — — t L — : | | 700 - — . oe=OmlData ——— |+ | | ¢ a — NayBeF 4 +LIQUID : g 600 - T - : — 2 | | 5 | | S @ i i | L ‘ | | ] - \ : _ | / a— NogBeFy 4 NoF | | 3'~ NaBeFy Bef, + LIQUID ‘ , | A+ 1QuID ‘ 400 f— | A=A A - BeFyp+ 3'—NaBeF3 = | \/ : “/ ’ a— N(_;IZBEFA + 55— NoBeF3 — W \ - — N 5 — NaBeF + LIQUID 300+ - a’— Na,BeF , + NaF 8 — NaBeFq —g————oer f . B:EFQ Tis Nug;F3 . 4 — NaBeF, + — NaBef "— Na,BeF ; _ c;es ¥ . quA\‘\i‘ NasbelF 4 /BeF2+3 NdeF3 200 v — NUQBeF4 + NaF i — T I i NaF 10 20 30 40 50 &0 70 80 90 BeFg BeF,, mole % Fi1g. 12-4. The system NaF-BeF. 574 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 BeFs Dotted Lines Represent 549 Incompletely Defined Phase Boundaries and Alkemade Lines The Symbol TC Represents All Temperatures Are in °C a Compound Whose Exact Composition Hos Not Been Determined 370 NaF-Befg 80 356 iF-Be et -—-‘4\ EE =7l T =X e N 700 - 20 750 (NaF - LiF-BeFo) . 8250 ’ - // / 70 LiF 800 750 700 64§ 700 750 800 850 %00 950 NaF 844 980 F16. 12-5. The system LiF-NaF-BeFo. The lowest melting binary systems are those containing Bel's and LiF or NaI". Since Bel's offers the best cross section of all the useful diluents, fuels based on these binary systems are likely to be of highest interest in thermal reactor designs. The binary system LiF-BeF2 has melting points below 500°C over the concentration range from 33 to 80 mole 9, BeF2. The presently accepted LiF-Bel's system diagram presented in Fig. 12-3 differs substantially from previously published diagrams [1-3]. It is characterized by a single eutectic between Bel'e and 2Lil" - BeF 2 that freezes at 356°C and contains 52 mole T Bel's. The compound 2LiF - BeFz melts incongruently to Lil' and liquid at 460°C; LiF - Bel'z is formed by the reaction of solid BeFs and solid 2LiI" - BeF2 below 274°C. The diagram of the Nal'-BeFy system (Fig. 12-4) is similar to that of the LiI'-Bel'; system. The ternary system combining both NaF and LilF with Bel's, shown in Fig. 12-5, offers a wide variety of low-melting compo- sitions. Some of these are potentially useful as low-melting heat-transfer liquids, as well as for reactor fuels. TaBLE 12-3 MEeLTIiNG Points, HEaT CapaciTies, AND EqQuATIiONs ForR DENSITY AND ViscosiTy oF FUueL BEARING SaLTs Liquid density, Viscosity, centipoise Composition Melting g/cc Heat capacity : mg’le o point, p=A — BTCC) at 700°C, n= AcB/TCK /0 °C - cal/gram At 600°C A B A B X 10—5 LiF-Bel»-UF4 (67-30.5-2.5) 464 2.38 40 0.57 8.4 NaF-BeF 2*~UF4 (55.5-42-2.5) 400 2.50 43 0.46 10.5 NaF-ZrF,-UF, (50-46—4) 520 3.93 93 0.26 0.0981 3895 8.5 [1-g1 SIIVS LNHATOS HO HSVH A0 HDIOHD juby ] 576 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAp. 12 TaBLE 124 TaerMmal ConbuctiviTy oF Typicarn FrLuoripE FurLs Thermal conductivity, Composition, Btu/ (hr)(ft) °F) mole %, Solid Liquid LiF-NaF-KF-UF, (44.5-10.9-43.5-1.1) 2.0 2.3 NaF-ZrF,-UF4 (50-46-4) 0.5 1.3 NaF-ZrF~UF4 (53.5-40-6.5) 1.2 NaF-KF-UF, (46.5-26-27.5) 0.5 UF4 1035 All Temperatures Are in °C 950 NaF-2 UF4 7 NaF-6 u 5 NaF+3 UF4 2 NaF'UFy A 3 NaF-UFy4 750 765 850 ANZrFy \ 918 7 NaF-6 ZrFy NaF-ZrFy 200 5NaF-2ZrF4 2NaF-ZiFg 3 NaF-2 ZrFy %0 3 NaF-ZrFy Fia. 12-6. The system NaF-ZrF;-UF4. 12-2] FUEL AND BLANKET SOLUTIONS 577 UF4 All Temperatures Are in °C E-- Eutectic P:: Peritectic LiF- 4UFy /... UF |- Primary Phase Field P P00 7LF6UF 4 \ VF | 0 ALiF- UFy O\ 5 - > B |7LiF6UF, 05 A2 © N1 300 650 ; CPOO _'}O“gd‘ \ 450 - 600~—— =T NN o : =V LiF 2LiF-BeF, P 400°g_ 400 BeF Fig. 12-7. The system LiF-BeFo-UF,. 12-2. FueL AND BLANKET SOLUTIONS 12-2.1 Choice of uranium fluoride. Uranium hexafluoride is a highly volatile compound, and it is obviously unsuitable as a component of a liquid for use at high temperatures. The compound UO2Fs, which is rela- tively nonvolatile, is a strong oxidant that would be very difficult to con- tain. Fluorides of pentavalent uranium (Ul's,Usl', ete.) are not thermally stable [4] and would be prohibitively strong oxidants even if they could be stabilized in solution. Uranium trifluoride, when pure and under an inert atmosphere, 1s stable even at temperatures above 1000°C [4,5]; however, it is not so stable in molten fluoride solutions [6]. It disproportionates appreciably in such media by the reaction 4 UFs == 3 UF;:+ U, at temperatures below 800°C. Small amounts of UF3 are permissible in the presence of relatively large concentrations of UF4 and may be beneficial insofar as corrosion 18 concerned. It is necessary, however, to use UF4 as the major uraniferous compound in the fuel. 578 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 UFy4 All Temperatures Are in °C E -=Eutectic P- Peritectic :Primury Phase Field 7NaF- éUF4 25 5NafF- 3UF4 800 7NaF- 6UF4 5NaF-3UF4 . ) . 4 . ) E E E BeFg 2NqF'8eF2/ \NaF ‘BeFy F1G. 12-8. The system NaF-Bel2-UF,. 12-2.2 Combination of UF, with base salts. The fuel for the Aircraft Reactor Experiment (Chapter 16) was a mixture of UF4 with the NaF-ZrI'4 base salt. The ternary diagram for this system is shown in Fig. 12-6. The compounds ZrF4 and UF4 have very similar unit cell parameters [4] and are isomorphous. They form a continuous series of solid solutions with a minimum melting point of 765°C for the solution containing 23 mole 9 UF4. This minimum is responsible for a broad shallow trough which pene- trates the ternary diagram to about the 45 mole 9, Nal' composition. A continuous series of solid solutions without a maximum or a minimum exists between a—3Nal - UFy and 3Nal’ - ZrF4; in this solution series the temperature drops sharply with decreasing ZrFs concentration. A con- tinuous solid-solution series without a maximum or a minimum also exists between the Isomorphous congruent compounds 7NaF -6UFs and 7Nal - 6ZrF4; the liquidus decreases with increasing Zrl'y content. These two solid solutions share a boundary curve over a considerable composition range. The predominance of the primary phase fields of the three solid solutions presumably accounts for the complete absence of a ternary eutectic in this complex system. The liquidus surface over the area below 8 mole 9, UI'4 and between 60 and 40 mole 9, NaF is relatively flat. All fuel compositions within this region have acceptable melting points. Minor 12-2] FUEL AND BLANKET SOLUTIONS 579 ThF, 1080°C \060 1000 7 950° 900° 850° 800 ) Oo 550°C %%, 750 I, I G- a 0,0 2N\ A\ N 2 TN b, 502N LiF AN\ = BeF 845°C ligBeF, LiBeFg (?) 543°C 475°C 360°C F1g. 12-9. The system LiF-BeF:-ThF,. advantages in physical and thermal properties acerue from choosing mix- tures with minimum ZrI's content in this composition range. Typical physical and thermal properties are given in Tables 12-3 and 12-4. The nuclear studies in Chapter 14 indicate that the combination of BeFs with NaF or with LiF (provided the separated Li? isotope can be used) are more suitable as reactor fuels. The diagram of Fig. 12-7 reveals that melting temperatures below 500°C can be obtained over wide com- position ranges in the three-component system Lil'-Bel's-UF4. The lack of a low-melting eutectic in the NaI'-UI'4 binary system is responsible for melting points below 500°C being available over a considerably smaller concentration interval in the NaF-Bel's-UF4 system (Fig. 12-8) than in its LiF—Bels—-UF4 counterpart. The four-component system LiF-NalF-BeF.-UF4; has not been com- pletely diagrammed. It is obvious, however, from examination of Fig. 12-5 that the ternary solvent LiF-Nal'-Bel's offers a wide variety of low-melting compositions; it has been established that considerable quantities (up to at least 10 mole 9;) of UF4 can be added to this ternary system without elevation of the melting point to above 500°C. 580 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 7 LiF-6 ThFy LiF-2 ThF4 T E=Eutectic P—Peritectic Liquidus Temperatures Are in °C LiF 0 25 2 LiF-BeF; BeFy fmole %) Fic. 12-10. The system LiF-BeF.-ThFy in the concentration range 50 to 100 mole 9, LiF. 12-2.3 Systems containing thorium fluoride. All the normal compounds of thorium are quadrivalent; accordingly, any use of thorium in molten fluoride melts must be as ThF4. A diagram of the LiF-BeFe—ThTF ternary system, which i1s based solely on thermal data, is shown as Fig. 12-9. Recent studies in the 50 to 100 mole 9 LiF' concentration range have demonstrated (Fig. 12-10) that the thermal data are qualitatively correct. Breeder reactor blanket or breeder reactor fuel solvent compositions in which the maximum ThF4 concentration is restricted to that available in salts having less than a 550°C liquidus may be chosen from an area of the phase diagram (I'ig. 12-10) in which the upper limits of ThI'4 concentra- tion are obtained in the composition 75 mole 9, LiF-16 mole 9, ThF4-9 mole 9, BeFs,, 69.5 mole 9, LiF-21 mole 9, ThF4+-9.5 mole 9, BeFs, 68 mole 9, LiF-22 mole 9, ThI'+-10 mole 9, Bel's. 12-2.4 Systems containing Thy and UF;. The LiF-Bel'>-UF4 and the LiF-BeFo—ThF4 ternary systems are very similar; the two eutectics in the LiF-BelF3—ThI'y system are at temperatures and compositions virtually identical with those shown by the UF4-bearing system. The very great 12-3] PROPERTIES OF FLUORIDE MIXTURES 581 similarity of these two ternary systems and preliminary examination of the Lil'-Bel"'s>-ThF4-UF4 quaternary system suggests that fractional re- placement of UF4 by ThIy will have lLittle effect on the freezing tem- perature over the composition range of interest as reactor fuel. 12-2.5 Systems containing PuF;. The behavior of plutonium fluorides i molten fluoride mixtures has received considerably less study. Plu- tonium tetrafluoride will probably prove very soluble, as have Uy and Thl'y, in suitable fluoride-salt diluents, but is likely to prove too strong an oxidant to be compatible with presently available structural alloys. The trifluoride of plutonium dissolves to the extent of 0.25 to 0.45 mole 9, in LiI'-Bel’s mixtures containing 25 to 50 mole ¢, Bel's. As indicated in Chapter 14, it is believed that such concentrations are in excess of those required to fuel a high-temperature plutonium burner. 12-3. Puysicar anp THeErMAL ProPERTIES OF FLUORIDE MIXTURES The melting points, heat capacities, and equations for density and vis- cosity of a range of molten mixtures of possible interest as reactor fuels are presented above in Tables 12-1 and 12-3, and thermal-conductivity values are given in Tables 12-2 and 12-4; the methods by which the data were ob- tained are described here. The temperatures above which the materials are completely in the liquid state were determined in phase equilibrium studies. The methods used included (1) thermal analysis, (2) differential- thermal analysis, (3) quenching from high-temperature equilibrium states, (4) visual observation of the melting process, and (5) phase separation by filtration at high temperatures. Measurements of density were made by weighing, with an analytical balance, a plummet suspended in the molten mixture. Iiuthalpies, heats of fusion, and heat capacities were determined from measurements of heat liberated when samples in capsules of Ni or Inconel were dropped from various temperatures into calorimeters; both ice calorimeters and large copper-block calorimeters were used. Measure- ments of the viscosities of the molten salts were made with the use of a capillary efflux apparatus and a modified Brookfield rotating-cylinder device; agreement between the measurements made by the two methods indicated that the numbers obtamed were within 4 1095. Thermal conductivities of the molten mixtures were measured in an apparatus similar to that deseribed by Lucks and Deem [7], in which the heating plate is movable so that the thickness of the liquid specimen can be varied. The uncertainty in these values is probably less than 4 25%. The variation of the thermal conductivity of a molten fluoride salt with temperature is relatively small. The conductivities of solid fluoride mix- tures were measured by use of a steady-state technique in which heat was passed through a solid slab. H82 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 The vapor pressures of PuF; [8], UF4 [9], and ThF4 are negligibly small at temperatures that are likely to be practical for reactor operations. Of the fluoride mixtures likely to be of interest as diluents for high-temperature reactor fuels, only AlF3, BeF» [9], and ZrF4 [10-12] have appreciable vapor pressures below 700°C. Measurements of total pressure in equilibrium with NaF-ZrF,~UF, melts between 800 and 1000°C with the use of an apparatus similar to that described by Rodebush and Dixon [13] yielded the data shown in Table 12-5. Sense et al. [14], who used a transport method to evaluate partial TaBLE 12-5 Varor Pressures oF FrLuoripE MixTUres CoNTAINING ZRF4 COIIEEI(; s1(1;7;0n, Vapor pressure constants* Vapor pressure at 900°C, mm Hg NaF ZrFy UF, A B X103 100 7.792 9.171 0.9 100 12 542 11.360 617 57 43 7.340 7.289 14 50 50 7.635 7.213 32 50 46 4 7.888 7.551 28 53 43 4 7.37 7.105 21 *For the equation log P (mm Hg) = A — (B/T), where T is in °K, pressures in the NaF-ZrF4 system, obtained slightly different values for the vapor pressures and showed that the vapor phase above these liquids is quite complex. The vapor-pressure values obtained from both investi- gations are less than 2 mm Hg for the equimolar NaF-ZrF4 mixture at 700°C. However, since the vapor is nearly pure ZrF4, and since ZrF4 does not melt under low pressures of its vapor, even this modest vapor pressure leads to engineering difficulties; all lines, equipment, and connections ex- posed to the vapor must be protected from sublimed ZrF4 “snow.” Measurements made with the Rodebush apparatus have shown that the vapor pressure above liquids of analogous composition decreases with in- creasing size of the alkali cation. All these systems show large negative deviations from Raoult’s law, which are a consequence of the large, posi- tive, excess, partial-molal entropies of solution of ZrF,. This phenomenon has been interpreted qualitatively as an effect of substituting nonbridging TaBLE 12-6 Varor Pressures oF NaF-BeFs MixTures* Composition, Vapor pressure constantsf mole % Temperature Vapor pressure interval, NaF Bel', NaF - BeFs at 800°C, °C mm Hg Nak BekF, A B A B A B x 104 X 10% x 101 26 74 785-977 10.43 1.096 9.77 1.206 1.69 41 59 802988 1006 1.085 9.79 1.187 0.94 50 50 796-996 9.52 1.071 9 82 1.187 0 41 60 40 855-1025 9 392 1.1667 9.080 1.1063 (.09 75 25 857-1035 9.237 1.2175 8.2 1.12 0.02 *Compiled from data obtained by Sensc et al. [15]. tFor the equation log /> (mm Hg) = A — (B/T), where 7' is in °K. SHUAIXIN JAIH0NTd A0 STILIHdONd [e-2T €8¢ H84 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHaP. 12 fluoride ions for fluoride bridges between zirconium ions as the alkali fluoride concentration is increased in the melt [12]. Vapor pressure data obtained by the transport method for NaF-BeF, mixtures [15] are shown in Table 12-6, which indicates that the vapor phases are not pure BeFs;. While pressures above LiF-BeFz must be ex- pected to be higher than those shown for NaF-BeF: mixtures, the values of Table 12-6 suggest that the “snow” problem with BeFo mixtures is much less severe than with ZrFs melts. Physical property values indicate that the molten fluoride salts are, in general, adequate heat-transfer media. It is apparent, however, from vapor pressure measurements and from spectrophotometric examination of analogous chloride systems that such melts have complex structures and are far from ideal solutions, 12-4. PropucTtoN AND PURIFICATION OF FLUORIDE MIXTURES Since commercial fluorides that have a low concentration of the usual nuclear poisons are available, the production of fluoride mixtures is largely a purification process designed to minimize corrosion and to ensure the removal of oxides, oxyfluorides, and sulfur, rather than to improve the neutron economy. The fluorides are purified by high-temperature treat- ment with anhydrous HF and Ho gases, and are subsequently stored in sealed nickel containers under an atmosphere of helium. 12-4.1 Purification equipment. A schematic diagram of the purification and storage vessels used for preparation of fuel for the Aircraft Reactor Experiment (Chapter 16) is shown in Fig. 12-11. The reaction vessel in which the chemical processing is accomplished and the receiver vessel into which the purified mixture is ultimately transferred are vertical cylindrical containers of high-purity low-carbon nickel. The top of the reactor vessel 1s pierced by a charging port which is capped well above the heated zone by a Teflon-gasketed flange. The tops of both the receiver and the reaction vessels are pilerced by short risers which terminate in Swagelok fittings, through which gas lines, thermowells, etc., can be introduced. A transfer line terminates near the bottom of the reactor vessel and near the top of the receiver; entry of this tube is effected through copper-gasketed flanges on l-in.-diameter tubes which pierce the tops of both vessels. This transfer line contains a filter of micrometallic sintered nickel and a sampler which collects a specimen of liquid during transfer. Through one of the risers in the receiver a tube extends to the receiver bottom; this tube, which is sealed outside the vessel, serves as a means for transfer of the purified mixture to other equipment. This assembly is connected to a manifold through which He, Hs, HF, or vacuum can be supplied to either vessel. By a combination of large tube 124] PURIFICATION OF FLUORIDE MIXTURES 085 N e~ | 3/8-in. Nickel | » Transfer Line U . Filter (S Reaction Vessel Receiver Vessel/ Fia. 12-11. Diagram of purification and storage system. furnaces, resistance heaters, and lagging, sections of the apparatus can be brought independently to controlled temperatures in excess of 800°C. 12—4.2 Purification processing. The raw materials, in batches of proper composition, are blended and charged into the reaction vessel. The material is melted and heated to 700°C under an atmosphere of anhydrous HF to remove Ho(O with a minimum of hydrolysis. The HF is replaced with Ho for a period of 1 hr, during which the temperature is raised to 800°C, to reduce U5t and U®* to U4 (in the case of simulated fuel mixtures), and sulfur compounds to S~ and extraneous oxidants (Fet**, for example) to 5806 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHar, 12 lower valence states. The hydrogen, as well as all subsequent reagent gases, is fed at a rate of about 3 liters/min to the reaction vessel through the re- ceiver and transfer line and, accordingly, it bubbles up through the molten charge. The hydrogen is then replaced by anhydrous HF, which serves, during a 2- to 3-hr period at 800°C, to volatilize HeS and HCI and to con- vert oxides and oxyfluorides of uranium and zirconium to tetrafluorides at the expense of dissolution of considerable NiF; into the melt through re- action of HF with the container. A final 24- to 30-hr treatment at 800°C with Hy suffices to reduce this NiFe and the contained FeF. to soluble metals. At the conclusion of the purification treatment a pressure of helium above the salt in the reactor vessel is used to force the melt through the transfer line with its filter and sampler into the receiver. The metallic iron and nickel are left in the reactor vessel or on the sintered nickel filter. The purified melt is permitted to freeze under an atmosphere of helium in the receiver vessel. 12-5. RADIATION STABILITY OF FLUORIDE MIXTURES When fission of an active constituent occurs in a molten fluoride solu- tion, both electromagnetic radiations and particles of very high energy and intensity originate within the fluid. Local overheating as a consequence of rapid slowing down of fission fragments by the fluid is probably of little consequence in a reactor where the liquid is forced to flow turbulently and where rapid and intimate mixing occurs. Moreover, the bonding in such liquids 1s essentially completely ionic. Such a solution, which has neither covalent bonds to sever nor a lattice to disrupt, should be quite resistant to damage by particulate or electromagnetic radiation. More than 100 exposures to reactor radiation of various fluoride mix- tures containing UF4 in capsules of Inconel have been conducted; in these tests the fluid was not dehberately agitated. The power level of cach test was fixed by selecting the U235 content of the test mixture. Thermal neu- tron fluxes have ranged from 10V to 10™ neutrons/(cm?)(sec) and power levels have varied from 80 to 8000 w/cm?. The capsules have, in general, been exposed at 1500°F for 300 hr, although several tests have been con- ducted for 600 to 800 hr. A list of the materials that have been studied is presented in Table 12-7. Methods of examination of the fuels after irra- diation have included (1) freezing-point determinations, (2) chemical analysis, (3) examination with a shielded petrographic microscope, (4) as- say by mass spectrography, and (5) examination by a gamma-ray spectro- scope. The condition of the container was checked with a shielded metal- lograph. No changes in the fuel, except for the expected burnup of U?32 have been observed as a consequence of irradiation. Corrosion of the Inconel 12-5] RADIATION STABILITY OF FLUORIDE MIXTURES H&87 TaBLE 12-7 MovreN Savts WaicH Have BEEN Stupiep iN IN-PiLe CaprsurLk TESTS . Composition, System mole NaF-KF-UF, 46.5-26-27.5 Nal-BeFo-UFy 25-60-15 NaF-BeFo-UF4 47-51-2 NaF-BeF-UF, 50-46-4 NalF-ZrF4UF, 63-25-12 NaF-ZrF4«-UF, 53.5-40-6.5 NalF-7ZrF~UF, 50-48-2 Nal-ZrF—UF; 50-48-2 TaBLE 12-8 DescripTions oF INncoNEL ForceED-CircuraTioN Looprs OperaTep IN THE LITR anp tae MTR Loop designation LITR LITR MTR Horizontal Vertical Horizontal NaF-ZrF,~UF4 composition, mole §5 62.5-12.5-25 | 63-25-12 | 53.5-40-6.5 Maximum fission power, w/cm? 400 500 800 Total power, kw 2.8 10 20 Dilution factor™® 180 7.3 b5 Maximum fuel temperature, °F 1500 1600 1500 Fuel temperature differential, °F 30 250 155 Fuel Reynolds number 6000 3000 5000 Operating time, hr 645 332 467 Time at full power, hr 475 235 271 *Ratio of volume of fuel in system to volume of fuel in reactor core. 588 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [CHAP. 12 capsules to a depth of less than 4 mils in 300 hr was found; such corrosion is comparable to that found in unirradiated control specimens [16]. In capsules which suffered accidental excursions in temperatures to above 2000°F, grain growth of the Inconel occurred and corrosion to a depth of 12 mils was found. Such increases in corrosion were almost certainly the result of the serious overheating rather than a consequence of the radiation field. Tests have also been made in which the fissioning fuel is pumped through a system in which a thermal gradient is maintained in the fluid. These tests included the Aireraft Reactor Iixperiment (deseribed in Chapter 16) and three types of forced-circulation loop tests. A large loop, in which the pump was outside the reactor shield, was operated in a horizontal beam hole of the LITR.* A smaller loop was operated in a vertical position in the LITR lattice with the pump just outside the lattice. A third loop was operated completely within a beam-hole of the MTR.7 The operating con- ditions for these three loops are given in Table 12-8. The corrosion that occurred in these loop tests, which were of short duration and which provided relatively small temperature gradients, was to a depth of less than 4 mils and, as in the capsule tests, was comparable to that found in similar tests outside the radiation field [16]. Therefore it is concluded that within the obvious limitations of the experience up to the present time there is no effect of radiation on the fuel and no accelera- tion of corrosion by the radiation field. 12-6. BEgavior orF FisstoN Propucts When fission of an active metal occurs in a molten solution of its fluo- ride, the fission fragments must originate in energy states and ionization levels very far from those normally encountered. These fragments, how- ever, quickly lose energy through collisions in the melt and come to equi- librium as common chemiecal entities. The valence states which they ulti- mately assume are determined by the necessity for cation-anion equivalence in the melt and the requirement that redox equilibrium be established among components of the melt and constituents of the metallic container. Structural metals such as Inconel in contact with a molten fluoride so- lution are not stable to Fo, Ul's, or Ul's. It 1s clear, therefore, that when fission of uranium as UFy4 takes place, the ultimate equilibrium must be such that four cation equivalents are furnished to satisfy the fluoride ions released. Thermochemical data, from which the stability of fission-product fluorides in complex dilute solution could be predicted, are lacking in *Low Intensity Test Reactor, a tank type research reactor located at Oak Ridge, Tennessee. fMaterials Testing Reactor, a tank type research reactor located at Arco, Idaho. 12-6] BEHAVIOR OF FISSION PRODUCTS H89 many cases. No precise definition of the valence state of all fission-product fluorides can be given; it is, accordingly, not certain whether the fission process results in oxidation of the container metal as a consequence of de- positing the more noble fission products i the metallic state. 12-6.1 Fission products of well-defined valence. The noble gases. The fission products krypton and xenon can exist only as elements. The solubilitiecs of the noble gases in NaF-ZrFs (53-47 mole 9) [17], NaF-ZrF4~UF4 (50-46—-1 mole %) [17], and LiF-Nal-KF (46.5-11.5-42 mole %) obey Henry's law, increase with increasing temperature, decrease with increasing atomic weight of the solute, and vary appreciably with composition of the solute. The Henry’s law constants and the heats of solution for the noble gases in the NaF-ZrF, and LiF-NaF-KF mixtures are given in Table 12-9. The solubility of krypton in the NakF-ZrF4 mix- ture appears to be about 3 X107% moles/(cm?)(atm). TasLE 12-9 SoLUBILITIES AT (G00°C aAxp HEATS OF SOLUTION FOR NOBLE Gases 1y MovLreN FrLuoripes MIXTURES 5 In NaF-7rF4 In LiF-NaF-KF (53-47 mole &, (46.5-11.5-42 mole %) Gas Heat of Heat of K* solution, K* solution, keal/mole keal /mole X108 X108 Helium 21.6 +£1 6.2 11.34£0.7 8.0 Neon 11.3 £0.3 7.8 4.440.2 8.9 Argon 5.1 £0.15 8.2 Xenon 1.94+0.2 11.1 *Henry’s law constant in moles of gas per cubic centimeter of solvent per atmosphere, The positive heat of solution ensures that blanketing or sparging of the fuel with helium or argon in a low-temperature region of the reactor cannot lead to difficulty due to decreased solubility and bubble formation in higher temperature regions of the system. Small-scale in-pile tests have revealed that, as these solubility data suggest, xenon at low concentration is re- tained in a stagnant melt but is readily removed by sparging with helium. Only a very small fraction of the anticipated xenon poisoning was observed 590 CHEMICAL ASPECTS. MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 during operation of the Aireraft Reactor Experiment, even though the sys- tem contained no special apparatus for xenon removal [18]. It seems certain that krypton and xenon isotopes of reasonable half-life can be readily removed from all practical molten-salt reactors. Elements of Groups [-A, II-A, I11-B, and IV-B. The fission products Rb, Cs, Sr, Ba, Zr, Y, and the lanthanides form very stable fluorides; they should, accordingly, exist in the molten fluoride fuel in their ordinary valence states. High concentrations of ZrFy and the alkali and alkaline earth fluorides can be dissolved in LiF-NaF-KF, LiI'2-BeFs, or Nal'-ZrF, mixtures at 600°C. The solubilitiés at 600°C of Y5 and of selected rare- earth fluorides in NaF-ZrF, (53-47 mole 9 ) and LiF-BeF 2 (65-35 mole %) are shown in Table 12-10. For these materials the solubility increases TasLe 12-10 SoLUBILITY oF YF3 axp or SoME Rare-FEarta FLUroripEs IN Nal-ZrFy axp 1x LaF-BulFs ar 600°C i | Solubility, mole ¢, MFE; Fluoride . | . - In Nalb 7rk, i In LifF-Bel, i (57—43 mole ©) g (62-38 mole Yp) | _ YT 3.6 f LaFg ‘ 2.1 “ Cng ‘ 2.3 .48 smkEs | 2.5 | about 0.5% /°C' and increases slightly with increasing atomic number in the lanthanide series; the saturating phase is the simple trifluoride. For solutions containing more than one rare earth the primary phase is a solid solution of the rare-carth trifluorides; the ratio of rare-earth cations in the molten solution is virtually identical with the ratio in the precipitated solid solution. Quite high burnups would be required before a molten fluoride reactor could saturate its fuel with any of these fission products. 12-6.2 Fission products of uncertain valence. The valence states as- sumed by the nonmetallic elements Se, Te, Br, and I must depend strongly on the oxidation potential defined by the container and the fluoride melt, and the states are not at present well defined. The sparse thermochemical data suggest that if they were in the pure state the fluorides of Ge, As, Nb, Mo, Ru, Rh, Pd, Ag, Cd, Sn, and Sh would be reduced to the cor- responding metal by the chromium in Inconel. While fluorides of some of 12-7] FUEL REPROCESSING 591 these elements may be stabilized in dilute molten solution in the melt, it is possible that none of this group exists as a compound in the equilibrium mixture. An appreciable, and probably large, fraction of the niobium and ruthenium produced in the Aireraft Reactor Experiment was deposited in or on the Inconel walls of the fluid circuit; a detectable, hut probably small, fraction of the ruthenium was volatilized, presumably as RulF'5, from the melt. 12-6.3 Oxidizing nature of the fission process. The fission of a mole of UF4 would yield more equivalents of cation than of anion if the noble gas isotopes of half-life greater than 10 min were lost and if all other elements formed fluorides of their Towest reported valence state. If this were the case .the system would, presumably, retain cation-anion equivalence by reduction of fluorides of the most noble fission products to metal and perhaps by reduction of some U*™ to U3*T. If, however, all the elements of uncertain valence state listed in Article 12-6.2 deposit as metals, the balance would be in the opposite direction. Only about 3.2 equivalents of combined cations result, and since the number of active anion equivalents i= & minimum of 4 (from the four fluorines of UF4), the deficiency must he alleviated by oxidation of the container. The evidence from the Aireraft Reactor Experiment, the in-pile loops, and the in-pile capsules has not shown the fission process to cause serious oxidation of the container; it is possible that these experiments burned too little uraniurm to yield significant results. If fission of UF4 1s shown to be oxidizing, the detrimental effect could be overcome by deliberate and occasional addition of a reducing agent to create a small and stable concentration of soluble UF3 in the fuel mixture. 12-7. FUuEL REPROCESSING Numerous conventional processes such as solvent extraction, selective precipitation, and preferential ion exchange could be readily applied to molten fluoride fuels after solution in water. However, these liquids are readily amenable to remote handling and serve as media in which chemical reactions can be conducted. Most development efforts have, accordingly, been concerned with direct and nonaqueous reprocessing methods. Recovery of uranium from solid fuel elements by dissolution of the element in a fluoride bath followed by application of anhydrous HF and subsequent volatilization of the uranium as UFg has been described [19,20]. The volatilization step accomplishes a good separation from Cs, Sr, aud the rare earths, fair separation from Zr, and-poor separation from Nb and Ru. The fission products I, Te, and Mo volatilize completely from the melt. The nonvolatile fission products are discarded in the fluoride solvent. Further decontamination of the UFy is effected by selective ab- 592 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 sorption and desorption on beds of NaF. At 100°C, UFg is absorbed on the bed by the reversible reaction UFs (g) + 3NaF—=3NaF - UFg, which was first reported by Martin, Albers, and Dust [21]. Niobium ac- tivity, along with activity attributable to particulate matter, is also absorbed; ruthenium activity, however, largely passes through the bed. Subsequent desorption of the UFg4 at temperatures up to 400°C is accom- plished without desorption of the niobium. The desorbed UFy is passed through a second NaF bed held at 400°C as a final step and is subsequently recovered in refrigerated traps. The decontaminations obtained are greater than 10° for gross beta and gamma emitters, greater than 107 for Cs, Sr, and lanthanides, greater than 10° for Nbh, and about 10% for Ru. Uranium was recovered from the molten-salt fuel of the Aircraft Reactor Experiment by this method, and its utility for molten-fluoride fuel systems or breeder blankets was demonstrated. Recovery of plutonium or thorium, however, is not possible with this process. There are numerous possible methods for reprocessing molten-salt fuels. The behavior of the rare-earth fluorides indicates that some decontamina- tion of molten-fluoride fuels may be obtained by substitution of Cel’; or Lals, in a sidestream circuit, for rare earths of higher cross section. It seems likely that Pul’y can be recovered with the rarc-earth fluorides and subsequently separated from them after oxidation to Pul’y. [Further, it appears that both selective precipitation of various fission-product ele- ments and active constituents as oxides, and selective chemisorption of these materials on solid oxide beds are capable of development into valu- able separation procedures. Only preliminary studies of these and other possible processes have heen made. 593 REFERENCES 1. D. N. Roy et al., Fluoride Model Systems: [V, The Systems LiF—BeF, and RbFe—BeFg, J. Am. Ceram. Soc. 37, 300 (1954). 2, A. V. Novosrrova et al,, Thermal and X-ray Analysis of the Lithium- Beryllium Fluoride System, J. Phys. Chem. USSR 26, 1244 (1952). 3. W. R. GrimEs et al., Chemical Aspects of Molten Fluoride Reactors, paper to be presented at Second International Conference on Peaceful Uses of Atomic Energy, Geneva, 1958, 4, J. J. Karz and E. RasiNowrrcH, The Chemistry of Uranium, National Nuclear Energy Series, Division VIII, Volume 5. New York: MeGraw-Hill Book Co., Inec., 1951. 5. W. R. Grimes et al., Oak Ridge National Laboratory. Unpublished, 6. . H. CravprrT et al., Oak Ridge National Laboratory, 1957. Unpublished. 7. C. F. Lucks and H. W. DrrM, Apparatus for Measuring the Thermal Conductwity of Liquid at Elevated Temperatures; Thermal Conductivity of Fused NaOH to 600°C, Am. Soc. of Mech. Eng. Meeting, June 1956. (Preprint 56SA31) 8. G. T. Seapora and J. J. Karz (Eds.), The Actitnide Elements, National Nuclear Energy Series, Division [V, Volume 14A, New York: MceGraw-Hill Book Co., Inc., 1953. 9. W. R. Grimes et al., Fused-salt Systems, Sec. 6 in Reactor Handbook, Vol. 2, Engineering, USAEC Report AECD-3646, 1955. (pp. 799-850) 10. K. A. SENsE et al.,, The Vapor Pressure of Zirconium Fluoride, J. Phys. Chem, 58, 995 (1954). 11. W. 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Maxvy et al., Metallurgical Problems in Molten Fluoride Systems, paper to be presented at Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958. 17. W. R. Grimes ct al,, Solubility of Noble Gases in Molten Fluorides. [. In Mixtures of NabF—ZrFy (53-47 mole 9) and NabF—ZrF4+—UF4 (50-46-4 mole 93), J. Phys. Chem. (in press). 594 CHEMICAL ASPECTS: MOLTEN-FLUORIDE-SALT FUELS [cHAP. 12 18. E. 8. BErris et al., The Aircraft Reactor Experiment—Qperation, Nuclear Sci. and Eng. 2, 841 (1957). 19. G. I. CaraErs, Uranium Recovery for Spent Fuel by Dissolution in Fused Salt and Fluorination, Nuclear Sci. and Eng. 2, 768 (1957). 20. F. R. Bruce et al., in Progress in Nuclear Energy, Series 1lI, Process Chemistry, Vol. I. New York: McGraw-Hill Book Co., Inc., 1956. 21. Von H. MarTIN et al.,, Double Fluorides of Uranium Hexafluoride, Z. anorg. u. allgem. Chem. 265, 128 (1951).