MOLTEN-SALT REACTOR CHEMISTRY
W. R. GRIMES Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 KEYWORDS: fused salt fuel, chemical reactions, reactors, beryllium fluorides, zirconium fluorides, Ii th i um fluorides, Received August 4, 1969 uranium hexafluoride, repro Revised October 7, 1969 cessing, protactinium, seporo tion pro c es s es, breeding, fission products, MSRE
aggressive toward some otherwise suitable ma terial of construction and toward some suitable This document summarizes the large program of chemical research and development which led moderator material. The fuel must be stable to selection of fuel and coolant compositions for toward reactor radiation, must be able to survive the Molten-Salt Reactor Experiment (MSRE) and fission of the uranium (or other fissionable ma for subsequent reactors of this type. Chemical terial} and must tolerate fission product accumu lation without serious deterioration of its useful behavior of the LiF-BeFrZrFrUF4 fuel mixture and behavior of fission products during power op properties. We must also be assured of a gen uinely low fuel-cycle cost; this presupposes a eration of MSRE are presented. A discussion of low-cost fuel associated with inexpensive turn the chemical reactions which show promise for around of the unburned fissile material, and recovery of bred 233Pa and for removal of fission effective and economical schemes for recovery of product poisons from a molten-salt breeder reac tor is included. bred fissile material and for removal of fission product poisons from the fuel. A suitable secondary coolant must be provided to link the fuel circuit with the steam-generating equipment. The demands imposed upon this cool ant fluid differ in obvious ways from those im INTRODUCTION posed upon the fuel system. Radiation intensities will be markedly less in the coolant system, and A single-fluid molten-salt thermal breeder the consequences of uranium fission will be ab (MSBR) of the type described by Rosenthal et al.,1 sent. The coolant salt must, however, be com Bettis and Robertson,2 and Perry and Bauman3 patible with metals of construction which will makes very stringent demands upon its fluid handle the fuel and the steam; it must not undergo 4 6 violent reactions with fuel or steam should leaks fuel. - This fuel must consist of elements having low capture cross sections for neutrons typical of develop in either circuit. The coolant should be the energy spectrum of the chosen design. The inexpensive, possessed of good heat-transfer prop fuel must dissolve more than the critical concen erties, and it should melt at temperatures suit tration of fissionable material (235U 233U or able for steam cycle start-up. An ideal coolant 239 Pu), and high. concentrations. of fertile' material ' would consist of compounds which would be easy 232 to separate from the valuable fuel mixture should ( Th) at temperatures safely below the tempera ture at which the salt leaves the MSBR heat they mix as a consequence of a leak. exchanger. The mixture must be thermally stable, This report presents, in brief, the basis for and its vapor pressure needs to be low over an choice of fuel and coolant systems which seem operating temperature range ( 1100 to 1400°F) optimum in light of these numerous-and to some sufficiently high to permit generation of high extent conflicting-requirements. quality steam for power production. The fuel mixture must possess heat transfer and hydro CHOICE OF FUEL COMPOSITION dynamic properties adequate for its service as a Compounds which are permissible major con heat-exchange fluid. It must be relatively non- stituents of fuels for single-fluid thermal breeders
NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 137 Grimes MSBR CHEMISTRY are those which can be prepared from beryllium, The only good moderator material truly compati bismuth, boron-11, carbon, deuterium, fluorine, ble with molten-fluoride fuel mixtures is graph lithium-7, nitrogen-15, oxygen, and the fissionable ite. 4- 6 and fertile materials. As minor constituents one might tolerate compounds containing the other Phase Behavior Among Fluorides elements in Table I. Uranium tetrafluoride and uranium trifluoride Many chemical compounds can be prepared are the only fluorides (or oxyfluorides) of uranium from the several "major constituents" listed which appear useful as constituents of molten above, Most of these, however, can be eliminated fluoride fuels. Uranium tetrafluoride (UF 4) is after elementary consideration of the fuel re relatively stable, nonvolatile, and largely non quirements. 4-6 No hydrogen- (or deuterium-) hygroscopic. It melts at l035°C (1895°F), but this bearing compounds possess overall properties freezing point is markedly depressed by useful that are practical in such melts. Carbon, nitro diluent fluorides. Uranium trifluoride dispropor gen, and oxygen form high melting binary com tionates at temperatures above ~ l000°C by the pounds with the fissionable and fertile metals; reaction these compounds are quite unsuitable as constit ( 1) uents of liquid systems. The oxygenated anions 6 7 either lack_ the required thermal stability (i.e., It is unstable ' at lower temperatures in most N03- or N02-) or fail as solvents for high concen molten-fluoride solutions and is tolerable in reac trations of thorium compounds (i.e., C03=). It tor fuels only with a large excess of UF4 so that quickly develops, therefore, that fluorides are the the activity of u° is so low as to avoid appreciable only suitable salts indicated in this list of ele reaction with moderator graphite or container ments. metal. Fluoride ion is capable of appreciable neutron Thorium tetrafluoride (ThF 4) is the only known moderation, but this moderation is by itself in fluoride of thorium. It melts at llll°C (2032°F) sufficient for good neutron thermalization. An but fortunately its freezing point is markedly de additional moderator is, accordingly, required. pressed by fluoride diluents which are also useful with UF4. Consideration of nuclear properties alone leads TABLE I one to prefer as diluents the fluorides of Be, Bi, 7 Elements or Isotopes Which may be Tolerable Li, Mg, Pb, and Zr in that order. Equally simple 8 9 in High-Temperature Reactor Fuels consideration of the stability of these fluorides ' toward reduction by structural metals, however, Absorption Cross Section eliminates the bismuth fluorides from considera Material (barns at 2200 m/sec) tion. This leaves BeF2 and 7 LiF as the preferred diluent fluorides. Phase behavior of systems Nitrogen-15 0.000024 based upon LiF and BeF2 as the major constitu Oxygen 0.0002 10 Deuterium 0.00057 ents, has, accordingly, been examined in detail. Carbon 0.0033 Fortunately for the molten fluoride reactor con Fluorine 0.009 cept, the phase diagrams of LiF-BeF2-UF4 and Beryllium 0.010 LiF-BeF2-ThF4 are such as to make these ma Bismuth 0.032 terials useful as fuels. Lithium-7 0.033 The binary system LiF-BeF2 has melting points Boron-11 0.05 below 500°C over the concentration range from 33 10 11 Magnesium 0.063 to 80 mole% BeF 2. ' The phase diagram, pre Silicon 0.13 sented in Fig. 1, is characterized by a single Lead 0.17 eutectic (52 mole% BeF2, melting at 360°C) be Zirconium 0.18 tween BeF2 and 2LiF·BeF2. The compound Phosphorus 0.21 2LiF·BeF2 melts incongruently to LiF and liquid Aluminum 0.23 at 458°C. LiF·BeF2 is formed by the reaction of Hydrogen 0.33 solid BeF2 and solid 2LiF·BeF2 below 280°C. 0 11 Calcium 0.43 The phase behavior of the BeF2-UF/ • and 12 Sulfur 0.49 Be Fr ThF4 systems are very similar. Both sys Sodium 0.53 tems show simple single eutectics containing very Chlorine-37 0.56 small concentrations of the heavy metal fluoride. Tin 0.6 ThF4 and UF4 are isostructural; they form a con Cerium 0.7 tinuous series of solid solutions with neither Rubidium 0.7 maximum nor minimum.
138 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY
i :::~
7001--~~-t--~~--1[~~~~-+~~~+-~~--t-~~--;~~~--r--~~--r~~~r-~~~ E600'--~-L-iF~+~L-IQ_U_ID~---+--~·--'~~~~~+-~~--+-~~----;~~~--+-~~--r~~~r-~~~ ~ ' ~~ ~ ~ _J.----f__ _, ~ 500f--~~-+-~~----j~~~~~-,-\-~--r~~~t-~~-t-~--:::....-ri....-::-:--'---~~-t--~~--j--~~--j 458 l ~ i-----+------i---+-.....,-._...... v BeF (HIGH QUARTZ TYPE) L 2 ~ +LIQUID -1
400 1----1---"'--i-h..l"~--+----+------i------36_0--fl I Li 2BeF.i + BeF2 (HIGH QUARTZ TYPE l 300~~·~-+-~~-+-+,_-.::.-.::.-.::.-:::.'.::..-.::.-.::.-.::.-.::.-.::.:'=------J+----~J---~....---28_0_,I ~<;t mN Li2B•F4 "' LiBeF3 + BeF2 (HIGH QUARTZ TYPE) ..J + ~ Li8eF3 + BeF2 (LOW QUARTZ TYPE~ j 220 Li BeF3 ::::; J. " 200'----'-----''----_.____.__~---~--~--~---~-.-0.-~--~l Li F 10 20 30 40 50 60 70 80 90 BeF2 BeF2 (mole%)
Fig. 1. The system LiF-BeF2 •
13 14 2 The binary diagrams LiF - UF4 and LiF-ThF4 3LiF·ThF4 can incorporate Be + ions in both are generally similar and much more complex interstitial and substitutional sites to form solid than the binary diagrams discussed immediately solutions whose compositional extremes are rep above. The LiF-UF4 system shows three com resented by the shaded triangular region near pounds (none are congruently melting) and a single that compound. Liquidus temperatures < 550°C eutectic, at 27 mole% UF4, melting at 490°C. The (1022°F) are available at ThF4 concentrations as LiF-ThF4 system contains four binary compounds, high as 22 mole%. The maximum ThF 4 concen one of which (3LiF·ThF4) melts congruently, with tration available at liquidus temperatures of 500°C two eutectics at 570°C and 22 mole% ThF4 and at (932°F) is seen to be just above 14 mole%. Inspec 560°C and 29 mole% ThF4. tion of the diagram reveals that a considerable 15 The ternary system LiF-ThF4- UF4, shown in range of compositions with > 10 mole% ThF4 will Fig. 2, shows no ternary compounds and a single be completely molten at or below 500°C. eutectic freezing at 488°C with 1.5 mole% ThF4 As expected from the general similarity of and 26.5 mole% UF4. Most of the area on the dia ThF4 and UF4-and especially from the substitu gram is occupied by the primary phase fields of tional behavior shown by the LiF-UF4-ThF4 sys the solid solutions UF4-ThF4, LiF·4UF4-LiF·4ThF4, tem (Fig. 2)-substitution of a small quantity of and LiF·UFrLiF·ThF4. Liquid us temperatures UF4 for ThF4 scarcely changes the phase be decrease generally to the LiF-UF4 edge of the havior. Accordingly, and to a very good approxi diagram. mation, Fig. 3 represents the behavior of the The single-fluid molten-salt breeder fuel will LiF-BeF2-ThF4-UF4 system over concentration need a concentration of ThF4 much higher than regions such that the mole fraction of ThF4 is that of UF 4, Accordingly, the phase behavior of much greater than that of UF4. the fuel will be dictated by that of the LiF-BeF2- ThF4 system. Figure 3 gives the ternary system Oxide Fluoride Equilibria LiF-BeF2-ThF4; this system shows a single ter nary eutectic at 47 mole% LiF and 1.5 mole% ThF4, Phase behavior of the pure fluoride system 10 11 melting at 360°C. ' The system is complicated LiF-BeF2-ThF4-UF4, as indicated above, is such to some extent by the fact that the compound that adequate fuel mixtures seem assured. The
NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 139 Grimes MSBR CHEMISTRY
PRIMARY-PHASE AREAS TEMPERATURE IN °C (a) UF -ThF.i (ssl COMPOSITION IN mole "lo 4 (b) LiF·4UF -LiF·4ThF (ss) 4 4 (cl LiF·2Th(UlF4 (ssl (d) 7LiF·6UF4-7LiF·6ThF4 (ss) (e) 3LiF·Th(U)F4 (ssl (fl LiF
LiF ,___..__.:L---'----'---lo"---'----41111--"'------"----""-'~~-__.._---'-''-"----'~~--->----~-~ UF4 848 1035
Fig. 2. The system LiF-ThF4-UF+ behavior of systems such as this, however, is When a mixture of LiF and BeF 2 containing markedly affected by appreciable ctmcentrations ThF4 and UF 4 is treated with a reactive oxide a of oxide ion. homogeneous cubic phase is produced; this phase When a melt containing only LiF, BeF 2, and is a solid solution of U02 and Th02 which is very UF4 is treated with a reactive oxide (such as H20) rich in U02. 17 Careful studies have shown that the precipitation of transparent ruby crystals of U02.oo reaction 6 7 results. ' If the melt contains, in addition, an TT4 + 4+ (2) appreciable concentration of ZrF4 the situation is Th02(ss) + U(f) '°" U02(ss) + Th(f) markedly altered. Zr0 is less soluble than is 2 [where the subscripts (ss) and (f) indicate the U02 in such melts, and the monoclinic Zr02 (the solid-phase solid solution and the molten-fluoride form stable below ~ 1125 °C) includes very little solution, respectively] approaches equilibrium with U02 in solid solution. Thus, inadvertent oxide reasonable speed. Values for the equilibrium contamination of a LiF-BeF2-ZrF -UF4 melt 4 quotient Q for this reaction yields monoclinic Zr02 containing 250 ppm of U02. 16 Precipitation of cubic U02 (containing a 4+ N VO 2(ss) N Th (f) small concentration of Zr02) begins only after (3) precipitation of Zr02 had dropped the ZrF4 con Q=------NThO 2(ss) 0 N4u(o centration to near that of the UF4· Slow precipitation of U02 followed by a sudden entrance of this material into the reactor core increase with U02 concentration of the oxide phase could result in undesired increased in reactivity. and decrease markedly with temperature. Since This possibility was assumed to represent a dan values of Q for mixtures similar to those chosen ger to the Molten- Salt Reactor Experiment. Ac as fuel compositions are typically 300 to 1000, it cordingly, the MSRE fuel was chosen to contain 5 is clear that oxide contamination of such salts will mole% of ZrF4 to eliminate such a possibility. selectively precipitate the uranium.
140 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY
Th~ H11
TEMPERATURE IN °C COMPOSITION IN mole "lo
LiF·Th~ LiF·2Th~ 3LFThF 4 ss
E 360
Fig. 3. The system LiF-BeF2-ThF4.
It is likely, though not certain, that addition of the type used for MSRE fuels can be successfully some ZrF 4 would afford protection of the sort ob applied on a large scale to the LiF-BeF2-ThF4- tained with the MSRE fuel. Such addition is un UF4 system. desirable, however, since the presence of ZrF4 would certainly complicate the separation pro MSRE and MSBR Fuel Compositions cesses described later in this paper and else where in this series. The fuel chosen for operation of MSRE (with a The successful operation of the MSRE over a 235U- 238 U isotopic mixture containing 33% of the three year period (discussed later) lends con fissionable isotope) was a mixture of 7 LiF, BeF2, fidence that oxide contamination of the fuel system ZrF4, and UF4 consisting of 65, 29.1, 5, and 0.9 can be kept to adequately low levels. This confi mole%, respectively. The uranium concentration dence, when added to the prospect that the breeder was fixed at ~ 1% so that there was less possibility 235 fuel will be reprocessed (and its oxide level re of fissile U precipitating (~ 0.3 mole% U was duced) at regular intervals, suggests very strongly necessary to achieve criticality and to provide a that successful operation can be achieved without small excess of fissionable material for power added ''oxide protection.'' operation of the machine). The ZrF4 was added, Tolerance levels for oxide concentration in as indicated above, to preclude possible inad LiF-BeF2-UF4 and LiF-BeF2-ZrF4-UF4 systems vertent precipitation of U02. Beryllium fluoride have been studied in detail and are relatively well is an extremely viscous material; its viscosity is 16 understood. • l8-ro Analogous values for the LiF markedly lowered by addition of LiF. The ratio of BeF 2-ThF4-UF 4 system are still largely lacking. LiF to BeF2 in the MSRE fuel was chosen to opti It is known, however, that processing of these mize the conflicting demands for low viscosity and quaternary melts with anhydrous HF and H2 serves a low liquidus temperature for the molten fuel. to remove oxide to a level below that required for The single-fluid breeder requires a high con precipitation of the solid solutions. There seems centration of ThF4; cone en t rations near 12 little doubt, therefore, that initial processing of mole% seem to be reasonable for good reactor
NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 141 Grimes MSBR CHEMISTRY performance. Criticality estimates suggest that nous fluoride (SnF 2 ) melts at 212°C. This ma such a fuel could be made critical in a practicable terial is probably not stable during long term 233 reactor with somewhat < 0.3 mole% UF4. The service in Hastelloy-N; moreover, its phase dia ratio of 7 LiF to BeF2 should, to decrease vis grams with stable fluorides (such as NaF or LiF) cosity, be kept at a value as high as is practicable. show high melting points at relatively low alkali If the liquidus temperature is to be kept at or fluoride concentrations. Coolant compositions below 500°C (932°F) for a melt with 12 mole% of which will meet the low liquidus temperature spec ThF4, the beryllium concentration limits range ification may be chosen from the NaF-BeF2 or from 16 to 25 mole%. The most likely choice for NaF-LiF-BeF2 system. These materials are al the MSBR fuel-and the present design composi most certainly compatible with Hastelloy-N, and tion-is, accordingly, 7 LiF-BeF2-ThF4-UF4 at they possess adequate specific heats and low 71. 7-16-12-0. 3 mole%, respectively. vapor pressures (discussed in the next section). They (especially those including LiF) are moder Choice of Coolant ately expensive, and their viscosities at low temperature are certainly higher than desirable. The secondary coolant is required to remove It is possible that substitution of ZrF4 or even heat from the fuel in the primary heat exchanger AlF3 for some of the BeF2 would provide liquids and to transport this heat to the power generating of lower viscosity at no real expense in liquidus system. In the MSBR the coolant must transport temperature. heat to supercritical steam at minimum tempera It now appears that the best choice for the tures only modestly above 700°F. In the MSRE the MSBR secondary coolant is the eutectic mixture of heat was rejected to an air cooled radiator at sodium fluoride and sodium fluoroborate. The markedly higher salt temperatures. binary system NaF-NaBF4 has been described as The coolant mixture chosen for the MSRE and showing a eutectic at 60 mole% NaBF 4 melting at 21 22 shown to be satisfactory in that application is 304 °C (580°F). ' Studies here have shown that BeF2 with 66 mole% of 7 LiF. Use of this mixture this publication is seriously in error. Boric oxide would pose some difficulties in design of equip substantially lowers the freezing point of NaF ment for the MSBR since its liquidus temperature NaBF4 mixtures; the original authors may well is 851°F; moreover, it is an expensive material. have used quite impure materials. Use of pure The eutectic mixture of LiF with BeF2 (48 NaF and NaBF4 leads to the conclusion that the mole% LiF) melts at near 700°F (see Fig. 1) but it system shows a single eutectic at 8 mole% NaF 23 is relatively viscous and is expensive, especially and a melting point of 385°C (725°F), as shown in if 7 LiF is used. Fig. 4. The alkali metals, excellent coolants with real At elevated temperatures the fluoroborates promise in other systems, are undesirable here show an appreciable equilibrium pressure of gas since they react vigorously with both fuel and eous BF3. The equilibrium pressure24 above a steam. Less noble metal coolants such as Pb0 or Bi0 undergo no violent reactions, but they are not compatible with Hastelloy-N, the Ni-based alloy lOOO~~,----,----,---,----,----,----r---r----~ used in MSRE and intended for use in MSBR's. i-r-..__ 1 i Several binary chloride systems are known to 900 f---1--fl'------""+=---+--l--+----J.-i __J------lJ----i have eutectics melting below (in some cases much 1 21 1~.1 f below) 700°F. These binary systems do not, p ~Or------t-~1t-l--+-~~i----'~"'K:.------ie------+----+----t------1 however, appear especially attractive since they contain high concentrations of chlorides (TlCl, - 700 r--~r------t--+--+--+---+-~+---!-----+---< ZnCb, BiCIJ, CdCb, or SnCb), which are easily ~ I '1' reduced and, accordingly, corrosive; or chlorides ~ soo r------1'------+--+--+--+---+--+G~.+---+----< (AlC!J, ZrC4, HfCl4, or BeCb) which are very ~ I ",, volatile. The only low-melting binary systems of ~ 500 r------1----+--+--+--+---+--+---L\.~---+---< stable, non-volatile chlorides are those containing ~ 400 ~ LiCl; LiCl-CsCl (330°C at 45 mole% CsCl), LiCl ....____,~-+~-•-~__.._~-·-~~--~- ~-~ KCl (355 °C at 42 mole% KCl), LiCl-RbCl (3 l2°C at 300 f------1--+--+--+--+--+---+---!---+-- 45 mole% RbCl). Such systems would be relatively expensive if made from 7 LiCl, and they could lead 200 L_____J_ __J__ __J__ _J__ _J__ _J___J__~L____J_ __ to serious contamination of the fuel if normal LiCl NoF 20 40 60 NoBF4 NoBF (mole%) were used. 4 Very few fluorides or mixtures of fluorides are known to melt at temperatures below 370°C. Stan- Fig. 4. The system NaF-NaBF4 •
142 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY melt maintained at the eutectic composition (8 TABLE ill mole% NaF, 92 mole% NaBF4 ) is given by Composition and Properties of Possible MSBR Secondary Coolants 5,920 c, C2 c, log PTorr = 9.024 - T oK (4) Composition (mole%! NaF 8 LiF 23 NaBF. 92 NaF 41 NaF 57 PHYSICAL PROPERTIES OF FUELS BeF2 36 BeF2 43 AND COOLANTS Liquidus Temperature: Tables II and III list some of the pertinent phys oc 385 328 340 ical properties24 for MSRE and MSBR fuels and OF 725 622 644 secondary coolants. Physical Properties at Many of the properties shown are estimates 850°F (454°C)" rather than measured values. These estimates Density, lb/ft3 121 136 139 have been carefully prepared from the best avail Heat capacity, able measurements on several salt mixtures of Btu/(lb °F) 0.36 0.47 0.44 similar composition. The values given are un Viscosity, cP 2.5 40 65 likely to be in error sufficient to remove the fluid Vapor pressure at from consideration. It is clear, however, from the 1125°F (607°C),b mm 200< Negligible Negligible fact that estimates rather than measured values Thermal conductivity, are shown that an experimental program must be W/(°C cm) 0.005 0.01 0.01 devoted to firming up the physical properties of "Mean temperature of coolant going to the primary heat exchanger. these materials. bHighest normal operating temperature of coolant. The densities were calculated from the molar NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 143 Grimes MSBR CHEMISTRY Thermodynamic Data for Molten Fluorides For the case where the total uranium in the mol ten solution of 0. 9 mole% (as in the MSRE) the A continuing program of experimentation over activity of metallic uranium is near 10- 15 with many years has been devoted to definition of the 1% of the UF4 converted to UF3 and is near 10 27 thermodynamic properties of many species in 2 x 10- with 20% of the UF4 so converted. Op molten LiF-BeF2 solutions. Many techniques have eration of the reactor with a small fraction been used in this study. Many of the data have (usually < 2%) of the uranium present as UF3 is been obtained by direct measurement of equilib advantageous insofar as corrosion and the conse rium pressures for reactions such as quences of fission are concerned (see subsequent sections of this report). Such operation with some H2(g) + FeF2(d).= Fe?c) + 2HF(g) (5) UF3 present should result in formation of an and extremely dilute (and experimentally undetectable) alloy of uranium with the surface of the container metal. Operation with > 50% of the uranium as UF3 would lead to much more concentrated (and [where (g), (c), and (d) represent gas, crystalline highly deleterious) alloying and to formation of solid, and solute, respectively] using the molten uranium carbides. Fortunately, all evidence to fluoride as reaction medium. Baes has reviewed date demonstrates that operation with relatively all these studies and by combining the data with the little UF3 is completely satisfactory. work of others has tabulated thermodynamic data 26 for many species in molten 2LiF.BeF2• TableIV below records pertinent data for the major com Oxidation (Corrosion) of Metals ponents of MSRE and MSBR fuels and for cor The data of Table IV reveal clearly that in rosion products in molten 2LiF·BeF2 • From these data one can assess the extent to reactions with structural metals (M) which UF3-bearing melt will disproportionate ac 2UF4(d)+ M(c).=2UF3(d)+ MF2(dJ , (8) cording to the reaction 4UF3(d) .= 3UF4(d) + U . (7) Cr is much more readily attacked than is iron, nickel, or molybdenum. 27 Few thermodynamic data exist for the fluorides of titanium in the pure state and none for dilute solutions in molten 8 9 fluoride solvents. Estimates • of free energies of formation suggest that Ti is somewhat more reac TABLE IV tive than Cr. Titanium on the surface layers of Standard Free Energies of Formation for Species the metal should, therefore, be expected to react in Molten 2LiF · BeF2 with the available oxidants, such as UF4. Such (773 to l000°K) oxidation and selective attack follow from reac tions such as the following: -t>.Gf -AG /(1000°K) Impurities in the melt Material a (kcal/mole) (kcal/mole) Cr + NiF2-+ CrF2 +Ni (9) 3 LiF (1) 141.8 -16.6 x 10- T°K 125.2 or 3 BeF2(1) 243.9 - 30.0 x 10- T°K 106.9 3 Cr+ 2HF -+CrF2 + H2 (10) UF3(d) 338.0 -40.3 x 10- T°K 99.3 3 UF4(d) 445.9 - 57 .9 x 10- T°K 97.0 Oxide films on the metal 3 ThF4(d) 491.2 - 62.4 x 10- T°K 107.2 3 ZrF4(d) 453.0-65.1x10- T°K 97.0 NiO + BeF2-+ NiF2 + BeO , (11) 3 NiF2(d) 146.9-36.3x10- T°K 55.3 followed by reaction of NiF2 with Cr. 3 FeF2 (d) 154. 7 - 21.8 x 10- T°K 66.5 Reduction of UF4 to UF3 3 CrF2(d) 171.8-21.4x10- T°K 75.2 3 Cr + 2UF4 .= 2UF3 + CrF2 (12) MoFs(g) 370.9 - 69.6 x 10- T°K 50.2 Reactions (9), (10), and (11) above will proceed essentially to completion at all temperatures aThe standard state for LiF and BeF2 is the molten within the MSBR circuit. Accordingly, such reac 2LiF · BeF2 liquid. That for MoFs (g) is the gas at one atmosphere. That for all species with subscript (d) is tions can lead (if the system is poorly cleaned) to that hypothetical solution with the solute at unit mole a noticeable, rapid initial corrosion rate. How fraction and with the activity coefficient it would have ever, these reactions do not give a sustained at infinite dilution. corrosive attack. 144 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY The reaction of UF4 with Cr, on the other hand, experimental study of this is much less extensive. has an equilibrium constant with a small tempera The free energy change for the chemical reaction ture dependence; hence, when the salt is forced to circulate through a temperature gradient, a pos BF3(gl + 3/2Cr(sl-> 3/2CrF2(1) + B(sl (13) sible mechanism exists for mass transfer and is ~ +30 kcal at 800°K. 29 The reaction is, there continued attack. fore, quite unlikely to occur, and similar reactions If nickel, iron, and molybdenum are assumed to with Fe, Mo, and Ni are much less so. In addition, be completely inert diluents for chromium (as is the above reaction becomes even less likely (per approximately true), and if the circulation rate in haps by 5 kcal or so) when one considers the the MSBR is very rapid, the corrosion process can energetics of formation of the compound NaBF be simply described. At high flow rates, uniform 4 and dilution of the NaBF by NaF. However, re concentrations of UF3 and CrF2 are maintained 4 actions which produce metal fluorides and metal throughout the fluid circuit; these concentrations borides are those which must be anticipated. Ti satisfy (at some intermediate temperature) the tanium, a minor constituent of Hastelloy-N, has a equilibrium constant for the reaction. Under these quite stable boride.30 The reaction steady-state conditions, there exists some tem perature (intermediate between the maximum and 2BF3(g) + %Ti(c);=e TiB2(c)+ fTiF4(d) (14) minimum temperatures of the circuit) at which the initial surface composition of the structural metal probably shows a small negative free energy is at equilibrium with the fused salt. Since the change and must be expected to proceed. The very equilibrium constant for the chemical reaction in small diffusion rate of Ti in Hastelloy-N would be creases with increasing temperature, the chromi expected to markedly limit the reaction. Thermo um concentration in the alloy surface tends to chemical data for the borides of Fe, Cr, Ni, and decrease at temperatures higher than T and tends Mo do not seem to have been established. Unless to increase at temperatures lower than T. [In the borides of these materials are very stable some melts (NaF-LiF-KF-UF4, for example) D..G (D.G f values more negative than - 25 kcal B atom) for the mass transfer reaction is quite large, and the Hastelloy-N should prove resistant to pure the equilibrium constant changes sufficiently as a NaF-NaBF4 coolant. function of temperature to cause formation of dendritic chromium crystals in the cold zone.] Compatibility of Graphite with Fluorides For MSBR fuel and other LiF-BeF2-UF4 mixtures, the temperature dependence of the mass-transfer Graphite does not react with molten fluoride reaction is small, and the equilibrium is satisfied mixtures of the type to be used in the MSBR. 8 9 at reactor temperature conditions without the for Available thermodynamic data ' suggest that the mation of crystalline chromium. Thus, in the most likely reaction MSBR, the rate of chromium removal from the (15) salt stream by deposition at cold-fluid regions is controlled by the slow rate at which chromium should come to equilibrium at CF4 pressures diffuses into the cold-fluid wall; the chromium < 10-a atm. CF4 concentrations over graphite-salt concentration gradient tends to be small, and the systems maintained for long periods at elevated resulting corrosion is well within tolerable limits. temperatures have been shown to be below the Titanium, which must be presumed to undergo limit of detection (< 1 ppm) of this compound by similar reactions, diffuses less readily than does mass spectrometry. Moreover, graphite has been Cr in Hastelloy-N. It seems most unlikely that the used as a container material for many NaF-ZrF4- 1% of Ti in the alloy will prove detrimental. UF4, LiF-BeF2-UF4, and other salt mixtures with The results of numerous long-term tests have no evidence of chemical instability. shown that Hastelloy-N has excellent corrosion resistance to molten-fluoride mixtures at temper CHEMICAL BEHAVIOR IN MSRE atures well above those anticipated in MSBR.25 The attack from mixtures similar to the MSBR fuel at The Molten-Salt Reactor Experiment was, as 31 hot-zone temperatures as high as 1300°F is barely detailed in an article by Haubenreich and Engel observable in tests of as long as 12 000 h. A fig operated during much of 1966, 1967, and early 28 1968 with the original fuel charge in which the ure of < 0.5 mil/year is expected. In MSRE, 235 attack was ~ 0.1 mil/year. Thus, corrosion of the uranium was 33% U with the balance consisting 238 container metal by the reactor fuel does not seem primarily of U. The reactor accumulated nearly 25 to be an important problem in the MSBR. 70 000 MWh of operation during this interval with It is likely that the NaF-NaBF4 coolant mixture the major fraction of this accumulated at the max will also prove compatible with Hastelloy-N, but imum power of ~ 8 MW(th). NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 145 Grimes MSBR CHEMISTRY Behavior of Fuel Components discrepancy in LiF and BeF2 concentration re mains a puzzle, but there was nothing in the an Samples of the molten salts were removed alyses (or in the behavior of the reactor) to routinely from the fuel and coolant circuits and suggest that any changes occurred. were analyzed for uranium, major fuel constitu Oxide concentration in the radioactive fuel salt ents, possible corrosion products, and (less fre were performed by careful evaluation of H2 0 pro quently) for oxide ion contamination. Early in the duced upon treatment of the salt samples with power runs, standard samples of fuel were drawn anhydrous gaseous HF. All samples examined three times per week; this schedule was markedly showed less than 100 ppm of o=; no perceptible diminished as confidence in the system grew. At increase in the values with time or with opera late stages in the power runs, analyses for these tional details was apparent.34 These facts are items was generally done on a one-per-week quite reassuring insofar as maintaining oxide basis. The LiF-BeF2 coolant system was sampled contamination at very low levels in future reactor every two weeks initially and less frequently dur systems is concerned. ing 1967 and 1968. MSRE maintenance operations have necessi Chemical determinations of uranium content of 32 tated flushing the interior of the drained reactor the fuel salt were run by coulometric titrations of circuit on numerous occasions. The salt used for dilute aqueous solutions. Such analyses showed this operation consisted originally of a 7 LiF-BeF2 good reproducibility and high precision (±0. 5%). (66.0 to 34.0 mole%) mixture. Analysis of this salt (On-site reactivity balance calculations proved to before and after each use shows that ~ 215 ppm of be about ten-fold more sensitive than this in es uranium is added to the flush salt in each flushing tablishing changes in uranium concentrations with operation; this corresponds to removal of 22. 7 kg in the fuel circuit.) Agreement of the chemically of fuel-salt residue (~ 0.5% of the charge) from determined uranium concentrations (in a statis the reactor circuit. tical sense) with the reactor inventory values diminished by the uranium burnup during opera Behavior of Corrosion Products tion was excellent. The inventory value at end of operation on 235U showed the uranium concentra 33 Analysis of many samples at regular intervals tion to be 4. 532% by weight. The difference (200 during reactor operation showed relatively high g of uranium out of 220 kg) is < 0.1 %. These data values for iron (120 ppm) and nickel (50 ppm) in strongly indicate that uranium losses (as to the the salt.33 -35 These values scattered significantly purge gas stream) have been extremely small and but showed no perceptible trends. They do not that, as anticipated, the fuel salt is chemically seem to represent dissolved Fe2+ or Ni2+. Molyb stable during long periods of reactor operation. A denum concentrations were shown, in the few further check on these numbers will be possible 235 238 attempts made, to be below the detectible limit when the U- U mixture removed by fluori (~ 25 ppm) for chemical analysis. nation from the MSRE fuel is recovered and The chromium concentration, as determined by assayed. chemical analysis, rose from an initial value near While analyses for UF4 and for ZrF4 agree 40 ppm to a final value after 70 000 MWh of quite well with the inventory data, the values for ~ 85 ppm. A considerable scatter (± 15 ppm) in 7 LiF and BeF2 have never done so; analyses for the numbers was apparent but the increase with LiF have shown higher and those for BeF2 have time and reactor operation is clearly real. All shown lower values than the inventory since evidence suggests that the analytically determined start-up. Table V shows a comparison of typical chromium was largely, if not entirely, present as 2 analysis with the original inventory• value. This dissolved Cr +. This observed increase in chro mium concentration corresponds to removal of < 250 g of this element from the MSRE circuit. If TABLE V this were removed uniformly it would deplete the chromium in the ailoy to a depth of 0. 2 mil. Typical and Original Composition < of MSRE Fuel Mixture Such an estimate of corrosion seems consistent with observations (reported elsewhere in this Original Value series) showing very slight attack by the fuel Constituent (mole%) Typical Analysis during MSRE operation. This virtual absence of corrosion-though in 7 LiF 63.40 ± 0.49 64.35 general accord with results from a variety of en BeF2 30.63 ± 0.55 29.83 gineering corrosion tests-is mildly surprising ZrF4 5.14 ± 0.12 5.02 since the MSRE was operated for appreciable UF4 0.821 ± 0.008 0.803 periods with the fuel salt more oxidizing than 146 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY necessary. The analysis for ratio of u4+ to u3 + must occur or the container must supply the de has proved especially difficult in the radioactive ficiency. Operation of MSRE has indicated that the salts, but it seems certain that the MSRE fuel has, presence of UF3 in relatively low concentrations at times, contained < 0. 2% of the total uranium as (see preceding sections) suffices to avoid oxida tf+. tion or corrosion from this source. The fission process (see following section of Behavior of the fission products in MSRE was 36 40 this report) appears to be mildly oxidizing toward studied - during the entire period of operation 235 dissolved lf+ in the fuel. Accordingly, and al with U. Samples of MSRE fuel, drawn from the though the corrosion product analysis did not sampling station in the MSRE pump bowl, have seem to require it, a convenient means was de been routinely analyzed for numerous fission prod veloped for adding if+ to the reactor circuit as uct species. Since samples drawn in open metal desired. Beryllium metal (as 3-in. rods of i-in. cups were shown (for reasons described below) to diameter encased in a perforated capsule of give erroneously high values for gas-borne spe nickel) was introduced through the sampling sys cies, evacuated bulbs of Ni sealed with a fusible tem and suspended in the flowing salt stream in plug of salt were used to obtain many of the data the MSRE pump bowl. This active metal reacts presented here. with UF4 by Samples of the helium gas purge within the MSRE were obtained by use of identical evacuated Be~cl + 2UF4(dl __, BeF2(d)+ 2UF3(dl (16) sample bulbs in which the fusible plug was per mitted to melt with the bulb exposed to the gas at a rate such that some 600 g of UF3 is produced phase. 37 ,3a during an 8-h treatment. The salt flowing past the In addition to these salt and gas samples, as Be appears to be locally over-reduced so that at semblies of surveillance specimens were exposed least 6% of the UF4 becomes UF3; crystalline Cr as the central stringer within the MSRE core. was observed upon the nickel capsule from one These assemblies included specimens of various such reduction attempt. The over-reduced salt graphites and many specimens of Hastelloy-N and mixture clearly reacted and achieved equilibrium some other metals. Their removal at reactor with the large excess of unreduced salt in the shutdowns permitted study of fission product de pump bowl and reactor circuit. positions after 8000, 20 000, and 70 000 MWh of A molten-salt reactor which included a repro operation. cessing circuit could use that external circuit to Samples of salt removed from the pump bowl maintain the UF .JUF3 ratio at the desired leveL were dissolved in analytical chemistry hot cells Use of techniques such as the Be addition for such (after careful leaching of activities from the out a reactor would seem to be quite feasible but side of capsules where necessary); the fission unnecessary. products were identified and their quantity de termined radiochemically. Analysis of the gas Behavior of Fission Products samples was accomplished similarly after careful When fission of UF4 occurs in a well-mixed leaching of the activities from within the capsules. high-temperature molten-fluoride, four fluoride Deposition of fission products upon the metallic ions associated with the u4+ are released and the surveillance specimens was determined in the two fission fragments must come quickly to a same way after repeated leachings of the metal steady state as common chemical entities. This surface. steady state is made very complicated by the Deposition on the graphite specimens was es radioactive decay of many species. The valence tablished as a function of depth of penetration. states that the fission product assume in the mol This was accomplished by careful milling of suc ten system are, presumably, defined by the re cessive thin layers from each surface of the rec quirements that cation-anion equiv a 1 enc e be tangular specimens. The removed layers were maintained in the liquid and redox equilibria be separately recovered and analyzed for several 38 40 established both among the components of the melt fission product isotopes. - 6 and the surface layers of the container metal. 4- The fission gases Kr and Xe form no com Fluorine and uranium species higher than 4 are, pounds under conditions existing in a molten-salt 41 in MSRE and MSBR fuels, unstable toward reduc reactor, and these elements are only very spar 42 44 tion by the active constituents of Hastelloy-N; the ingly soluble in the molten fluoride. - The fission product cations must satisfy the four helium purge gas introduced into the MSRE pump fluoride ions plus the fission product anions. bowl serves to strip these gases from the in Should they prove inadequate, or if they become coming salt to charcoal filled traps downstream in adequate only by assuming oxidation states incom the exit gas system. This stripping ensures that patible with Hastelloy-N, then oxidation of UF3 the fission product daughters of these gases will NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 147 Grimes MSBR CHEMISTRY appear within the reactor system at lower than the TABLE VI theoretical concentration. The MSRE graphite is, Fission Product Deposition on Surface* of course, permeable to Kr and Xe; radioactive of MSRE Graphite daughters of these gases are, accordingly, ex pected in the graphite specimens. Graphite Location The fission products Rb, Cs, Sr, Ba, the lan thanides and yttrium, and Zr all form quite stable Top Middle Bottom and well-known fluorides which are soluble in Percent of Percent of Percent of 45 46 Isotope Total a Total a Total a MSRE fuel. • These fission products were ex pected (except to the extent that volatile precursor 99 Mo 13.4 17.2 11.5 Kr and Xe had escaped) to be found almost com 132Te 13.8 13.6 12.0 pletely in the molten fuel. These expectations 103Ru 11.4 10.3 6.3 have been confirmed. Isotopes such as 95 Zr, 91Sr, 95Nb 12 59.2 62.4 and 143Ce which have no precursors of consequence 131I 0.16 0.33 0.25 are typically found in the salt at 90+3 of the cal 9szr 0.33 0.27 0.15 144Ce culated quantity. Isotopes such as 89Sr and 140 Ba 0.052 0.27 0.14 00 sr 3.24 3.30 2.74 whose Kr and Xe precursors have appreciable 140Ba 1.38 1.85 1.14 half-lives are found in less than the calculated 141Ce 0.19 0.63 0.36 quantity; these isotopes are, as expected, found in 137Cs 0.07 0.25 0.212 the graphite specimens and the gas samples as well. *Average of values of 7- to 10-mil cuts from each of The elements (molybdenum, niobium, ruthen three exposed graphite faces. ium, and tellurium) whose higher valence fluorides a Percent of total in reactor deposited on graphite if each 2 6 2 are volatile and relatively unstable toward re cm of the 2 x 10 cm of moderator had the same con 8 19 29 30 47 duction by UF3 • ' , • are virtually absent from centration as the specimen. the salt. Typical analyses of samples (which have been carefully protected from the pump bowl gas by use of the evacuated bulb samples) show drops a factor of 100 from the top 6 to 10 mils to 2% of the calculated concentration of these iso < the second layer. More recent examinations which topes. used cuts as thin as 1 mil show the materials to Deposition of fission products on the ~ 7- to 10-mil outer layers of graphite specimens ex be concentrated in an extremely thin layer. Typ ically 95% of the metallic species are within posed for 8000 MWh in MSRE is shown in Table VI. 3 mils of the surface. It is clear that, with the assumption of uniform deposition in or on the moderator graphite, ap preciable fractions of Mo, Te, and Ru and a larger fraction of the Nb are associated with the graph ite. Subsequent examinations of graphite after TABLE VTI 235 22 000 and 70 000 MWh have shown somewhat Fission Product Distribution for MSRE at End of U Operation smaller fractions deposited for all these isotopes. Quantity Found (% of Calculated Inventory) The data of Table VII described below show values more typical of the longer exposures. 40 On In On Hastelloy- Purge A careful determination of uranium in or on the Nuclide In Fuel Graphite" N" Gasb Total 2 40 graphite specimens led to values < 1 µg/cm • 99 This quantity of uranium-equivalent to a few Mo 0.17 9.0 19 50 78 132Te 0.47 5.1 9.5 74 89 grams in the moderator stack-seems completely i2"Te 0.40 5.6 11.5 31 48 negligible. 103Ru 0.033 3.5 2.5 49 55 io6Ru Figure 5 shows the change in concentration of 0.10 4.3 3.2 130 138 the fission product isotopes with depth in the 95Nb 0.001 to 2.2 41 12 11 64 140 95Zr 94 0.14 0.06 0.43 95 graphite. Those isotopes (such as Ba) which 89Sr 83 8.5 0.08 17 109 penetrated the graphite as noble gases show 14oBa 96 1.9 0.05 0.48 98 straight lines on the logarithmic plot; they seem 141Ce 0.33 0.03 0.88 to have remained at the point where the noble gas i«ce 0.92 0.03 2.7 131! 60 0.11 0.3 19 80 decayed. As expected, the gradient for 140 Ba with 140 16-sec Xe precursor is much steeper than that •Calculated on assumption that average values for surveillance for 89Sr, which has a 3.2-min 89Kr precursor. All specimens are representative of all graphite and metal surface in MSRE. the others shown show a much steeper concentra b These values are the percentage of production lost assuming mean tion dependence. Generally the concentration values found in gas samples apply to the 4-liter/min purge. 148 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY precursors. A small quantity of 95 Zr appears in the gas phase; this isotope, along with several others with stable soluble fluorides, almost cer tainly appears in the gas as fine particles of salt mist. 39 However, high concentrations of 99Mo, 132Te, 95Nb, 103Ru, and 106Ru are found in the gas phase. Indeed, as Table VII indicates, the gas w 4012 samples represent the largest fractions observed r tt-::i:-i------t----+-----+------+----+---1~}t====t==::::::::::::=====t==::::::::::::=====t==::::::::::::=t==::::::::::::=1=:=:j 95 ~ Nlc--1-1~---i------t----t----t------+--~ for each of these isotopes except Nb. In addition, appreciable fractions of 110Ag and of palladium ~ H1~~~_l~---r--~~t---~--+~~-+-~~+---~ isotopes have been observed in the purge gas samples. An appreciable quantity of 131I appears 11 in the gas phase; it seems very likely that its i 40 CH:] <.!) appearance there is due to "volatilization" of its a:: w tellurium precursor. a. w The mechanisms by which the Mo, Nb, Te, and r :::> Ru isotopes appear in the gas phase are still not 2 ::;: fully understood. NbFs, MoFs, RuFs, and TeF6 are all volatile (although Ag and Pd have no volatile fluorides); these volatile fluorides, however, are far too unstable, with the possible exception of NbF 5, to exist at the redox potentials in MSRE. Stability of lower fluorides of these elements is poorly known; it is possible that some of these may contribute to the "volatility." It seems much more likely, however, that all these species exist in the reduced MSRE fuel in ~ ...... t----+-~~.-+-""'-..:-=--+----1-103Ru-+- elemental form. They originate as (or very rap idly become) individual metal atoms. They ag gregate in the fuel at some slow but finite rate, and become insoluble as very minute colloidal particles which then grow at a slower rate. These 10 20 30 40 50 colloidal particles are not wetted by the fuel, they DISTANCE FROM SURFACE OF GRAPHITE (mils) tend to collect as gas-liquid interfaces, and they can readily be swept into the gas stream of the Fig. 5. Concentration profile of fission products in MSRE core graphite after 8000 MWh. helium purge of the pump bowl. They tend to plate upon the metal surfaces of the system, and (as extraordinarily fine "smoke") to penetrate the Examination of the Hastelloy-N surveillance outer layers of the moderator. While there are specimens exposed in the MSRE core reveals very difficulties with this interpretation it seems more low concentrations of the stable fluoride-forming plausible than others suggested to date. fission products such as 89Sr, 95 Zr, and 141Ce. The Deposition of fission products on or in the more noble metals such as Mo, Nb, and Ru are graphite is, of course, undesirable since there deposited, along with Te, in appreciable though not they serve most efficiently as nuclear poisons. Of in large quantities. If the surveillance specimens the species studied only Nb (whose carbide is are truly representative of all the metal surface stable at MSRE temperatures) seems to prefer the in MSRE then (see Table VII) only a relatively graphite. small fraction of these metals are so deposited. It seems likely that in a MSBR the molten fuel However, flow conditions in the MSRE heat ex will contain virtually all of the zirconium, the changer (where most of the metal surface exists) lanthanides and yttrium produced, and a large is turbulent while laminar flow occurs over the fraction of the iodine, rubidium, cesium, stron core specimens. It seems very likely that the tium, and barium. Thess species would, therefore, actual fraction deposited on MSRE metal is higher be available for removal in the processing plant. than that shown here. The salt will contain very little Mo, Nb, Te, Ru, Analysis of samples of gas from MSRE yielded and (presumably) technetium. These will report some real surprises. Isotopes such as 89Sr and to the MSBR metal, graphite, and purge gas sys 140Ba are found in amounts consistent with con tems in fractions which are probably not very sideration of the half-lives of their Kr and Xe different from those indicated for MSRE. NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 149 Grimes MSBR CHEMISTRY SEPARATIONS CHEMISTRY was adsorbed on the added oxide or participated in the solid solution formation is not known. The Molten-salt breeder reactors will require ef process was shown to be reversible; treatment of fective schemes for recovery of the fissionable the oxide-fluoride system with anhydrous HF dis material which is produced and for removal of solves the oxide and returns the Pa to the fluoride fission product poisons from the fuel. MSRE pro melt from which it can readily be reprecipitated. duced no significant quantity of fissionable isotope It seems likely that Pa might be removed from and made no provision for on-stream removal of the MSBR fuel by effective oontact of a fuel mix fission products other than the noble gases Kr and ture side stream with U02 -Th02 solid solution. Xe. Such a possibility-which is still under considera Recovery of uranium from molten fluorides by tion in laboratory-scale equipment-would add volatilization as UFs and the subsequent decon appreciably to the 233 U inventory through holdup of tamination of this UF6 by sorption-desorption on U02 in the solid solution. It would also probably beds of NaF is well demonstrated. Recovery of require that the oxide ion concentration in the fuel uranium from MSRE or MSBR fuel by this scheme after contact be diminished (as by treatment with is clearly feasible. However, performance of a HF) before the melt could be returned to the single-fluid MSBR is markedly improved if the reactor. bred protactinium is removed from the circulating Separations based upon reduction of the Pa 233 fuel and permitted to decay to U outside the (present in the fuel as PaF4) to metal and extrac reactor core. Fluorination is not effective in re tion into liquid metals appears at present to be moval of protactinium from molten fluoride mix more attractive.48 Such separations have the fol tures. Development of other processes for Pa lowing basis. recovery has, accordingly, been necessary. When thorium metal, dissolved as a dilute alloy In molten-salt reactors, from which Kr and Xe in bismuth (ThB;), is equilibrated with a molten are effectively removed, the most important fis mixture of LiF, BeF2, and ThF4 several equilibria sion product poisons are among the lanthanides. must be satisfied simultaneously. These include These fission products, as indicated above, form stable fluorides which are soluble in the fuel and LiF (d) + iTh(Bi) ;o=" iThF4(d) + Li(Bi) ( 17) will report almost quantitatively with the fuel to the separations plant. Their removal especially BeF2(d)+ ±Th(Bi) ;o="±ThF4 + Be(B;) (18) from fuels with high ThF4 concentrations is diffi 3 I cult. Rare earths have, accordingly, received 4 ThF4(d) + 4Th(Bi) ;o=" ThF3(dl , (19) more attention than other fission products. where (d) and (Bi) indicate that the species are in Processes for recovery of Pa and the lan thanide fission products from MSBR fuels are fluoride solution and alloyed with the Bi, respec described in some detail by Whatley et al.48 The tively. Reaction (19) seems to be of little im chemical basis for such processes is described portance since only a small quantity (if any) ThF3 very briefly below. forms. Reaction (18) does not occur perceptibly; Be is "insoluble" in Bi and forms no intermetallic compounds with that liquid metal. Reaction ( 17) Separation of Protactinium occurs to an appreciable extent even though LiF 19 Protactinium is produced in the MSBR fuel by (see Table IV) is much more stable than ThF4. reaction of a neutron with 232 Th to yield 233 Th Uranium forms stable dilute alloys in Bi, and its fluorides are less stable than ThF . 19 Accord which transmutes to 233 Pa; the 233 Pa decays with a 4 27.4-day half-life to 233 U. Clearly, a separation ingly, the reactions 233 of Pa from the fuel needs, if it is to be effec UF4(d) + iTh(Bi) ;o=" iThF4(dl + UF3(dl (20) tive, to be accomplished on a cycle time short compared with this half-life. It is desirable to and process the fuel on a cycle of three to five days. The process needs, therefore, to be relatively UF3(dl + t Th(Bil ;o=" U(Bi) + t ThF4(d) (21) simple. proceed very far to the right with low concentra Removal of Pa from LiF-BeF2-ThF mixtures 4 tions of Th (and Li) in the Bi at equilibrium. Uran by addition of BeO, Th02, or even U02 has been 49 50 ium is, therefore, very readily extracted from demonstrated in laboratory-scale experiments. • 51 52 Addition of these oxides to MSBR fuel composition MSBR fuel melts. ' If the LiF-BeF2 -ThF4 melt contains PaF4, the would, as indicated in a previous section of this reaction report, result in a uranium-oxide-rich solid solu tion of U02-Th02. Whether the protactinium oxide PaF4(d) + Th(Bi) ;o=" ThF4(d) + Pa(Bi) (22) 150 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY 52 55 proceeds to quite a convenient extent - as shown Separation of Rare Earths in Fig. 6 although protactinium is reduced less easily than is uranium. Accordingly, if a melt There is no real doubt that the rare-earth containing both UF4 and PaF4 is contacted with elements, which form fluorides that are more bismuth and thorium metal added in small incre stable than ThF/9 will be present in MSBR fuel ments the uranium and the protactinium are ex sent to the processing plant. tracted. The limited solubility of these trifluorides Preliminary experiments56 disclosed that Pa (though sufficient to prevent their precipitation could indeed be removed from the salts by reduc under normal MSBR conditions) has suggested a tion with Th but material balance data were very possible recovery scheme. When a LiF-BeF2- poor and little Pa appeared in the Bi. More recent ThF 4- UF 4 melt (in the MSBR concentration range) experiments have used equipment of molybdenum, that is saturated with a single rare-earth fluoride which seems essentially inert to the salt and the (LaF3, for example) is cooled slowly, the precipi alloy, and very careful purification of the ma tate is the pure simple trifluoride. When the melt 53 55 terials and the cover gas. - By these techniques contains more than one rare-earth fluoride the 53 54 two sets of investigators • have shown values in precipitate is a (nearly ideal) solid solution of the excess of 1500 for the separation factor trifluorides.45•46 Accordingly, addition of an ex- cess of CeF3 or LaF3 to the melt followed by Npa NThF4 (23) heating to effect dissolution of the added tri NPaF 4 NTh ' fluoride and cooling to effect crystallization ef fectively removes the fission product rare earths where N indicates mole fraction of the indicated from solution. It is likely that effective removal species either in metal or in LiF-BeF2-ThF4 so of the rare earths and yttrium (along with UF3 and lution. The separation factor for U from Pa has PuF3) could be obtained by passage of the fuel been shown to be above 25. These separation fac through a heated bed of solid CeF3 or LaF3. How tors are, of course, quite adequate for process 48 ever, the price is almost certainly too high; the design, as discussed in detail by Whatley et al. resulting fuel solution is saturated with the sca venger fluoride (LaF3 or CeF3, whose cross sec tion is far from negligible) at the temperature of contact. Similar solid solutions are formed by the rare earth trifluorides with UF3. It might, accordingly, 233 c 90<-----t----+-----1--+---+-----+---+----+-----< be possible to send a side-stream (with the Pa 233 ~ and U removed by methods described above) ...... 12. "'Q_ , .. - ---+---i--+--u-..,,--+--=-0 -+---+---i-... 80 11-4--+----+-___,__ +P ___,_ ___,__ _g_ -i - through a bed of UF3 to remove these fission ::;; I p_n 238 :::> I product poisons; UF3 would presumably be used z I f:= ~- ______, _ a PROTACTINIUM IN BISMUTH 70 for economic reasons. The resulting LiF-BeF2- u p i-- o PROTACTINIUM IN SALT ThF4 solution would be saturated with UF3 after g JI I 6 THORIUM METAL IN BISMUTH its passage through the bed. This 238 U would have ; 60 - 1---1-r- -- C----r------t------1 to be removed (for example, by electrolytic re 1-- duction48 into molten Bi or Pb) before the salt LFLOWING ARGON- 1-- _ FLOWING ~ 50 g-j L STATIC -;RG~~-- FLOWING ARGON HELIUM could be returned to the cycle. The UF3 bed could 1 be recovered by fluorination to separate the ura ~ 40 .--+-- __ I _J _ -r--- J _ _ _ -J __ ~ nium and rare earths. While this process probably o I r I I deserves further study, the instability of UF3 in J I :~ 30..-~I--+--+----+------+--l T l melts with high UF3/UF4 ratios and the ease with w I I I which uranium alloys with most structural metals ::;; I i l would tend to make application of such a process ~ 20 I ~ l I unattractive. g-j \"I " ~ .: 1_" " I ~I ~ IQ J_ I j_ Removal of rare-earth ions (and other ionic I : I : fission-product species) by use of cation ex I I I changers has always seemed an appealing possi : J I I 5 -v 25 45 65 85 105 125 145 l65 bility. The ion exchanger would, of course, need CONTACT TIME {h) (a) to be quite insoluble, (b) to be extremely un reactive (in a gross sense) with the melt, and Fig. 6. Distribution of protactinium and material bal (c) to take up rare-earth cations in exchange for ance for equilibrium of LiF-BeF2-ThF4 (72-16- ions of low-neutron cross section. The bed of 12 mole%) with Bi-Th alloy (0.93 wt% Th). CeF3 described above functions as an ion ex- NUCLEAR APPLICATIONS & TECHNOLOGY VOL. B FEBRUARY 1970 151 Grimes MSBR CHEMISTRY changer; it fails to be truly beneficial because it ff!1 is too soluble in the melt. 5l 103 TIII ! I I c Unfortunately, not many materials are truly r---+---+-+-++++++---j---+-NEODYMIUM>----+-+++++! c stable to the LiF-BeF2-ThF4-UF4 fuel mixture. 0 Zirconium oxide is stable (in its low temperature ~ form) to melts whose zrH/u4+ ratio is in excess 0.... l of ~ 3, and U02 -Th02 solid solutions are stable at - I I equilibrium U/Th ratios. It is conceivable that I v sufficiently dilute solid solutions of Ce2 0 3 in these oxides would be stable and would exchange 0 Ce3+ for other rare earth species. Intermetallic Qi E compounds of rare earths with moderately noble .!:: metals (or rare earths in very dilute alloys with c LANTHANUM"- I I I .Q 101 l . . Nii such metals) seem unlikely to be of use because t; they are unlikely to be stable toward oxidation by e UF4. Compounds with oxygenated anions (such as -OJ rI z silicates and molybdates) are decomposed by the 0 E fluoride melt; they precipitate U02 from the fuel mixture. It is possible that refractory compounds CERIUM j v . (such as carbides or nitrides) of the rare earths, either alone or in solid dilute solution with an alogous uranium compounds, may prove useful. By Reduction The rare-earth f 1 u or ides are very stable toward reduction to the metal. For example, at lOOO"K the reaction • IL 2/3La.F3(1) + Be(c) _, 2/3La(c) + BeF2(l) , (24) 0:: t) {SAMARIUM 0 10-2 lul l l where (c) and (1) indicate crystalline solid and liq 10-2 10-1 100 101 uid, respectively, shows ~ +30 kcal for the free en ergy of reaction.19 With the LaF3 in dilute solution LITHIUM FOUND IN METAL PHASE (at. 3) and BeF2 in concentrated solution in LiF-BeF2 mix ture the situation is, of course, even more un Fig. 7. Effect of lithium concentration in metal phase favorable. However, the rare-earth metals form on the distribution of rare earths between LiF BeF2 (66-34 mole%) and bismuth at 600"C. extremely stable solutions57 in molten metals such as bismuth; the activity coefficient for La at low concentrations in Bi is near 10-14 . Therefore, the + tThF4(d) + Li(Bi)-=" LiF(d) + tThBb(c) (26) reaction ~Bi(l) iLaF3(d) +Be(c)-=" iLa(Bi) + BeF2(d) , (25) proceeds at lower concentrations than these to yield solid ThBb. Accordingly, the most reducing where (d) indicates that the species is dissolved in metallic phase which is tolerable is that repre 2LiF·BeF2, and (c) indicates crystalline solid can sented by the (barely) saturated solution of ThBb be made to proceed essentially to completion. in Bi. When such solutions are used the rare Accordingly, LaF3 can be reduced and extracted earth fluoride is much less completely extracted into molten Bi from LiF-BeF2 mixtures. Figure 7 and relatively small separation factors from tho 51 53 55 shows the behavior of several rare-earth ele rium are obtained. • , Figure 8 shows data ments58 upon extraction into bismuth using Li0 obtained55 in extraction of Ce at 600°C from sev (which is more convenient to use) as the reductant. eral LiF-BeF2 -ThF4 mixtures (the points are Such a process would seem to be useful. labeled to indicate mole percentages of LiF, BeF2, However, when the melt is complicated, as is and ThF4, respectively) in this way. Separation necessary for the single-fluid breeder, by the factors appreciably above unity are obtained even addition of large quantities of ThF4 the situation in the worst case, and compositions containing 0 becomes considerably less favorable. Use of Li 12 mole% ThF4 with < 20 mole% BeF2 appear concentrations as high as those shown in Fig. 7 is somewhat more attractive. Unfortunately, Ce3 + is prohibited since the reaction one of the easiest of the rare earths to reduce. 152 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 Grimes MSBR CHEMISTRY Others in the series show even smaller separation properties of these fluids, estimated in most cases factors, Another paper in this series presents in from data on similar materials, seem adequate formation showing how processes with separation for their use in MSBR. factors of this magnitude could be operated.48 Compatibility of the LiF-BeF2-ThF4-UF4 melt There is little doubt, however, that improved with Hastelloy-N and with moderator graphite (larger) separation factors would be desirable; seems assured, and operation of LiF-BeF2-ZrF4- there is reason to believe that improvement can UF 4 melts in MSRE indicate very strongly that the be obtained by modification of the alloy phase. compatibility will not be adversely affected by the consequences of radiation and fission of the ura nium. Compatibility of the secondary coolant, while less satisfactory, appears to be adequate with Hastelloy-N if salt purity is maintained. Fission product Kr and Xe are virtually insolu ble in the fuel and can be removed, if the moder 5 ator graphite is sufficiently impermeable, by -.c simple equilibration with an inert gas (helium). >- 0 Fission products with stable fluorides (Rb, Cs, Sr, Ba, the lanthanides, and Zr) appear almost ' ~4. D entirely in the fuel as fluorides except as they are II m ~ I lost through volatilization of precursors. More er: N 0 N I noble fission products (Nb, Mo, Ru, and Te) are ~ 0 t; 3 I --··- f'-. virtually absent from the fuel, but plate (in ap ~ preciable amounts) on the graphite and the Hastel z !r N• IN BISMUTH uranium and protactinium from the fuel salt and I I from each other have been demonstrated in small r This research was sponsored by the U.S. Atomic SUMMARY Energy Commission under contract with the Union Phase behavior of LiF-BeF2-ThF4-UF4 mix Carbide Corporation. tures appears suitable to permit use of a high concentration of ThF in melts whose freezing 4 REFERENCES point will prove acceptable for single-fluid molten-salt breeder reactors. Oxide tolerances 1. M. W. ROSENTHAL, P. R. KASTEN, and R. B. of such mixtures are not entirely defined, but BRIGGS, "Molten Salt Reactors-History, Status, and laboratory experience and the MSRE operating Potential," Nucl. Appl. Tech., 8, 107 (1970). experience combine to suggest that no oxide scavenger (such as ZrF4) need be added. The 2. E. S. 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