I
D. 0.CAMPBELL and G. I. CATHERS Oak Ridge National Laboratory, Oak Ridge, Tenn.
Processing of Molten Salt Power Reactor Fuels The basic method appears well suited to a continuous process for this reactor type, but more work is needed to develop chemical flowsheets for specific reactors
H I G H temperature fluid-fuel reactors using molten fluorides have been proposed for producing nuclear power (2, 7). For optimum neutron moderation and economy the core salt would be a lithium-7-beryllium fluoride mixture which would also act as a solvent for fluorides of uranium, plutonium, or thorium. A new chemical process for this type of fuel is based on two principles-i.e., recovery of uranium by volatilization of uranium hexafluoride and recovery of the lithium-7-beryllium fluoride solvent salt for re-use by hydrogen fluoride dissolution. The Aircraft Reactor Experiment demonstrated the feasibility of a high temperature molten salt reactor ( Z ) , and detailed calculations for molten salt power reactors have been published (7). For the work described here the reactor design used as reference has a uranium inventory of 600 to 1000 kg., initially fully enriched, an 8-foot-diameter core surrounded by a thorium blanket, and a total fuel volume of 530 cubic feet (330 cubic feet external to the core). The core salt, 63 mole % ' LiF-37 mole % BeFz plus UF4, weighs about 75,000 pounds, and the blanket salt, 71 mole yo LiF-16 mole % ' BeFt-13 mole % ThF4, about 150,000 pounds. At least 90% of the power is produced in the core, yielding about 180 kg. of fission products per year. After operation for 1 year without processing (except inert gas removal) fission products absorb 3.8% of all neutrons, uranium (-236 and -238) about 3.9y0',. If the fuel is processed to remove fission products and neptunium at the rate of one fuel volume per year, after 10 years the fission products would absorb 2.7y0 of all neutrons, uranium (-234, -236, and -238) 10.4%, and neptunium-237 about 0.9%. Without fuel processing, neutron absorption by fission products would increase almost linearly with time, exceeding 10% after 10 years; neptunium-237 would built up at an accelerating rate, absorb-
ing about 3.7y0 of the neutrons after 10 years; the inventory would increase by about 200 kg. per year; and the conversion ratio would decrease markedly. Even-numbered uranium isotopes, particularly uranium-236, are the worst poisons; but their removal is beyond the scope of chemical reprocessing. Of the 180 kg. of fission products per year, 22 atom % with half lives of more than 78 minutes are subject to removal from the reactor as rare gases; these would contribute 26y0 of the fission product poisoning for 100-e.v. neutrons. About 26 atom 7 0of the long-lived fission products are rare earths, which contribute 40% of the total fission product poisoning. The rest of the fission products consist of a wide variety of elements, no one of which is outstanding from the nuclear poisoning point of view. After reasonably long operation neptunium-237 is the worst individual poison other than the rare earths. The first requirements of a process for this fuel are to recover valuable materials, especially the isotopically enriched lithium and uranium, and to separate them from the major poisons-rare earths and neptunium. Fluoride Volatility Process for MSR Fuel Uranium is recovered from molten fuel salt by direct fluorination to convert UFI to volatile UFO. Similar volatility processes are being developed for uranium recovery from zirconium alloy fuel elements following dissolution in a fused salt (3, 5 ) . The O R N L Volatility Pilot Plant successfully recovered and decontaminated uranium from NaF-ZrF4-UF4 fuel of the Aircraft Reactor Experiment (4). T h e Molten Salt Reactor volatilization process would not require high decontamination of the product UFs because the UF6 could be remotely reduced to UF4 and reconstituted into reactor fuel salt.
Laboratory scale fluorinations of 48 mole yo LiF-52 mole % EkFz eutectic salt containing 0.8 mole yo UF, showed a more complete uranium recovery as the temperature was increased from 450' to 550' C. (Table I). T h e Molten Salt Reactor fuel would contain 0.25 to 1.0 mole % UF4, depending on operating history, Because of its higher melting point, the thorium-containing blanket salt would be processed at a higher temperature. With annual processing the uranium content of the blanket salt is low, increasing from 0.004 mole yo (140 p.p.m.) after 1 year to 0.014 mole yo after 20 years. Fluorination of LiF-BeFz-ThF, salt containing 140 p.p.m. uranium for 90 minutes at 600' C. gave residual uranium concentrations in the salt of only 1 to 2 p.p.m., and over 90% of the uranium was removed in 15 minutes. T h e behavior of other volatile fluorides has been only partially investigated. There was little volatilization of plutonium from fuel salt containing 0.003 mole yoPuFa during 2 hours of fluorination at temperatures up to 750' C. Neptunium was volatilized a t 500' and 600' C. from fuel salt containing 0.002 mole yo NpF,, but less than 10% was collected with the UFs. Trace protactinium-233 was not volatilized from blanket salt by fluorination at 600' C. Fission products which form volatile fluorides are not expected to be troublesome, as they generally have small neutron cross sections, are short-lived, or occur in small yields. Molybdenum occurs in large yield and is a corrosion product, so will probably be present at a concentration in equilibrium with the container. Li'F-BeFs Salt Recovery with HF The Li'F-BeFz salt can be processed by dissolution in anhydrous or nearly anhydrous liquid hydrogen fluoride. T h e relative insolubility of their fluorides in VOL. 52, NO. 1
0
JANUARY 1960
41
"7 ySOLVENT CONDENSER
CORE FLUORINATOR
i 2 5 kg.SALT / DAY
HF- H20
Th,
Fi?
F6
HF-HZO
t 250 kg.SALT/ DAY
SOLUTION
i I
B L A N K E T FLUORINATOR
SALT
i-
90% HF R E F L U X
t
10YoSALT SOLUTION
+
T h , Et? SOLIDS
+
u
F
4
1
w '"FL H2
W A S T E EVAPOR A T 0 R
p
T h RECOVERY WASTE STORAGE
ThF4 Li F Be F2
This is a tentative flowsheet for molten salt reactor fuel processing
such solution makes possible decontamination from the major neutron persons, The LiF-BeF2 salt is recovered from solution by evaporation. BeF2 is very soluble and LiF is insoluble in water ; in liquid H F the reverse is true. The polyvalent element fluorides generally are rather insoluble in both solvents ( 6 ) . Initial measurements indicated that LiF and BeF2 are appreciably soluble in H F containing 10 to 30 weight % water (Figure 1). The LiF solubility decreases rapidly as water is added to anhydrous HF, and the BeFz solubility increases from near zero ; the solubilities are the same in about 80 weight yoHF, 25 to 30 grams per kg. No further measurements were made with LiF or BeFz alone, as the solubilities of the two components together were of primary importance. The solubility of BeFz was measured a t various H F concentrations in the presence of 15 grams of LiF per kg. of solvent (middle line of BeFz solubility, Figure 1) and in solutions saturated with both LiF and BeFz (upper two lines, Figure 1). The increase with increase in LiF concentration confirms that BeFz forms a soluble complex with LiF; this is analogous to the known solubility in H F of the normally insoluble AlFs when NaF is present (7). The effect of BeFz on LiF solubility is similar but much smaller, at least in this range of solvent composition. Thus, two major factors affect solubility:
42
INDUSTRIAL AND ENGINEERING CHEMISTRY
the water content of the H F solvent and complexing between the two salts. Solubility measurements were made also with 63 to 37 mole yo LiF-BeFz salt (48 to 5 2 weight yo),which is approximately the fuel salt composition, in the 80 to 98 weight yo H F range (Table 11). This salt contained about 0.1 mole yo ZrF4, 0.2 mole 76 mixed rare earth fluorides (Lindsay Code 370), and trace fission products. The salt was crushed but not ground to a powder; average particles were flakes about 10 mils thick and 50 to 100 mils across. The solutions were sampled 15 and 60 minutes after salt addition to permit an estimate of the rate of dissolution. The results indicate that an appreciable concentration is reached rapidly, but the solutions are not saturated, especially with respect to
Table I. Higher Temperature Increases Rate of Fluorination Fluorination Time, Hr. 0 0.5 1.0 1.5 2.5 a
Uranium in Salt after Treatment, Wt. % 450" C. 3.39" 1.96 0.39 0.21 0.32
500OC.
550'C.
5.10 4.91 0.20 0.55 0.17 0.20 0.12 0.06 0.11 0.05 5 wt. % added: some uranium probably
precipitated as oxide.
BeF2, in less than several days. The salt behaves like the two components added individually. The maximum total salt solubility, about 160 grams per kg., occurs in 90 weight yo HF. However, the solubility in approximately 1 0 0 ~ oH F may be high enough for process purposes. The results presented here must be considered preliminary. Beryllium fluoride solubilities, in particular, may be generally higher than indicated, because of analytical problems. Difficulty was encountered in determining accurately the water content in highly concentrated HF. This was partly a problem of sampling the volatile solutions and partly the result of determining the small amount of water by the difference of the sample weight and the total weight of LiF, BeF2, and HF. Fission Product Decontamination in HF Dissolution The aqueous H F solutions of salt from the preceding experiment were filtered through sintered nickel, and radiochemical analyses were made to determine the fission product solubilities or decontamination effect. Table I11 shows that rare earths, as represented by cerium, are relatively insoluble in the H F solutions, and are therefore effectively separated from the salt. Solubility decreases as H F concentration is increased. The total rare earth (TRE) analyses do not
REACTOR FUELS Total Solubility of Lithium and Beryllium Fluorides Is Highest near 90% HF Salt in Solution, Mg./G. of Solution T ~ ~ 79.5 ~ .Wt. , 70 H F 89.6 Wt. % H F 95 Wt. % H F 98 Wt. % H F O C. LiF BeFl LiF BeF2 LiF BeFn LiF BeFz
Table 11.
Time 15 min. 1 hi. 20 hr. 4 days 5 hr. 20 hr.
12 12 12 12 - 60 12
28 29 31 32 21 34
33 35 58 68" 65 100
50 51 68 79 74 76
44 50 74 82" 102 92
40 43
17 17
28 22
4 3
63 70 88"
28 36 53
705 71a 60"
22 22 22
Essentially all the component indicated had dissolved and therefore solubility may be higher than the value given. Insufficient salt was added to some solutions to ensure an excess of both components. dissolved. Reactor fuel with a 1-year processing cycle will contain about 0.05 mole yorare earths, so the corresponding decontamination factors would vary from 500 to 16. The results for strontium indicate slight decontamination. Because stronx tium fluoride is fairly soluble in HF, no decontamination was expected. Ce-
show such a large separation because of the presence of the soluble yttrium daughter of strontium, which was then analyzed as a rare earth. The rare earth solubility in these H F solutions saturated with LiF and BeFz increased from about mole % in 98 weight % H F to 0.003 mole in 80 weight HF, based on the amount of LiF and BeF2
LiF A BeF2
I OC
Saturated in BeF2 \
I
0
Saturated in
A
\A
8C
LiF
6C
/\
4c
No L i F
\4
// -\
20
sium fluoride is soluble in both H F and water, and, as expected, there was no cesium decontamination. I n fact, all cesium in the salt added appeared to dissolve. Solubilities of Heavy Elements in HF Neptunium is the most serious nuclear poison other than the rare earths in a molten salt reactor burning uranium-235. With annual processing its concentration in the fuel will slowly increase with time from about 0.002 mole 7 0 after 1 year. Solubility in 80 to 100 weight % ' HF saturated with LiF and BeF2 is sufficiently low to permit removal with the rare earths (Table IV) Small increments of aqueous Np(N03)d solution (-0.02% of the fluoride present) were added to H F solutions saturated with LiF and BeF2 until the solubility limit was reached. Solubilities are reported in milligrams of neptunium per gram of solution and mole per cent of neptunium relative to dissolved LiF and BeF2. Neptunium was determined by alpha counting and pulse analysis to distinguish it from some plutonium impurity. The plutonium appeared to carry with the neptunium to a considerable extent. Addition of iron and nickel metal to the solutions significantly reduced solubility, probably as a result of reduction to Np(II1) (columns 4 and 5, Table IV). U F 4 and ThF4 also decreased the neptunium solubility (columns 6 and 7, Table IV). Trivalent neptunium might be expected to behave somewhat like the trivalent rare earths. Trivalent rare earths, neptunium, plutonium, uranium, and thorium will probably exhibit lower solubilities together than when present individually. A neptunium solurelability of 0.00005 to 0.0002 mole tive to dissolved salt may be expected in actual processing. Solubility measurements indicate that uranium and thorium fluorides are relalively insoluble in 90 to 100% HF. All thorium fluoride determinations showed less than 0.03 mg. of thorium per gram of solution, the limit of detection. Solubilities of UF4 were generally 0.005 to 0.010 mg. per gram, relative to dissolved salt on the order of 0.001 mole %.
.
Solubility of Corrosion Product Fluorides in HF
\A I
0
70
HF
I
I
I
I
\ A '
1
1
,
80
90 10 CONCENTRATION IN SOLVENT (wt,% )
Figure 1. Both lithium fluoride and water increase beryllium fluoride solubility in liquid hydrogen fluoride
Corrosion product (iron, nickel, and chromium) solubilities were measured in H F solutions saturated with LiF. Chromium fluoride was relatively soluble, with values of about 8, 12, and 18 mg. per gram in 100, 95, and 90 weight yo HF, respectively. Iron and nickel fluorides were less soluble. Iron fluoride solubilities varied from 0.08 to 2 mg. per VOL. 52, NO. 1
JANUARY 1960
43
Table 111.
Fission Product Analyses Show Rare Earth Fluorides Insoluble, Cesium and Strontium Fluorides Soluble, in HF Solutions [LiF-BeF2 (63-37 mole %) -0.2 mole % ’ rare earth fluorides trace fission products between 1 and 2 years old] Activity in Original Salt and in HF Solution, Counts/Min-G., of Salt Fission Original 79.5 89.5 95 98 Product’ silt wt.%HF wt.%HF wt.%HF wt.%HF
+
+
Gross@
745 X 104
230 X 104
200 X 104
225 X 104
250 X 104
Gross y
228 174 104 650 510 105
293 251 73 97 7.9 73
2 13 196 67 92 3.6 66
244 223 72 94 1.3 74
282 258 75 101 0.28 81.5
cs Y Sr P TRE Ce P
Y90 @
Zr and Nb precipitated from molten salt before this experiment and were not present in significant concentration. a
gram, and nickel fluoride varied from 0.01 to 1.3 mg. per gram. The results were so inconsistent that no trend with HF concentration could be established in this concentration range, but at lower HF concentrations the solubilities appear much higher. When iron, nickel, and chromium fluorides were present together, the solubilities of all were lower. Recovery of Salt from Solution
LiF and BeFz can be recovered from HF solutions by evaporation. There was the possibility of some hydrolysis of the salt during recovery, resulting mainly from the existence of a high-boiling azeotrope a t 38 weight HF, obtained during evaporation if any water were present, A 90% H F solution saturated with LiF and BeF2 was slowly evaporated to dryness and heated to 450’ C. X-ray diffraction indicated that the resulting salt was about 90% 2LiF.BeF2 and 10% BeF2, and petrographic examination indicated less than 3% of material other than these fluorides. A second sample was evaporated after addition of a little ammonium fluoride (which would decompose and hydrofluorinate the salt at high temperatures) and fused at 700” C. This salt appeared to be entirely the binary compound, and the x-ray pattern was somewhat cleaner than in the first case. Hydrolysis did
Table IV. HF Concn., Wt. % 80 90 95 100 0
44
not appear to be a significant factor in either case. Fuel Processing Flowsheet
A tentative flowsheet for application of the fluoride volatility and HF dissolution processes to molten salt reactor fluids has been prepared. The uranium is volatilized from the salt as LTF8before HF dissolution, but the reverse might be feasible in some circumstances. The lithium-beryllium fluoride salt is dissolved in ca. 90 weight yo HF for separation primarily from the rare rarth and neptunium neutron poisons. The salt is reformed by evaporation and then proceeds to final make-up and purification. This last would probably involve the hydrogen and hydrogen fluoride sparges now required for reactor grade salt (7). The UFE produced in the volatility process would be converted to UF4 for recycle to the reactor; this could be done by conventional hydrogen reduction or possibly by in situ reduction in the fused salt. Although the volatility process achieves high decontamination and the dissolution process largely eliminates rare earths and neptunium, the salt recycle would be extremely radioactive and would require shielding and remote operation. The scale of the process is based on a 1-
Neptunium Solubility Is Below Required 0.002 Mole % Concentration in Fuel Salt Neptunium Solubility UFa-ThFc Bearing Salt” 63-37 Mole % ’ LiF-BeF2 Fe and Ni Metals Added Mg./g. Mole 70 based Mg.!g. Mole % based Mg.!g. Mole % based solution on salt solution on salt solution on salt 0.026 0.011 0.0086 0.0029
0.0031 0.0012 0.00072 0.00024
0.0013 0.00046 0.00054 0.0047
0.00012 0.00005 0.00005 0.00039
0.0043 0.0014 0.0029 0.0029
HF solutions saturated with L~F-BCFS-T~FPUF~ (61.5-37-1.0-0.5mole 96).
INDUSTRIAL AND ENGINEERING CHEMISTRY
0.00052 0.00016 0.00025 0.00025
year cycle. The core salt processing rate would be 125 kg. containing 4 kg. of uranium, per day. Assuming a conservative 10% solubility for the lithiumberyllium fluoride saIt in 90% HF, the H F solution processing rate would be less than 1 liter per minute. The daily blanket processing rate would be about 250 kg., containing only 0.1 kg. of uranium-233. The uranium would be recovered by fluorination, and the blanket salt returned to the reactor. T h e blanket salt need not be processed as frequently to remove fission products, as at most a few per cent of the fissions occur in the blanket and the neutron loss to these fission products is negligible. Although more development of some of the process chemistry is needed: the principal features of the process appear to be suited to the objectives for processing molten salt reactor fuel. Much more work will be needed for developing chemical flowsheets for specific reactor systems. Acknowledgment The authors express appreciation of assistance given by J. T. Roberts and L. G. Alexander on the process requirements for molten salt reactors, and by R . E. Thoma on fused salt systems. Grateful acknoLvledgment is made to members of the Analytical Chemistry Division of Oak Ridge Sational Laboratory for the analyses necessary to the process development work. and to M. R. Bennett for the fused salt fluorination )cork. Literature Cited (1) Audrieth, L. F., Kleinberg, J., “NonAqueous Solvents,” Chap. 10, Wiley, New- York: 1952. (2) Briant, K. C., Weinberg, A . ?vi., others, ATuclear Sci. and En,p. 2, 797-853 (1957). ( 3 ) Cathers, G. I.: Ibid., 2, 768-77 (1957). (4)Cathers, G. I., others, “Second International Conference on Peaceful Trses of
Atomic Energy,” United Nations Paper 535, vol. 17, pp. 473-9, United Nations New York, 1958. (5) Hyman, H. H., Vogel, R. C., Katz, J. J., Ibid., V O ~ .9, pp. 613-26, 1956. (6) Jache, A . W., Cady, G. W., J . Phys. Chem. 5 6 , 1106-9 (1952). (7) Lane, J. A , , MacPherson, H. G., Maslan, F., “Fluid Fuel Reactors,” -4ddison-Whaley Publishing Co., New York, 1958.
RECHVED for review April 6, 1959 ACCEPTED September 24, 1959 Division of Industrial and Engineering Chemistry, Symposium on Chemical Considerations in Circulating Fuel Nuclear Reactors, 135th Meeting, ACS, Boston, Mass., April 1959.
