ENGINEERING, DESIGN, AND PROCESS DEVELOPMENT'
Electrorefining for Removing Fission Products from Uranium Fuels LEONARD W. NIEDRACH AND ARTHUR C. GLAMM Knolls Atomic Power laboratoryl, General Electric Co., Schenecfady, N. Y.
C
ONSIDERABLE work aimed a t the development of compact processes for the recycling of fuel from nuclear reactors is currently in progress. A group of these processes known as high temperature or pyrochemical methods is characterized by the employment of high temperatures and the retainment of the fuel material in the metallic states, at most they involve only a momentary deviation from the metallic state. While none of the pyrochemical processes achieve decontamination from fission products comparable with that obtainable with aqueous processes they do show potential of being less expensive than their aqueous counterparts. Examples of processes included in this group are extraction with molten metals (IS, I8),extraction with molten salts (2, IS), oxidative slagging (7, 18))volatilization (IS), and electrorefining (9: 14). All of these processing methods accomplish decontamination from active metal fission products, but few succeed in removing the more noble fission products from the fuel. Of the processes mentioned, only in the case of the electrorefining process is there theoretical or experimental evidence to indicate that general decontamination from the noble metals can be obtained. While the noble metals are of little concern from a nuclear standpoint, they may eventually affect the metallurgical properties of recycled metal. For this reason, the inherent ability of an electrorefining process to remove both active and noble fission products seemed important, and investigation of the feasibility of such a method was started. The possibility of applying electrorefining to fuel processing is discussed in this article. For this purpose a process employing a fused-salt electrolyte was chosen because the chemical activity of uranium precludes the use of aqueous electrolytes, and danger from radiation damage makes the use of organic solvents unattractive. With an electrorefining process fission products more noble than uranium should accumulate as an anode sludge, while the active metals should concentrate in the salt bath. By the proper choice of operating conditions it may be possible to concentrate any plutonium in the salt phase or to cause it to follow uranium into the cathode deposit. For some reactors, great decontamination from the fission products is not essential. A reduction in concentration of approximately a factor of 10 is satisfactory in many cases. This fact permits the consideration of electrorefining and other high temperature processes because some fission products possess chemical and physical properties close enough to those of uranium to preclude separations of great magnitude. Background literature on electrorefining process for uranium i s limited
In the field of fused-salt electrolysis, important work was performed by Driggs and Lilliendahl ( 6 ) , who used a salt bath con1 Operated for the U. 9. Atomic Energy Commisaion by the General Electric Co.
June 1956
taining sodium chloride, calcium chloride, and potassium uranium pentafluoride for the electrowinning of uranium. The electrolysis operation was performed a t 775" C. and produced powdered or dendritic deposits. Rosen ( 1 6 ) described a similar procedure in which powdered uranium is obtained by the electrolysis of uranium tetrafluoride in a mixture of alkafi and alkaline earth halides a t 800' to 850" C. Recently a discussion was given of a quantitative study of the effects of current density and temperature on the deposition characteristics of uranium obtained from mixed fluoride-chloride baths (11). The temperature range covered was 725' to 900' C. and finely divided metal was again obtained as the product. Workers a t the Argonne National Laboratory used a n electrorefining process t o obtain uranium of exceptional purity when fairly pure uranium was used as the starting material (12). Work was also done at Argonne on electrorefining as a processing method for spent reactor fuels (9). I n both cases the electrolyses were performed at temperatures below the melting point of uranium, and dendritic deposits were obtained. I n all of this work, the metal produced a t the cathode required thorough washing t o remove adhering salts before the metal was compacted. Because of the nature of the products, a batch type of operation was dictated to remove the bulky deposits from the electrolysis cells. I n spite of the awkwardness of this type of procedure, no process for uranium purification has been described in the literature in which a readily handled, molten product is formed. Preliminary considerations include need for molten products and effect of chemical activity of fission products
For a successful electrorefining process for irradiated uranium, it appeared advantageous to obtain a molten product that could readily be removed from the cell by a tapping operation. Dendritic deposits with nonirradiated uranium can be washed free of salt, but such an operation would be difficult t o perform with irradiated fuels and would result in the accumulation of undesirable aqueous wastes and appreciable recycling of salt. Although a cell that produced molten uranium would be ideal for a process, and is the ultimate goal, the high melting point of uranium, 1130' C., puts severe restrictions upon such a system. Therefore, attention has so far been given to the possibility of operating a t lower temperatures with alloys of uranium with the transition elements. An example is the system with manganese, for which the phase diagram (I 9) indicates a eutectic melting a t 716" C. that contains only 6.0% by weight of manganese. Also important, manganese can be distilled from such an alloy so that the alloy need be used only as a transient intermediate in a process. To provide for continuous formation of the alloy in a process, the cathode is constructed of the desired agent. Thus as the uranium deposits, interdiffusion occurs and a low melting alloy
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ENGINEERING, DESIGN, AND PROCESS DEVELOPMENT forms and drips from the electrode to collect in a pool in the bottom of the cell. 4110y formation in this manner, while never employed with uranium, has been observed with titanium ( 8 ) , zirconium (16),and other materials. I n considering an electrorefining process for uranium, an important difference in the behavior of the active metals and the noble metals becomes evident. The active metals presumably go into solution in the salt bath, and as a result their chemical activity continually builds up with each successive throughput. This chemical activity is roughly proportional to the concentration in the bath. As a result, on the tenth throughput the decontamination should be approximately one tenth that obtained on the first throughput. -4n activity increase should not occur for the noble metals because they collect as a sludge a t the anode. This sludge is presumably the elemental material] or at least an undissolved compound whose make-up remains the same whether it is the first throughput, the tenth, or the hundredth. Therefore, no concentration effect is involved, and the chemical activity will remain constant through the entire duration of a run. As a result] decontamination from these noble metals will remain a t its constant and its expected high value throughout the process. This is a very significant fact because one of the first items of concern mith any process is whether the decontamination factor can be maintained. As the diluent salt for an electrorefining bath, the halides of metals less readily reducible than uranium are applicable. The alkali and alkaline earth chlorides and fluorides would seem most desirable for the purpose. Since present interest is in operating a cell at the lowest temperature compatible n i t h a molten product, the chloride systems are preferred to the higher melting fluorides. Thermal data indicate that in chloride systems the trichloride of uranium is the most stable state (S). If necessary] however, the tetravalent salt is charged to the bath, and rapid reduction to the trivalent state occurs. Description of the process
A flow sheet in which a manganese alloy is formed in the electrolysis step is shown in Figure 1. Uranium fuel elements] if clad, would first go to a stripping station, preferably of a mechanical type, and the cladding would be removed and sent t o waste. The stripped elements would then be charged to the basket-type anode of the electrolysis cell. Charging would be through an air lock, since it would be necessary to operate the cell itself in an inert atmosphere. During electrolysis the uranium and active fission products would be anodically dissolved; the more noble fission products would accumulate in the bottom of the basket as an anodic sludge. T h e uranium would then migrate through the salt bath and be deposited upon the manganese cathode, while the active fission products accumulate in the bath. Alloying of the uranium and manganese would occur, and molten alloy would drip from the cathode t o the bottom of the crucible. The product alloy could be removed from the cell periodically by tapping, or continuously by the use of a satisfactory arrangement of weirs. The alloys, which in the case under consideration mould contain 5 to 10% by weight of manganese, mould then be transferred to a volatilization crucible and heated to about 1300"C. to remove the manganese (6). The molten uranium then runs to a casting crucible and from there to satisfactory molds in preparation for recycle. The manganese would probably be discarded. Periodically the basket could be removed and discarded to dispose of the accumulated fission products. The salt bath also would have t o be discarded periodically because of the build-up of the active fission products which had concentrated in it. The uranium could be recovered from the discarded bath by adding an oxidizing agent such as chlorine t o produce the volatile uranium
978
tetrachloride (boiling point 787" C.) ( 3 ) . This would be distilled from the salt, condensed, and recycled to the electrolysis cell. If plutonium had been concentrated in the salt, it too is distilled from the melt with the aid of the oxidizing agent. Because of the instability of the oxidation states higher than plutonium(III), only a small concentration of the more volatile plutonium(1V) chloride exists at any time (1, i ) ,and it will revert t o the trichloride and chlorine upon condensation ( 1 ) . The rate of distillation of plutonium from the salt should therefore be much lower than that of uranium. Hence the separation of the two should be feasible with the uranium being volatilized from the salt before the plutonium. If separation of plutonium from the uranium were desired it might be better to employ an extraction step prior to the electrorefining. Extraction with salt ( 2 , I S ) or metal (18)-e.g., Bilvermight be performed on the molten fuel. In either case active metal fission products would be removed as well as the plutonium, and a considerable burden would be removed from the electrolysis step. Then the latter would be required mainly for the removal of the more noble fission products from the uranium. If the salt extraction procedure were used, the plutonium and any uranium extracted into it could be separated by the oxidation and volatilization procedure shown on the flowsheet. If metal extraction were employed] a distillation step would be used to recover the plutonium. Process discussion considers problems of radiation and advantages and disadvantages of electrorefining
To date work has been concentrated on demonstrating the electrolysis step of the process as well as the possibility of achieving decontamination. Before scale-up vi11 be possible many additional features will have to be studied. For example, a t the present stage of development it appears necessary to have a fuel that can be mechanically dejacketed. Elimination of this restriction would be desirable. Further work is also required on the auxiliary step for recovering uranium and plutonium from the fused salt. Although a volatilization process is indicated on the flowsheet, an electrolysis step in which chlorine (or carbon tetrachloride) is formed a t the anode while uranium and plutonium deposit on the cathode may be more suitable. Such a process for depleting the uranium and plutonium content of the bath appears feasible in view of earlier processes for depositing uranium electrolytically from fused halide media (6, if, 15). Corrosion of container materials may prove to be another serious problem and testing of a variety of container materials or liners will be required. The intense radioactivity of the materials to be processed, presents additional problems both from the point of view of the possibility of radiation induced reactions in the bath and the problem of disposing of the heat generated by absoiption of the radiation. The magnitude of these effects is illustrated by the processing of a fuel having undergone 1% burn-up in 150 days and having cooled for 20 days. When processed a t a rate of about 30 kg. per day through a cell containing 30 kg. of salt, if an average salt life of 25 days (25 throughputs) is assumed, it is found that at equilibrium heat is generated in the cell a t a rate of 39 kw. This will maintain the temperature in the electrolysis cell a t about 950" C. While this temperature is close to the desired operating temperature it would appear that adequate control should be possible through proper cell design, control of the number of throughputs, and adjustment of cooling time. Under the corresponding conditions the specific activity of the radiation in the bath is about 3.5 watts per cubic centimeter of salt. Fused fluorides and chlorides have been found to withstand reactor and cyclotron irradiations of 1000 watts per cc. and electron bombardment a t 120 watts per cc. without decomposition(l7). Therefore, it would not appear that gross decomposition of the salt bath should occur under the conditions of an electrolysis.
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ENGINEERING, DESIGN, AND PROCESS DEVELOPMENT MECHANICAL DECLADDINS
c-o-~-u
1
CLAD FUEL CLADDING
io WASTE
j
EXTRACTION CRUCIBLE
I
1 '
r- SALT TO
!ua PU
R
I
--C
E
GRAPHITE ,BASKET ANODE
~
Tr M kf
easier with the electrorefining process than with most other high temperature processes. On the negative side is the need for remote refabrication of fuel elements from the partially decontaminated fuel. This is a disadvantage of all high temperature processes relative t o aqueous processes. The fact that fission products must be allowed t o accumulate in the salt bath is a disadvantage of electrorefining relative t o other pyro processes. I n addition the time of contact of hot uranium with the cell structural material will be magnitudes greater in the present case. As a result, requirements of structural materials will be high in order t o avoid contamination of the metal with extraneous impurities.
MN TO WASTE
STILL
U
CASTING CRUCIBLE
U
TO u a PU RECQV AND DISCARD
1 "2
LRESIWAL SALT AN& RESIDUE TO WASTE
PERIODICALLY TO WASTE
U TO CANNING AN0 RECYCLE TO REACTOR
Economic evaluation requires consideration of costs for chemicals, equipment, and processing steps
Figure 1. High temperature electrorefining process for fuel recovery
There is, however, a possibility that the presence of excited and other nonequilibrium species in the bath may cause a reduction in the current efficiency of the electrolysis. The intense radiation field associated with the fuel also imposes the restriction of remote operation of the electrolysis equipment. I n view of the experiences existing today in performing chemical operations remotely, major difficulties are not anticipated which are specific t o electrorefining. Again, the product metal from the process will not be in the highly decontaminated form t h a t is obtained from aqueous processes. This feature is typical of all of the high temperature processing schemes and means t h a t a simple fuel capable of remote refabrication will be required. Recycle of casting residues and other metal scrap should be relatively simple in this process. Most of this scrap consists of metal coated with oxide and other refractory films. As such, it can be recycled to the anode basket of the electrolysis cell. Some evidence has been obtained which indicates that oxides of uranium (and presumably plutonium) will settle out in the anode sludge. Whether or not a recovery step will be required for them is not certain at this time but it would appear likely that losses of this type would be small. The path of neptunium in the process isanother t h a t must receive attention if short cooling is t o be used. Presumably its behavior will fall between that of uranium and plutonium and no new problems are introduced. The manganese volatilization step also requires further attention. Preliminary results indicate t h a t temperatures of 1300' C. are required to remove the manganese. However, the use of an alloying agent such as manganese can be eliminated from the flowsheet. This should be possible by going t o higher temperature operation with the formation of molten uranium itself as the direct product of electrolysis. Such a step, of course, would enhance corrosion problems but would seem to be an appropriate step to take. This is actually a direction of present attack on the problem and success has been achieved in initial attempts to drip molten uranium from both carbon and tungsten cathodes. As with other compact high temperature processes, a real possibility is t h a t processing costs can be reduced below those of aqueous processes. This is related' t o the fact t h a t compact waste streams are obtained, and by eliminating aqueous solutions great reductions in plant bulk are possible. I n addition to this potential general advantage, the electrorefining process has an advantage over most other high temperature processes in that i t accomplishes decontamination from both noble and active fission products. Recycling of skulls and residues also promises t o be June 1956
*
At present the process is not developed t o a sufficient degree that a worth-while economic analysis can be made. It is clear, however, that the fundamental concept is not radical and embodies many well developed principles t h a t are economical for the processing of even inexpensive metals. The cost for electricity for the refining of 1 kg. of uranium assuming trivalent uranium in the bath, 50% current efficiency, a cell potential of 2 volts, and power a t 10 mils per kw.-hr. is 1.4 cents. This does not include the cost of power for heating the bath, but, in view of the heat generated by the fission products themselves, this additional cost should not prove excessive.
#
t
0
I
I = 2.5
amp.;
I
66 wt.
I
1
I
% BaClz-34
wt.
% UF4
General equipment costs are parallel to those for other high temperature processes. The size of a n electrolysis cell for processing at a rate of 30 kg. per day is relatively small, roughly a 1-foot cube. This would fit into a relatively small shielded isolation area and could economically be discarded a t frequent intervals if made of graphite. The use of more exotic materials such as carbides or nitrides would require longer life but may result in purer products. Perhaps the most important and expensive of the auxiliary steps is t h a t of recovery of uranium and especially plutonium from the
INDUSTRIAL AND ENGINEERING CHEMISTRY
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-
ENGINEERING, DESIGN, AND PROCESS DEVELOPMENT salt before it is discarded. This operation will, however, be performed on a relatively small side stream and should not contribute excessively t o the cost of the metal in the main process stream. Recycling of skulls and residues from crucibles should not be an expensive operation because they can be sent directly back to the anode basket of the main electrolysis cell. The cost of make-up salt, calcium chloride and uranium trichloride, in the present process m i l l not be excessive if high throughputs can be realized. ilctually most of the uranium salt would be recycled from the discarded salt and little make-up would be required.
100
I
I
I
I
I
I
ih I
I
I
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i -
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00
CURRENT-AMR
Figure 3.
0 66
wt.
0 6 2 wt.
Effect o f current on efficiency
70BaClz-34 wt. 70UFd at 975-1075' 70CaClp38 wt. % UC13 at 850-950'
C. C.
Some expense for manganese will also enter the costs n-ith the piesent process. If not recycled, assuming an alloy of 57G by Ji-eight of manganese and manganese selling a t 82.00 per pound, these costs are 20 cents per Bg. of uranium. It is probable that this cost can be reduced by a factor of ten if recycling were employed. However, the entire cost of manganese as well as the cost of removing it can be eIiminated by going to higher temperature operation with the formation of molten uranium itself as the direct product of electrolysis. Supporting work includes study of variables of electrolysis step using uranium-nickel and uranium-manganese systems
\i7ith the exception of the effect of radiation, the important variables associated with the electrolysis step of the process have been investigated in the laboratoi y and have been discussed (114). I n the laboratory studies the formation of the molten alloy with a manganese cathode was demonstrated, but most of the detailed experimental studies on the electrolysis steps were performed with the uranium-nickel system. This system has a eutectic melting at 740" C. and containing lO.57, by weight of nickel ( I O ) . Because of the similarity of the two systems, the results obtained with the uranium-nickel system should be applicable t o the uranium-manganese sj-stem. The negligible effect of temperature on cell efficiency is shown by the data in Figure 2 . However, the range of operating temperature possible with a uranium alloy system is limited by salt volatilization at high temperatures and by dendrite formation a t low temperatures. An operating temperature as high as 1100" C. is possible with chloride-containing baths and dendrites Twre observed a t 850' C. The marked effect of cell current on efficiency is shown in Figure 3. Cell current, rather than the more pertinent current density, is used as the abscissa because the continual change in electrode size during a run causes variations in current density. It is estimated, however, that the initial current densities varied from 0.1
%O
to 1 amp. per sq. cm. as the electrolysis current was changed from 0.4 to 5 amperes. The effect of current transcends that of temperature and of bath composition. The scatter among the data is caused in large part by the ever changing current densities a t the consumable electrodes employed. Although the effect of current density, especially for the individual electrodes, has not been examined, it is still possible on the basis of the chemistry of uranium t o arrive a t a mechanism that explains the decrease in efficiency with increasing current. Thermal data indicate that trivalent uranium is the stable state in the bath a t the operating temperatures involved ( 3 ) . A t low currents, then, uranium should enter the bath as U+3, and diffusion away from the electrode should occur rapidly enough to prevent polarization to the extent that U f 4 is foimed. Since there are no stable oxidation states between I;+3and Uo, the only possible reaction a t the cathode is reduction to the metal, and since the nickel alloying agent is present, difficulties from fog formation are minimized. High cell efficiencies are therefore in order a t low current densities. As the current is increased, a point is eventually reached when diffusion of the oxidized species away from the anode is not rapid enough to prevent polarization of the electrode to the extent that U+* is formed in significant amounts. When such a state is reached, current could be carried in part by a cyclic oxidation and reduction involving U+3 and U+4, and the current efficiency would decrease. The data in Figure 3 also indicate that the composition of the salt bath has only a minor effect on the cell efficiency. In addition, no effect on the general operability of the cell was observed with the baths indicated. Baths composed of barium fluoride with 38 to 66yc by weight of uranium trichloride also performed satisfactorily but offered no advantages over the barium chlorideuranium tetrafluoride mixture which replaces the hygroscopic uranium trichloride with the readily handled uranium tetrafluoride. When baths-'containing sodium fluoride with 80 to 91 % by weight of uranium tetrafluoride were used, a marked distillation of sodium metal was observed except when the highest concentration of uranium tetrafluoride was employed. I n this case, however, production of uranium trifluoride in the bath caused difficulties with the formation of a solid phase a t the operating temperature, 1038" C. The use of such baths was considered unfeasible. The effect of uranium concentration in the salt bath on product contamination by the bath diluent is shown by the data in Table I. On the basis of these data the bath should contain a definite minimum amount of the uranium salt if the deposition of significant amounts of the diluent cation is to be avoided.
Table I.
Effect of Uranium Concentration in Bath on Product Alloy Composition
(CaClrUCla b a t h : S i cathode: T = 950° C.: 1-3 a m p . ) Xi in Ca in UC18 in Salt Bath, Alloy, Alloy, TVt. 70 n-t. % mt. % 2.6 10 2.7 7.7 25 1.3 14.4 38 0.1-0.3 13.8 76 0.06
Both the current and the operating temperature have significant effects on the nickel content of the product alloy. This is shown by the data in Figure 4 The points correspond to actual alloy compositions obtained a t the temperatures indicated. Because of the variations in current density and probably other more subtle factors, the scatter in the data is too great to distinguish the effect of minor changes in current. For this reason the data have been grouped: alloys obtained a t 0.4 and 1 amp. are in one group; those obtained a t 2 , 2.5, 3, and 5 amp. are in a second
INDUSTRIAL AND ENGINEERING CHEMISTRY
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ENGINEERING, DESIGN, AND PROCESS DEVELOPMENT group; and those obtained a t 10 amp. are in the third. In this way it is evident that a t low currents the alloy composition tends toward that at equilibrium on the nickel-rich side of the eutectic. At higher current densities the composition moves toward the
* o o z
‘
‘look
preciable decontamination was achieved on both the ruthenium and the molybdenum. A decontamination from the chemically active fission products also would be obtained by simple salt extraction. The two runs reported (Table 11)were made for a duration of only two bath throughputs, and the need for runs of greater throughput is evident. These data indicate the magnitude of decontamination that is possible with the method. The information obtained to date on plutonium decontamination is inconclusive. However, there is some indication that the plutonium tends to concentrate in the salt phase when the chloride bath is used. Summary and conclusions
t \
8
?i 700
L
/
0
20
IO
L 30
WEIGHT PER CENT N I C K E L
Figure 4. Relationship between product composition and U-Ni phase diagram Liquidur (70); elecfrolyris current, amp.: 0 0.4 and 1.0 0 2,2.5,3, and 5 0 10
uranium-rich side of the eutectic. This is t o be expected since a greater rate of deposition of uranium a t the cathode should result in less chance for the nickel to diffuse into the deposit before the drop grows heavy enough to fall. These results indicate that cathode current densities higher than those that have been employed would be desirable in order to increase the uranium content of the alloy. However, care must be taken to avoid excessive cathode current densities; otherwise the rate of uranium deposition will so exceed the rate of diffusion of nickel into the deposit that a solid deposit will result and dendrites will grow.
Table
II. Decontamination of Factors Obtained by
Decontamination
Electrorefining counts/min./g. U a t start counts/min./g. U in product Hot Slloy Run Run ~
The general feasibility of an electrorefining process for uranium has been demonstrated both with regard to operability and ability to give decontamination. The effect of a number of cell operating variables has also been demonstrated and stability of operation has been shown. The encouraging results obtained t o date make it desirable to continue with more exhaustive testing of major variables including the effects of radiation and continued operation for extended periods. Larger scale tests are also in order to obtain pertinent engineering data. Concurrent with such tests, methods for remote fuel refabrication must also receive attention, because direct handling of fuel processed by electrorefining is not possible. literature cited
Abraham. B. If.,Brody, B. B., others, National Nuclear Energy Series IV-l4B, p. 753, McGraw-Hill, New York, 1949. Bareis, D. W., Wiswall, R. H., Jr., Winsche, W. E., ,Vzdeonics 12, No. 7 , 16 (1954).
Brewer, L., Bromley, L. A., others, MDDC-1543, 9 (1945). Brewer, L., Bromley, L., Gilles, P. W., Lofgren, N. L., Xational Kuclear Energy Series IV-l4B, p. 873, McGraw-Hill, New York, 1949.
Dearing, B. E., Glamm, A. C., Niedrach, L. W., Knolls Atomic Power Lab. Rept. KAPL-1140, 201 (1954). Driggs, F. H., Lilliendahl, W. C , IND.ENG.CHEM.22, 516 (1930); (to Westinghouse Lamp Co.), U. S.Patent 1,861,625 (June 7 , 1932). Feder, H. M., “Pyrometallurgical Processing of Reactor Fuel,” International Conference on Peaceful Uses of Atomic Energy, Geneva, Paper UN-544, July 1, 1955. Fischer, H., Dorsch, K. (to Siemans and Halske), German Patent 615,951 (July 16, 1935). Glassner, A., Chellew, N. R., Vogel. R. C., Argonne i’iatl. Lab. Rept. ANL-4872, 147, 1952; ANL-4922, 154, 1952. Grogan, J. D., Pleasance, R. J., J . I n s t . Metals 82, 141 (1953). Kantan, 9. K., Shreenivasan, N., Tendolkar, G. S., Chem. Eng. Progr. Symposium Ser No. 12, 50, 63 (1954).
Marzano, C., Noland, R. A., Argonne Natl. Lab. Rept. ANL5102, 1953.
Two runs have also been performed to determine the magnitude of the decontamination from fission products that can be achieved with this method of processing. In one of these runs the anode charge consisted of a mixture of unirradiated uranium with about one tenth its weight of irradiated metal that had been cooled for 2 to 3 years. This gave enough activity that the decontamination through the cell could be followed on the basis of the radioactivity. The second run was made using a uranium alloy that was prepared t o contain 1%by weight of each of ruthenium, molybdenum, zirconium, and lanthanum. Actually the lanthanum was volatilized or otherwise lost during the preparation of the material. The data obtained are summarized in Table 11. An ap-
lune 1956
Motta, E. E., “High Temperature Fuel Processing blethods,” International Conference on Peaceful Uses of Atomic Energy, Geneva, Paper UN-542, June 23, 1955. Niedrach, L. W., Glamm, A. C., J. Electrochem. SOC.,in press. Rosen, R. (to United States of America), U. S.Patent 2,519,792 (Aug. 22, 1950). Steinberg, M. A., Sibert, M. E., Wainer, E., in “Zirconium and Zirconium Alloys,” 58-60, Am. SOC.for Metals, Cleveland, 1953.
U. S. Atomic Energy Commission, “The Reactor Handbook,” vol. 2, p. 838, AECD-3646, May 1955. Voigt, A. F., “The Purification of Uranium Reactor Fuel by Liquid Metal Extraction,” International Conference on Peaceful Uses of Atomic Energy, Geneva, Paper UN-545, July 1, 1955. Wilhelm, H. A., Carlson, 0. N., Trans. Am. SOC.Metals 42, 1311 (1950). RECEIVED for review September 6, 1955. ACCEPTED March 5, 1956. Nuclear Engineering and Science Congress, Cleveland, Ohio, December 1955.
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