Plutonium recovery from spent reactor fuel by uranium displacement

A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

BACKGROUND OF THE INVENTION 
This invention relates to a pyrometallurgical reprocessing of irradiated 
nuclear reactor fuel elements to recover purified uranium and a mixture of 
uranium and plutonium for use as a fresh blanket and core in a nuclear 
reactor. More particularly, this invention relates to a process for 
refining spent blanket and core fuel in a single electrorefining cell by 
dissolving uranium and plutonium from the spent fuel in a molten chloride 
salt and thereafter selectively electrolytically removing first uranium 
and then thereafter using molten cadmium having uranium values dissolved 
therein chemically to displace plutonium values from molten salt and 
replace those values with uranium to reestablish the original salt 
composition. 
The disposal of radioactive waste which results from the reprocessing of 
irradiated nuclear power reactor fuel elements is one of the major 
problems facing the nuclear power industry. One approach is to solidify 
the radioactive waste as it comes from the reprocessing facility into a 
stable solid material which can be stored in the earth for a period of 
time sufficient for the radiation to decay to acceptable levels. However, 
storage times required for spent reactor fuels to achieve such levels of 
radioactivity are on the order of a million years. This is far longer than 
the geologic stability of a waste repository can be expected to be 
maintained. One solution is to remove the extremely long lived radioactive 
components such as the transuranic elements of neptunium, plutonium, 
americium and curium from the waste so that the remaining radioactive 
elements, representing the bulk of the radioactive waste, need only be 
stored for up to about one thousand years for the radioactivity to decay 
to radioactive levels of uranium used in making original fuel. It is 
acceptable to ensure the integrity of a repository for one thousand years. 
The actinides thus recovered from the waste can then be reprocessed and 
recycled to provide additional fuel for nuclear reactors and for isotopic 
power sources. 
As used herein, the phrase "rare earth fission product values" includes 
yttrium and the lanthanide fission product elements while the phrase, 
"transuranic values" or TRU elements include neptunium, plutonium, 
americium and curium values. 
Molten cadmium cathodes in combination with chloride salts have been used 
in processing spent fuel elements from the Integral Fast Reactor (IFR) as 
reported in U.S. Pat. No. 4,880,506 issued Nov. 14, 1989 to Ackerman et 
al. and assigned to the assignee of the present application, the 
disclosure of which is herein incorporated by reference. 
The IFR concept is a complete, self-contained, sodium-cooled, fast reactor 
filled with a metallic alloy of uranium, plutonium and zirconium, and 
equipped with a close-coupled fuel cycle. Close-coupling of the reactor 
and the fuel cycle facility is achieved by locating the reactor and the 
reprocessing, fuel refabrication, and management efficient product waste 
on one site. With this arrangement, it is unnecessary to ship fuel to or 
from the reactor site. Fission products may be processed and stored on 
site for long periods of time, perhaps the entire life of the reactor, 
before shipment to a waste repository where ultimate disposal is required. 
Accordingly, it is clear that reducing the volume of waste product 
produced for each reprocessing is inherent in and required by the IFR 
concept. 
A pyrometallurgical process utilizing electrorefining for purification of 
the core fuel has been developed to reprocess the reactor fuel. In this 
process, the chopped fuel rods are dissolved, or transferred by anodic 
solution to a solid cathode and thereafter molten cadmium is used 
chemically to transfer plutonium from the salt to the cadmium and replace 
the plutonium in the salt with uranium. The apparatus disclosed in the 
'506 and U.S. pat. No. 4,814,046 patents may be used to accomplish part of 
the process of the present invention. 
In general, a low carbon steel container may be used to hold the chopped up 
fuel spent fuel elements or rods and a low carbon steel cathode may be 
used on which to deposit uranium, as will be disclosed. Subsequent to the 
electrotransport of the spent fuel through the electrolyte, a quantity of 
molten cadmium having uranium values dissolved therein is put in contact 
with the electrolyte so that uranium dissolved in the cadmium replaces 
plutonium dissolved in the salt by chemical transport. The overall result 
is that the uranium and plutonium values in the spent fuel are transferred 
electrochemically and chemically to the molten cadmium with the salt 
eventually regaining composition it had before the onset of the 
electrochemical transfer of uranium and plutonium from the anode into the 
salt. Thus, repeated batches of spent fuel elements can be treated without 
substantially altering the salt composition. During the hereinafter 
described process, rare earth values tend to remain in the salt and build 
up in concentration whereas uranium and transuranic values transfer to the 
electrodes. 
It is known in the art how to remove the rare earth values which build up 
in the chloride electrolyte with reducing agents such as lithium cadmium 
alloys or lithium-cadmium-potassium alloys in order to isolate the rare 
earths in a metal matrix, all as previously disclosed in the Johnson et 
al. U.S. Pat. No. 4,814,046. 
SUMMARY OF THE INVENTION 
An improved method or process for treating IFR spent fuel has been 
discovered which involves combining both electrical transport and chemical 
replacement in a single process to permit repeated processing of batches 
of spent fuel elements using the same molten chloride bath. 
It is therefore an object of the invention to provide a combination 
electrochemical and chemical process for the treatment of spent nuclear 
reactor fuel. 
Another object of the invention is to provide a process for treating 
repeated batches of IFR spent fuel with substantially the same salt 
electrolyte. 
Yet another object of the invention is to provide a process for recovering 
uranium and transuranic values from spent nuclear fuel in which the 
processing media including the chloride salt and cadmium solvent may be 
recovered and reused for successive batches of spent fuel. 
Yet another object is to provide a process for separating uranium values 
and transuranic values from fission products containing rare earth values 
when the values are contained together in a molten chloride salt 
electrolyte comprising providing a molten salt electrolyte having a first 
ratio of plutonium chloride to uranium chloride, contacting the molten 
salt electrolyte with both a solid cathode and an anode having values of 
uranium and fission products including plutonium, electrolytically 
transferring uranium and plutonium from the anode to the electrolyte while 
uranium values in the electrolyte electrolytically deposit as uranium 
metal on the solid cathode causing the electrolyte to have a second ratio 
of plutonium chloride to uranium chloride, removing the solid cathode with 
the uranium metal deposited thereon and establishing chemical 
communication between the electrolyte having the second ratio and molten 
cadmium having uranium dissolved therein, and transferring plutonium 
values from the electrolyte to the molten cadmium and transferring uranium 
values from the molten cadmium to the electrolyte. 
Yet another object of the invention is to provide a process for separating 
uranium values and transuranic values from fission products containing 
rare earth values when the values are contained together in a molten 
chloride salt electrolyte comprising providing a molten salt electrolyte 
having a first ratio of plutonium chloride to uranium chloride, contacting 
the molten salt electrolyte with both a solid cathode and an anode having 
values of uranium and fission products including plutonium, 
electrolytically transferring uranium and plutonium from the anode to the 
electrolyte while uranium values in the electrolyte electrolytically 
deposit as uranium metal on the solid cathode in an amount equal to the 
uranium and plutonium transferred from the anode causing the electrolyte 
to have a second ratio of plutonium chloride to uranium chloride, removing 
the solid cathode with the uranium metal deposited thereon and 
substituting therefor molten cadmium having uranium dissolved therein, and 
chemically transferring plutonium values from the electrolyte to the 
molten cadmium and transferring uranium values from the molten cadmium to 
the electrolyte until the first ratio of plutonium chloride to uranium 
chloride is reestablished. 
A final object of the invention is to provide a process for separating 
uranium values and transuranic values from repeated batches of fission 
products containing rare earth values when said values are contained 
together in a molten chloride salt electrolyte comprising providing a 
molten salt electrolyte having a first ratio of plutonium chloride to 
uranium chloride, contacting the molten salt electrolyte with both a solid 
cathode and an anode having values of uranium and fission products 
including plutonium, electrolytically transferring uranium and plutonium 
from the anode to the electrolyte while uranium values in the electrolyte 
electrolytically deposit as uranium metal on the solid cathode in an 
amount equal to the uranium and plutonium transferred from the anode 
causing the electrolyte to have a second ratio of plutonium chloride to 
uranium chloride larger than the first ratio, removing the solid cathode 
with the uranium metal deposited thereon and dissolving some of the 
uranium metal in molten cadmium, contacting the molten cadmium with 
uranium values dissolved therein with the molten salt electrolyte having 
the second ratio, and chemically transferring plutonium values from the 
electrolyte to the molten cadmium and transferring uranium values from the 
molten cadmium to the electrolyte until the first ratio of plutonium 
chloride to uranium chloride is reestablished. 
The invention consists of certain novel features and a combination of parts 
hereinafter fully described, and particularly pointed out in the appended 
claims, it being understood that various changes in the details may be 
made without departing from the spirit, or sacrificing any of the 
advantages of the present invention.

DETAILED DESCRIPTION OF THE INVENTION 
These and other objects of the invention for recovering uranium and 
transuranic values from spent fuel elements and separating those values 
from rare earth fission product values when these values are contained 
together in a fused chloride salt may be met by first electrolytically 
transferring the values from the metal state into the electrolyte salt and 
thereafter chemically transferring uranium values for plutonium values 
with a molten cadmium solution. 
The fused salt useful herein is a mixture of alkali metal or alkaline earth 
metal chlorides, except beryllium and magnesium, that has a low melting 
temperature and in which chlorides of the rare earth fission products and 
the transuranic elements have high solubilities. The salt is a mixture of 
one or more chlorides of lithium, sodium, potassium, calcium, strontium 
and barium that are thermodynamically more stable than the rare earth and 
actinide chlorides. For example, a salt consisting of about 23 weight 
percent lithium chloride, about 35 weight percent barium chloride, about 
32 weight percent calcium chloride and about 10 weight percent sodium 
chloride and a eutectic mixture of 56% potassium chloride and 44 weight 
percent lithium chloride have been found to be satisfactory. However, any 
number of different combinations of chloride salts meeting the above 
criteria will be satisfactory and may be substituted one for the other 
without serious deleterious consequences. 
The amount of molten cadmium used in the chemical transfer of uranium 
values therein for plutonium values present in the electrolyte depends 
upon the amount of uranium which needs to be transferred back to the salt. 
The solubility of uranium in cadmium depends upon the temperature of the 
molten cadmium metal. For example, at 500.degree. C., cadmium is saturated 
with about 2.3 weight percent uranium. Accordingly, depending upon the 
amount of uranium required to be dissolved in the molten cadmium, more or 
less cadmium will be required, it being within the skill of the art to 
determine the total weight of cadmium required to dissolve the necessary 
amount of uranium for the chemical transfer portion of the invention 
process. The temperature at which the entire process operates must be at 
least above the melting temperature of the cadmium and the salt and below 
the temperature at which the components begin to vaporize, except for the 
cadmium distillation step which requires a temperature in excess of the 
boiling point of cadmium at the pressure used. The temperature of the 
electrochemical portion of the process may vary from about 450.degree. C., 
depending upon salt composition, to about 550.degree. C. Generally, a 
temperature of about 500.degree. C. has been found to be satisfactory both 
for the electrotransport portion and the chemical transport portion of the 
subject invention. The cadmium distillation, on the other hand, would 
require temperatures in excess of about 1040.degree.K or about 
767.degree. C. when the distillation is carried out at one atmosphere, but 
when vacuum conditions are employed such as 5 mm Hg, then the temperature 
can be about 400.degree. C. As is understood by those skilled in the art, 
since the boiling point of uranium is about 4,407.degree.K and that of 
plutonium about 3500.degree.K, cadmium is easily distilled from a solution 
of cadmium with plutonium and uranium values dissolved therein. As 
hereinbefore stated, a principal object of the invention is to provide a 
method of harvesting the uranium and transuranic values from spent fuel 
elements while conserving the chemicals used in the process so that the 
amount of waste chemicals produced is small and the amount of new 
chemicals required for processing is likewise small. 
For a feedstock containing 15.6 kilograms of uranium and 4.4 kilograms of 
plutonium, 390 kilograms of lithium-chloride-potassium chloride eutectic 
has been determined to be satisfactory based on smaller scale studies. The 
390 kilograms of eutectic salt contain 21.5 kilograms of plutonium and 
11.6 kilograms of uranium. With an anode of spent fuel and a solid cathode 
of low carbon steel, uranium can be electrolytically deposited on the 
cathode by impressing a voltage across the anode and the cathode. The 
electrochemical cell is run until about 20 kilograms of uranium are 
deposited on the solid electrode at which time the feedstock of uranium 
and plutonium has been depleted and the salt now contains 25.9 kilograms 
of plutonium and 7.2 kilograms of uranium. Thereafter, 300 kilograms of 
cadmium are heated to approximately 500.degree. C. and 6.6 kilograms of 20 
kilograms uranium deposited on the solid cathode are dissolved in the 
molten cadmium. 
The molten cadmium with the 6.6 kilograms of uranium dissolved therein is 
then put in contact with the electrolyte containing the 25.9 kilograms of 
plutonium and 7.2 kilograms of uranium at which time the uranium in the 
molten cadmium displaces plutonium in the salt until the original salt 
composition is reestablished. 2.2 kilograms of uranium remain dissolved in 
the cadmium and 4.4 kilograms of plutonium have transferred from the 
electrolyte to the molten cadmium. At this time, the original salt 
composition of 390 kilograms of lithium chloride, potassium chloride 
eutectic having 21.5 kilograms of plutonium and 11.6 kilograms of uranium 
has been reestablished. Thereafter, the 2.2 kilograms of uranium and the 
4.4 kilograms of plutonium are separated from the molten cadmium by 
elevating the temperature of cadmium above its boiling point and 
distilling the cadmium from the solution leaving the uranium and plutonium 
values. 
The ratio of plutonium chloride to uranium chloride prior to uranium 
displacement controls the ratio of plutonium to uranium in the mixed 
product. Because the former ratio depends only on the amounts of uranium 
chloride and plutonium chloride in the electrolyte before introduction of 
a fuel batch and the amount of uranium and plutonium in the fuel, it is 
not subject to control, provided that uranium in the amount of all the 
actinide values in the fuel is transferred to the solid electrode. The 
amount of plutonium chloride and uranium chloride in the electrolyte 
before introduction of the feed is established before the first batch of 
mixed uranium-plutonium product is processed. For stable process 
operation, the salt composition should return to the original value after 
processing of each feed batch. In order to achieve this steady state 
composition, it is only necessary to remove as much uranium and as much 
plutonium as introduced with the feed batch. Constancy of the uranium 
chloride/plutonium chloride ratio requires that the amount of plutonium or 
uranium removed from any batch be equal to the amount of plutonium or 
uranium introduced from the feed for that batch. This will naturally occur 
if the salt composition is correct to begin with. There is enough uranium 
chloride and plutonium chloride in the electrolyte at the beginning of 
each process that removal of the exact amount is not critical. If the 
average plutonium and uranium concentrations in the feed are well known, 
the process will average out differences of about 10 percent in batch 
compositions without any operator adjustment of the amount of uranium in 
the chemical displacement step. If the electrotransport portion of the 
invention method is carried to completion and the uranium from the 
displacement step is removed from the uranium recovered in the electrode 
transport step, the total amount of uranium and plutonium removed is 
constrained to be exactly the amount put in. The effect of minor errors in 
the salt composition is not great and tends to be averaged out over 
several batches. 
It is not possible with a fuel batch having 20 kilograms of actinides to 
use more than 20 kilograms of uranium in the displacement step without an 
additional uranium supply. On the other hand, it is required to use more 
uranium than the amount of plutonium to be removed. The ratio of plutonium 
to uranium in the mixed product can be varied over a considerable range, 
depending on the amount of uranium used for displacement. Typically, a 2/1 
ratio of plutonium to uranium in the product is preferred and the amount 
of uranium in the displacement step becomes 1.5 times the amount of 
plutonium in the feed. 
For a given amount of uranium in the displacement step, there is one and 
only one ratio of PuCl.sub.3 /UCl.sub.3 in the salt that will result in 
the desired plutonium removal. The required initial ratio (which is the 
same ratio that should be achieved after each batch) can be calculated 
from the equilibrium partition coefficient by assuming the desired amount 
of plutonium removed and the desired PuCl.sub.3 /UCl.sub.3 ratio. The 
partition expression is: 
##EQU1## 
Using the desired Pu/U ratio and amount of plutonium in the product, we 
calculate the required PuCl.sub.3 /UCl.sub.3 ratio in the intermediate 
step: 
##EQU2## 
Noting that the electrotransport step adds all the plutonium in the feed 
to the salt and removes the same amount of uranium from the salt, we can 
work backward to get the initial salt composition: 
______________________________________ 
PuCl.sub.3 + 4.4 = 25.9 
PuCl.sub.3 = 21.5 
UCl.sub.3 - 4.4 = 7.2 
UCl.sub.3 = 11.6 
______________________________________ 
For the same amount of plutonium removed, the required initial salt 
compositions for several product compositions are given in Table 1. Table 
2 shows the effect of an error in initial salt composition, assuming that 
6.6 kilograms of uranium is used for the displacement reaction. All these 
values are calculated as shown below for the first entry in Table 2. We 
take x as the amount of plutonium in the final product. 
##EQU3## 
Here X is the amount of uranium in the product 
TABLE 1 
______________________________________ 
Required Initial Salt Composition for Several 
Product Compositions 
(Constrained to Remove 4.4 kg of Pu Present in Feed) 
Product Salt 
Pu U PuCl.sub.3 UCl.sub.3 
PuCl.sub.3 /UCl.sub.3 
______________________________________ 
4.4 kg 4.4 kg 16.9 kg Pu 16.2 kg U 
1.0 
4.4 kg 2.2 kg 21.5 kg Pu 11.6 kg U 
1.9 
4.4 kg 1.1 kg 24.7 kg Pu 8.4 kg U 
2.9 
______________________________________ 
TABLE 2 
______________________________________ 
Product Obtained With Various Initial Salt 
Compositions 
(6.6 kg U used for displacement) 
Product Salt 
Pu U PuCl.sub.3 UCl.sub.3 
PuCl.sub.3 /UCl.sub.3 
______________________________________ 
4.8 kg 1.8 kg 27.7 kg Pu 5.4 kg U 
5.1 
4.4 kg 2.2 kg 21.5 kg Pu 11.6 kg U 
1.9 
2.7 kg 3.9 kg 16.6 kg Pu 16.4 kg U 
1.0 
______________________________________ 
While there has been disclosed what is considered to be the preferred 
embodiment of the present invention, it is understood that various changes 
in the details may be made without departing from the spirit, or 
sacrificing any of the advantages of the present invention.