As soon as sufficient plutonium was available for pilot-plant studies of more efficient processes, research was initiated on solvent extraction processes, which were already in large-scale use for recovery of uranium from ore leach liquors. Compared with the bismuth phosphate process, solvent extraction had the advantages that it could be operated continuously rather than batchwise, and that it could recover both uranium and plutonium in good yield and with high decontamination factors. The first solvent extraction process used in the United States for large-scale separation of uranium and plutonium from irradiated fuel was the Redox process. This process was developed by Argonne National Laboratory, tested in a pilot plant at Oak Ridge National Laboratory in 1948 to 1949, and installed by the General Electric Company at the Hanford plant of the U. S. Atomic Energy Commission (AEC) in 1951 [LI].

The solvent used in the Redox process was hexone, methyl isobutyl ketone, an extractant already in use for purifying uranium ore concentrates (Chap. 5).. Hexone is immiscible with water and will extract uranyl nitrate and plutonyl nitrate selectively from fission-product nitrates if the aqueous solution has a sufficiently high nitrate ion concentration. In the Redox process, aluminum nitrate was used as salting agent because high concentrations of nitric acid would decompose the hexone solvent.

Figure 10.1 is a material flow sheet for the first cycle of one form of the Redox process [F3]. Plutonium in the feed was converted to hexavalent plutonyl nitrate PuVI02(N03)2, by oxidation with dichromate ion Cr2072”, as this is the plutonium valence with highest distribution coefficient into hexone. In the decontamination contactor, hexavalent uranium and plutonium nitrates were extracted into hexone solvent, and fission-product nitrates were removed from the solvent by a scrub solution containing aluminum nitrate, sodium nitrate, and sodium dichromate.

In the partition contactor, plutonium was converted to inextractable, trivalent Punl(N03)3 by a reductant solution of ferrous sulfamate containing aluminum nitrate to keep uranium in the hexone phase. Plutonium was thus separated from uranium and transferred back to the aqueous phase along with the aluminum nitrate. Impure plutonium nitrate was purified by additional cycles of solvent extraction, not shown.

In the uranium stripping contactor, uranyl nitrate was transferred back to the aqueous phase by 0.1 M nitric acid strip solution. Fission products were separated from the impure uranyl nitrate by additional cycles of solvent extraction and by adsorption on silica gel, not shown.

A modification of the Redox process, the 235 U-hexone process, was used at the Idaho Chemical Processing Plant of the U. S. AEC, to recover highly enriched uranium from M5U-A1 alloy fuel elements irradiated in the Materials Testing Reactor. The aluminum nitrate needed as salting agent was provided when the fuel was dissolved in nitric acid. The plutonium content of the fuel was too low to warrant recovery. Plutonium was made trivalent and inextractable before solvent extraction and thus routed to the aqueous high-level waste.

Disadvantages of the Redox process were the volatility and flammability of the hexone solvent and the large amount of nonvolatile reagents such as A1(N03)3 added to the radioactive wastes.