Three examples of pyrochemical processes that have been developed for purifying uranium or plutonium oxides are listed in Table 10.2.

Skull-reclamation process. The skull-reclamation process was developed by Argonne National Laboratory and used at the Idaho EBR-II Fuel Cycle Facility in the 1960s to recover uranium

from the crucible oxide residues, or skulls, remaining after the partial oxidation, melt-refining process described above. The process involved selective reduction and extraction of the oxides by magnesium-zinc alloys at controlled temperatures and reductant metal concentrations, followed by removal of the magnesium-zinc solvents by distillation.

Salt-transport process. The salt-transport process was studied by Argonne, with the objective of reprocessing short-cooled, high-burnup LMFBR fuel oxide with nonaqueous systems in which radiation damage of solvents would not be a problem. In this process, stainless steel cladding is removed from the fuel by solution in molten zinc at 850°C. The U02-Pu02 fuel is then reduced by a copper-magnesium-calcium alloy with a CaCl2-CaF2 flux at 800°C. This produces a salt solution of the more stable fission-product oxides (Cs2 0, SrO, BaO, and some rare earth oxides), a copper-magnesium solution of plutonium, the rest of the rare earths and the more noble fission-product metals (ruthenium, molybdenum, palladium, etc.), and a solid phase consisting mostly of uranium metal. Plutonium in the liquid copper-magnesium phase is purified by countercurrent extraction with 50 w/o MgCl2, 30 w/o NaCl, 20 w/o KC1, which extracts the rare earths selectively. Finally, the plutonium is separated from the noble fission-product metals by transport through a second 50 w/o MgCl2 salt phase to a 95 w/o Zn-5 w/o Mg Pu-acceptor alloy. This last salt-transport step, suggested by Chiotti and Klepfer [C7], gave the process its name. Argonne [V2] tested individual steps of this process, but did not conduct a complete demonstration with full-bumup fuel.

Salt-cycle process. The salt-cycle process was developed by Battelle Northwest Laboratory [H4] with the following objectives:

To permit reprocessing short-cooled fuel at the reactor site

To handle U02 and U02-Pu02 fuel without requiring conversion to other chemical forms To recover 99 percent of the plutonium and remove at least 80 percent of the neutron­absorbing fission products

To permit control of the plutonium/uranium ratio in recovered fuel

In this process, oxide fuel is dissolved in a molten chloride salt mixture through which C12-HC1 gas is flowing. Dissolved uranium and plutonium are then recovered as oxides by cathodic electrodeposition at 500 to 700°C. The process was demonstrated with kilogram quantities of irradiated fuel, with production of dense, crystalline U02 or U02-Pu02 reactor-grade material. Difficulties were experienced with process control, off-gas handling, electrolyte regeneration, and control of the plutonium/uranium ratio. Development has been discontinued.