A Novel Methodology for Processing of Plutonium-Bearing Waste as Ammonium Plutonium(III)-Oxalate

A novel methodology has been developed for the recovery of Pu from different types of waste solutions generated during various operations involved in the chemical quality control/assurance of nuclear fuels. The method is based on the precipitation of Pu as ammonium plutonium(III)-oxalate and involves the adjustment of acidity of the Pu solution to 1 N, the addition of ascorbic acid (0.05 M) to reduce Pu to Pu(III), followed by the addition of (NH4)2SO4 (0.5 M) and a stoichiometric amount of saturated oxalic acid maintaining a 0.2 M excess of oxalic acid concentration in the supernatant. The precipitate was characterized by X-ray powder diffraction and thermal and chemical analysis and was found to have the composition NH4Pu (C2O4)2.3H2O. This compound can be easily decomposed to PuO2 on heating in air at 823 K. Decontamination factors of U, Fe, and Cr determined showed quantitative removal of these ions during the precipitation of Pu as ammonium plutonium (III)-oxalate. A semiautomatic assembly based on the transfer of solutions by suction arrangement was designed and fabricated for processing large volumes of Pu solution. This assembly reduced the corrosion of the glove-box material and offered the advantage of lower radiation exposure to the working personnel.