Dissolution of irradiated MOX fuel with high Pu content

Irradiation experiments were conducted with the aim to assess the irradiation behaviour of high Pu content mixed oxide fuels, intended for operation in fast reactors. The MOx fuel was fabricated by hybrid sol-gel method, all pellets have annular form, with Pu content Pu/(Pu+U) = 0.45. The fuel was irradiated for 74.1 effective full power days. The samples made available for the dissolution study were slices about 2 mm thick, weight of fuel about 0.37 g each. Due to its high plutonium content, the fuel was expected to dissolve slowly with high residues poorly in standard nitric acid (PUREX process) dissolution steps. Therefore, a series of dissolution experiments with varying conditions were proposed: (i) Dissolution in hot concentrated nitric acid, (ii) Dissolution in hot concentrated nitric acid with fluoride as a catalyst and (iii) Oxidation with in-situ chemically generated Ag(II). All methods have been applied and the samples have been dissolved. During the dissolution experiments sub-samples have been taken and analysed for U and Pu content using TIMS method to get information on dissolution rate