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Nuclear fuel safety after discharge: property evolution and testing

The safety of nuclear fuel after discharge from the reactor depends to a large extent on the integrity and/or the behaviour of the spent fuel rod during the cooling, storage and handling / transportation stages. The mechanisms potentially affecting the mechanical integrity of spent fuel rods under normal conditions are associated with the radioactive decay process, which sets the power level in the rods and determines the build-up of alpha-decay damage and helium in the fuel, and with the behaviour of the hydrogen present in the cladding. In addition to the property evolution of intact fuel rods, the behaviour of defective rods and the consequences of potential accidents causing rod failure must be considered. This paper presents an overview of past and ongoing spent fuel characterization studies focused on long-term storage and transportation performed at JRC-ITU. Irradiated fuel alteration as a function of time and specific activity are investigated at the microstructural level (defects configuration and lattice parameter swelling), and at the level of macroscopic properties. Properties measured on tailor-made materials (e.g. alpha-doped UO2) are compared to actual LWR fuel (UO2, MOX). The trend observed for all the measured property changes includes reaching a saturation level after an initial degradation stage. Hardness and thermal conductivity show maximum alteration at a cumulative decay dose of ~0.2-0.4 displacements per atom (dpa), corresponding to storage times of decades for high-burnup UO2. The maximum UO2 lattice swelling of ~0.5% was measured at ~1.2 dpa, corresponding to storage times of the order of centuries. At very high dose, evidence for microstructure reorganization is obtained by transmission electron microscopy analysis. In addition to intact fuel rods, data on the behaviour of defective rods and the consequences of potential accidents are presented. Short segments of commercial ~60 GWd/tU UO2 with pre-set defects were exposed to inert and hydrogen-containing atmospheres at 90°C for several months in a moist autoclave environment. Limited surface alteration and grain boundary etching was observed after 4 months of testing in inert atmosphere; essentially no alteration was observed when hydrogen was present. Similar trends as the fuel corrosion were observed for the fission gas release. Solution analysis revealed significant concentrations of the mobile fission products Cs and I dissolved in water due to condensation films contacting the fuel. These tests are part of a program of studies aimed at characterizing safety relevant aspects potentially affecting LWR fuel rods with different burnup, and simulating different scenarios (ages and thermal histories).