Investigation of high temperature irradiated fuel-liquefied Zircaloy interactions in support of severe accident safety studies

摘要

The problem of irradiated fuel (both UO_2 & Mixed Oxide Fuels) interactions with liquefied Zircaloy at high temperatures is central to the understanding of bundle degradation mechanisms during reactor power transients or severe accidents. These initial interactions of the cladding and the irradiated fuel result in a melt (corium) and then to a loss of bundle geometry and the corium accumulation in a pool. ITU investigated the interaction of irradiated fuel and compared it with non-irradiated fuel with its Zircaloy cladding at 2000 °C for various short times. This was its contribution to the COLOSS (Core Loss of Geometry) project carried out under an EC framework programme. The tests were investigated by optical microscopy with image analysis and then by SEM-EDS analysis. The dissolution of the irradiated fuel by the Zircaloy melt was very variable and heterogeneous, but for non-irradiated fuel was reasonably uniform and constant. The kinetics of the non-irradiated UO_2-liquefied Zircaloy interactions was shown in another work package of the project to follow diffusion-limited mechanisms that could be modelled. The large variation in the results with the irradiated fuel rods made it difficult to model these interactions, nevertheless, they appear to have similar parabolic kinetics seen in non-irradiated fuel. The cracked condition of the fuel and the fission gas release during these interactions are major factors for fuel break-up, dispersion and dissolution in the melt under temperature transients.