Chromia-doped UO2 fuel: An engineering model for chromium solubility and fission gas diffusivity

Abstract

Increasing the average grain size of fuel pellets by doping them with chromium oxide is one strategy to improve oxide nuclear fuels performance. The promoted fission gas retention is thought to improve the performance of the fuel at high burnup. In this work, we review models for the solubility of chromium in UO2, and the evolution of the chromium phases in the fuel matrix during irradiation. These models are implemented in SCIANTIX, an open-source mesoscale code describing inert gas behaviour in nuclear fuel. We adjusted the chromium solubility model keeping each parameter within its range of compatibility with experimental data, targeting a better representation of available electron probe microanalysis data of chromium content in fuel after irradiation. As for fission gas behaviour, we considered a physics-based description of the chromium impact on the fission gas diffusivity in fuel grains. The expression for the fission gas diffusivity in standard non-doped uranium oxide has been extended by introducing the impact of the concentration of defects introduced by interstitial oxygen excess representing the effect of chromium content in the fuel itself. A preliminary integral assessment of the proposed models has been carried out against the available experimental data.