This study is part of the Sodium-Cooled Fast Reactor projects. Uranium and plutonium dioxide (U,Pu)O2, known as MOX, is the reference fuel. During operation, fuel pellets are subjected to a high thermal gradient that induces mass transport, thermodiffusion, and vaporization phenomena, coupled with irradiation effects. Fuel performance codes are developed to simulate the behavior of fuel rods under nominal and incidental conditions, up to and including meltdown.
The objective of this study is to improve the thermokinetic model of MOX used in these codes. This model is based on the description of the U-Pu-O system using the CALPHAD method, coupled with a database of element mobilities developed using DICTRA software. The description of defects will be extended with the introduction of metal vacancies and oxygen clusters. The description of thermodynamic data (oxygen potential and heat capacity) and the phase diagram will also be improved by taking into account the most recent data. Finally, the mobility database, coupled with the Calphad model, will be improved to better describe diffusion in MOX. New experimental data as well as data calculated using atomic-scale calculation methods (molecular dynamics, ab initio) will be used.