The large radial thermal gradient in Sodium-cooled Fast reactors (SFR) leads to mass transport of uranium and plutonium in the fuel pellet. Knowledge of self-diffusion of all these elements in (U,Pu)O2 fuel is therefore of utmost importance for the prediction of its in-pile behaviour. Very few experimental data, however, are available on self-diffusion in this material.
In the inception of the study, the lack of experimental measurements is circumvented by the so-called ‘cBΩ’ model. The cBΩ model has been utilized to describe the plutonium self-diffusion in MOx using the bulk properties of the fuel and a single fitting-parameter.
Using these results of cBΩ and the experimental data together, a plutonium mobility database has been composed for the MOx fuel. Hence, by implementing this database, a detailed model has been developed to describe the plutonium self-diffusion in MOx using the state-of-art DICTRA code.
The self-diffusion of Uranium in MOx has not been experimentally measured yet. However, by implementing the thermodynamic definition of MOx and the plutonium self-diffusion model, developed under this work, the self-diffusion of Uranium has been estimated from the experimentally measured cation inter-diffusion coefficient data. Hence, a similar Uranium mobility database for MOx has been also composed. Further, using this database, a model for the uranium diffusion in MOx has been developed.
Both the models for self-diffusion of Uranium and Plutonium has been validated against the experimental results. In combination of these two models of cation self-diffusion in MOx, the inter-diffusion profile of cations has been simulated.