Grain-Scale Modeling of Active Fission Product Behavior During Power Ramps with the Fission Gas Model MARGARET-PAF

Due to their influence on the general behavior of the fuel rod, fission gases in oxide fuels are a major topic of research in the nuclear community. Fuel swelling, thermal conductance degradation of the fuel-cladding gap, and over-pressurization of the fuel rod belong to the potential negative impacts of fission gases during irradiation. The fission gas model MARGARET has been developed by the CEA for many years, and takes into account relevant phenomena related to inert fission gas (xenon and krypton) behavior in nuclear fuels, at the grain scale. MARGARET has been included in fuel performance simulations, and in particular in ALCYONE, the code co-developed by the CEA, EDF, and FRAMATOME within the PLEIADES computational environment. The MARGARET model is mainly used to evaluate the release of inert fission gases during normal or incidental loading sequences. In recent years, the calculated noble gases percolation flux has been employed to assess the release of the minor and chemically reactive fission gases (mostly iodides) from the fuel pellet during slow power transients. This approach was however limited to long- lived fission products, and considered that they were immediately available on grain boundaries. In order to overpass these two limitations, a sub-model of MARGARET, called MARGARET-PAF, have been coupled to thermochemistry and used. MARGARET-PAF calculates the production, disappearance, and transport of a whole series of isotopes (mainly xenon, krypton, iodine, cesium, tellurium, and strontium) in the fuel. MARGARET-PAF uses all the microstructural quantities calculated by MARGARET, which may influence active fission product transport, such as intra- and inter-granular bubble concentrations (in the grain and at grain boundaries) and material restructuring (High Burnup Structure formation). In this study, the model undergoes validation through the simulation of a power transient, replicating an experiment conducted on a refabricated irradiated fuel rod within the OSIRIS experimental reactor. The simulation is conducted using the code stand-alone and input data (temperature, pressure) previously computed by the fuel performance code ALCYONE. Simulation results are discussed and compared to available experimental data.