12th International Symposium on Symbiotic Nuclear Power ( ISSNP-2026)

Sodium-cooled fast reactors are important candidates for improving nuclear-fuel utilization and supporting the transmutation of minor actinides. Reliable simulation of mixed-oxide (MOX) fuel pins remains challenging because heat transfer, fission-gas behavior, fuel restructuring, species redistribution, and cladding deformation are strongly coupled. This work presents a multi-case validation and parameter sensitivity assessment of the open-source multidimensional fuel-performance code OFFBEAT for MOX fuel pins irradiated under sodium-cooled fast-reactor conditions. The validation matrix includes the SUPERFACT-1 SF4/16 and Santenay-1 fuel pins irradiated in the Phénix reactor and the B5206 power-to-melt experiment conducted in the Joyo reactor. The americium-bearing SF4/16 pin is selected as the principal case for evaluating long-term irradiation behavior. Santenay-1 extends the validation to annular MOX fuel under higher-power and higher-dose conditions, while B5206 provides a complementary assessment of fuel thermal behavior during a short beginning-of-life high-power irradiation. The 1.5-D simulations account for fuel and cladding heat conduction, gap heat transfer, mechanical deformation, thermal and irradiation creep, cladding swelling, fission-gas behavior through SCIANTIX, porosity migration, and actinide redistribution. Calculated results are compared quantitatively with available post-irradiation measurements, including fission-gas release or plenum gas inventory, fuel-column elongation, cladding diametral deformation, residual fuel–cladding gap, central-hole size, columnar-grain-region diameter, and radial plutonium distribution. For the B5206 experiment, the predicted thermal response and fuel melting extent are compared with the reported post-irradiation observations. Preliminary calculations for SF4/16 predict an end-of-life fission-gas release of approximately 59%, compared with the measured value of approximately 68%. Sensitivity analyses are performed for selected parameters associated with fuel thermal conductivity, initial gap size, fuel and cladding deformation, porosity migration, and fission-gas behavior. The cross-case comparison aims to identify the dominant sources of calculation–experiment discrepancies and to evaluate the transferability of a consistent OFFBEAT modeling approach without case-specific recalibration. The results provide a basis for assessing and improving fuel-performance models for advanced sodium-cooled fast-reactor fuels.