Tootfinder

Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP
Javier Ruiz-Pineda, Jaime Romero-Barrientos, Francisco Molina, Marcelo Zambra, Franco L\'opez-Usquiano
https://arxiv.org/abs/2506.22559

Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP
OpenMC is an open-source Monte Carlo code with increasing relevance in criticality safety and reactor physics applications. While its validation has covered a broad range of systems, its performance in spent nuclear fuel storage scenarios remains limited in the literature. This work benchmarks OpenMC against MCNP for eleven configurations based on the KBS-3 disposal concept, involving variations in geometry, fuel composition (fresh vs spent), and environmental conditions (e.g., air, argon, floo…