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, flooding scenarios). Effective multiplication factors ( k eff ) and leakage fractions were evaluated for both codes. Across all cases, OpenMC systematically yielded slightly lower k eff values than MCNP, with absolute code-to-code differences bounded by 136 ( 10 ) pcm. In the normal-operation configurations, differences were − 32 ( 3 ) pcm (spent) and − 38 ( 3 ) pcm (fresh), with negligible leakage. Under periodic boundary conditions, both codes showed the expected boundary-induced rise in k eff ; for both OpenMC and MCNP, the increase relative to vacuum was + 237 ( 3 ) pcm. By providing reproducible input decks and quantifying agreement under equivalent modeling assumptions, this study represents the first systematic OpenMC–MCNP benchmark for KBS-3–type spent fuel canisters. These findings support extending the validation domain of OpenMC toward spent fuel transport and geological disposal applications.