Speaker
Description
A comparative assessment between OpenMC 0.15.3 and MCNP 6.2 was performed on a three-section concrete labyrinth experiment from the ICSBEP database. The already existing MCNP benchmark model was converted to OpenMC and mesh-based weight windows were generated using the MAGIC method.
Neutron transport from a 252Cf spontaneous fission source was simulated, with neutron flux computed at multiple positions along the labyrinth. Results obtained with each transport code were compared and benchmarked against the available experimental measurements. The ENDF/B-VIII.0 nuclear data library was employed in both cases.
Multiple experimentally tested labyrinth configurations were simulated, including cases with added material to the labyrinth walls and polyethylene source filtering. Given the relevance of polyethylene in the benchmark, a particular focus was placed on thermal neutron scattering, both on the generation pipelines of ACE and HDF5 S(α,β) files and on the treatment of this type of scattering by the two different transport codes.
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