Methodology and preliminary verification of generating heterogeneous multigroup microscopic cross-section libraries for neutron transport codes based on OpenMC

Advancements in reactor technology, particularly Generation IV and modular reactors, have introduced new challenges on the neutronics analysis due to their complex geometries and spectra. This study addresses these challenges by developing a methodology to generate heterogeneous multigroup microscopic cross-section libraries for three-dimensional neutron transport calculations using the OpenMC Monte Carlo code. The approach involves two-dimensional transport calculations in OpenMC for various fuel pins or supercells, generating multigroup microscopic cross-section libraries for isotopes relevant to burnup, temperature, and moderator density. These cross-sections are then post-processed and used in three-dimensional core neutron transport calculations with the CRANE deterministic code. This method combines the high accuracy of Monte Carlo methods with the computational efficiency of deterministic approaches. Preliminary 2D verification was conducted using benchmark problems, including PWR fuel assemblies from the VERA series, a fast reactor pin, a 3600 MWth subassembly, and a 1000 MWth metallic fuel core. Results indicate that the coupled OpenMC/CRANE method accurately captures reactivity and isotopic evolution during burnup, suggesting potential improvements in accuracy and efficiency for neutronic simulations of advanced reactor designs.

Mots clés

• Monte Carlo
• cross section generation
• multigroup microscopic cross-sections