Multigroup Cross-Section Generation Using OpenMC Fixed-Source Neutron Transport Simulation for Spent TRISO Fuel

One of the main tasks during repository studies is to demonstrate canister subcriticality under various conditions. If this task is tackled using deterministic codes, a suitable set of cross sections must be generated. In this work, OpenMC fixed-source neutron transport simulation was utilized to generate representative condensed and homogenized multigroup cross sections for spent fuel pebbles in a spent fuel canister. The results show a noticeable difference between reactor and canister cross sections. For isotopes of interest such as 235U and 239Pu, fission cross-section differences of up to 29.41% and 16.31% were seen, respectively, showing that such methodologies are necessary to accurately simulate spent fuel canisters via deterministic methods.