Calculations of Fast-Reactor Data-Testing Assemblies Utilizing Cross Sections from the MC² and SDX Codes

Reactor physics calculations were performed with ENDF/B Version-Ill cross-section data for several of the fast-reactor data-testing assemblies specified by the Cross Section Evaluation Working Group. In these calculations, multigroup cross sections were generated using both the Argonne MC² and SDX codes for comparative purposes. The multigroup cross sections were then used in S_n transport-theory calculations to obtain k_eff and central activation ratios, and in perturbation-theory calculations to obtain central worths. Results with MC² and SDX cross sections are in good agreement except when regions containing large amounts of a structure material are involved.