Uncertainty Quantification of ²³⁷ Np, ²⁴¹ Am, and ²⁴³ Am Reaction Rates in Highly Enriched Uranium Fuel Cores at Kyoto University Critical Assembly

Experimental analyses of 237Np, 241Am, and 243Am fission, as well as 237Np capture reaction rates, are conducted with the Serpent 2 code together with ENDF/B-VIII.0 and JENDL-5 using experimental data for the neutron spectra of thermal and intermediate regions obtained in the solid-moderated and solid-reflected cores with highly enriched uranium fuel at the Kyoto University Critical Assembly. Also, uncertainty quantification of the fission and capture reaction rate ratios of the test samples of 237Np, 241Am, and 243Am with reference samples of 235U and 197Au are evaluated by the MARBLE code system.