Uncertainty Quantification of ²³⁷Np, ²⁴¹Am, and ²⁴³Am 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.

In terms of the fission reaction rate ratios of ²³⁷Np/²³⁵U, ²⁴¹Am/²³⁵U, and ²⁴³Am/²³⁵U, a comparison between experiments and Serpent 2 calculations shows an accuracy of about 5%, 15%, and 10%, respectively, together with ENDF/B-VIII.0 and JENDL-5. For the capture reaction rate ratios of ²³⁷Np/¹⁹⁷Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of the ²³⁷Np/²³⁵U, ²⁴¹Am/²³⁵U, and ²⁴³Am/²³⁵U fission reaction rate ratios by MARBLE with the covariance data of ENDF/B-VIII.0 and JENDL-5 are found to be about 4% at most in all cores, except for about 8% for ²⁴³Am/²³⁵U with ENDF/B-VIII.0 at the intermediate neutron spectrum.