Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the JEFF-3.1.1 library by using integral data assimilation

Abstract

Comparisons of calculated and experimental isotopic compositions of used nuclear fuels can provide valuable information on the quality of nuclear data involved in neutronic calculations. The experimental database used in the present study − containing more than a thousand isotopic ratio measurements for UOX and MOX fuels with burnup ranging from 10 GWd/t up to 85 GWd/t − allowed to investigate 45 isotopic ratios covering a large number of actinides (U, Np, Pu, Am and Cm) and fission products (Nd, Cs, Sm, Eu, Gd, Ru, Ce, Tc, Mo, Ag and Rh). The Integral Data Assimilation procedure implemented in the CONRAD code was used to provide nuclear data trends with realistic uncertainties for Pressurized Water Reactors (PWRs) applications. Results confirm the quality of the 235U, 239Pu and 241Pu neutron capture cross sections available in the JEFF-3.1.1 library; slight increases of +1.2 ± 2.4%, +0.5 ± 2.2% and +1.2 ± 4.2% are respectively suggested, these all being within the limits of the quoted uncertainties. Additional trends on the capture cross sections were also obtained for other actinides (236U, 238Pu, 240Pu, 242Pu, 241Am, 243Am, 245Cm) and fission products (103Rh, 153Eu, 154Eu) as well as for the 238U(n,2n) and 237Np(n,2n) reactions. Meaningful trends for the cumulative fission yields of 144Ce, 133Cs, 137Cs and 106Ru for the 235U(nth,f) and 239Pu(nth,f) reactions are also reported.