Nuclear data sensitivity and uncertainty analysis using a set of benchmarks containing 233U comparing ENDF/B-VII.1 and ENDF/B- VIII.0 libraries

In nuclear reactor physics, the integral quantities (reaction rates, k(eff), etc.) are estimated by calculation codes (Monte Carlo and deterministic codes). The determination of the bias and uncertainty that may exist between the calculated results and the reality is necessary. These uncertainties are mainly due to the approximations used during modeling, the approximation associated with the calculation method, as well as the uncertainties of the nuclear data. It is necessary to perform the Sensitivity/Uncertainty Analysis using a set of benchmarks in order to provide an overview of nuclear data uncertainty propagation and to identify isotopes and reactions which can contribute to the uncertainty of integral parameters. In this study, a set of benchmarks contained U-233 were selected in order to estimate the impact of the uncertainties related to the reaction of this isotope on k(eff) and to validate the ENDF/B-VIII.0 library in comparison to ENDF/B-VII.1 nuclear data.

Automated summary

In nuclear reactor physics, there are biases and uncertainties in the estimation of integral quantities using calculation codes. Conducting sensitivity/uncertainty analysis helps understand the propagation of uncertainties and identify isotopes and reactions that contribute to the uncertainties in integral parameters.