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

In nuclear reactor physics, the integral quantities (reaction rates, k e f f , 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 233 U were selected in order to estimate the impact of the uncertainties related to the reaction of this isotope on k e f f and to validate the ENDF/B-VIII.0 library in comparison to ENDF/B-VII.1 nuclear data. • Study of eigenvalue sensitivity due to the different reactions of 233U isotope using MCNP code. • Analysis of the K eff uncertainty caused by several nuclear data. • Nuclear data uncertainties comparison between ENDF/B-VIII.0 and ENDF/B-VII.1 libraries are performed. • 233U uncertainties of ENDF/B-VIII.0 is significantly reduced compared with that of ENDF/B-VII.1. [ABSTRACT FROM AUTHOR]