Validation of ENDF/B-VI.2, JEF-2.2 and JENDL-3.2 through MCNP analysis of BWR-simulated critical experiments

Abstract

The ENDF/B-VI.2, JEF-2.2 and JENDL-3.2 evaluated nuclear data were validated and intercompared for applications in light water reactor analyses. A set of twelve BWR-simulated critical experiments at temperatures representative of cold as well as operating conditions was analyzed by using the MCNP-4A code with continuous energy data. Calculated eigenvalue and fission density distribution results from the MCNP simulations coupled with each of three different nuclear data files showed good agreements with the measurements in the all considered cores.