Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

New releases of nuclear data files have been made available during the few recent years. The reference MCNP5 code for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community: ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modularNJOY system (release 99.9) in conjunction with its most recent updates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the International Handbook of Evaluated Criticality Safety Benchmarks Experiments from OECD. Some revised benchmarks were taken from references. These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(a,b) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically ±13 to ±20 pcm. (author)