VALIDATION OF ENDF/B-VII.0 NUCLEAR DATA LIBRARY FOR SHIELDING CALCULATIONS USING THE MONTE CARLO METHOD

The ENDF/B-VII.0 nuclear data library has been widely adopted by researchers in nuclear science and technology applications and especially for the Integral validation testing for a variety of quantities: For nuclear criticality, shielding applications ...; in this paper we have been validated this library through some benchmarks: the PCA-REPLICA and NESDIP experiments for the Light-Water reactor (LWR) pressure vessel facility benchmark. The principal goal of those experiments is the acquisition of benchmark-quality integral neutron data to validate the shielding transport methods and the nuclear data library used in the calculation of the ex-core fluxes in LWR. For this analysis, a 3-D model of the benchmarks experiments chosen for this study, are established by the mean of the continuous energy Monte Carlo code MCNPX. The benchmarks analysis was made based on the calculated-to-measured (C/M) dosimetry reaction ratesof , and monitors, at different depth in a water/iron shields reproducing the ex-core radial geometry of a PWR. The comparisons between reference Monte Carlo calculations and experimental results of PCA-REPLICA and the NESDIP showed us that the calculation method with the ENDF/B-VII.0 nuclear data library is effective for the shielding study of the REP, and generally, the obtained results show a high degree of agreement with the experiment an especially for PCA-REPLICA benchmark. Our choice of MCNPX code is based on the main are used routinely for studying radiation shielding analysis and the fact that feature a rich palette of Variance Reduction (VR) techniques, which play a very important role to reduce the computer required to obtain results of sufficient precision. Some of these VR techniques have been used in the frame of this protection study such as the energy cut-offs (CUT card), the Geometry splitting and Russian roulette (IMP card) and the phase space