ENDF/B-VII data testing with ICSBEP benchmarks

Abstract
Continuous energy Monte Carlo eigenvalue calculations have been performed for several hundred critical benchmarks described in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. These calculations were performed with MCNP5 using either ENDF/B-VI.8 or ENDF/B-VII.0 cross sections. ENDF/B-VII cross section data files yield significantly more accurate calculated eigenvalues than those obtained with ENDF/B-VI.8 cross sections for moderated, low-enriched uranium fuel rod lattice configurations and for unmoderated, bare or reflected, critical benchmark assemblies. Accurate calculated eigenvalues were previously obtained with ENDF/B-VI.8 cross sections for both highly-enriched and low-enriched uranium solution assemblies. These accurate eigenvalues are retained with ENDF/B-VII.0 cross sections.