Open Access
ND 2007
Article Number 214
Number of page(s) 5
Section Benchmarking for fission applications
Published online 17 June 2008
International Conference on Nuclear Data for Science and Technology 2007
DOI: 10.1051/ndata:07285

ENDF/B-VII data testing with ICSBEP benchmarks

A.C. Kahler and R.E. MacFarlane

Los Alamos National Laboratory, Los Alamos, NM, USA

Published online: 21 May 2008

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.

© CEA 2008