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

Current usage metrics show cumulative count of Article Views (full-text article views including HTML views, PDF and ePub downloads, according to the available data) and Abstracts Views on Vision4Press platform.

Data correspond to usage on the plateform after 2015. The current usage metrics is available 48-96 hours after online publication and is updated daily on week days.

Initial download of the metrics may take a while.