Assessment of standard point-wise neutron data libraries for criticality safety analysis with a Monte Carlo code
- Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland)
This study addresses the assessment of standard continuous-energy neutron data libraries using the Monte Carlo radiation transport code MCNPX for light water reactor criticality safety applications based on a suite of low-enriched, thermal, compound uranium benchmarks and represents a continuation of previously performed analysis using the JEF-2.2 and JENDL-3.3 nuclear data libraries. The new work enhancing the previous study includes the application of the ENDF/B-6.8 neutron data library and employs the most recent official release of the code (MCNPX-2.5.0) with an improved S({alpha}, {beta}) thermal neutron scattering treatment. Particular attention is paid to the analysis of the spectrum-related characteristics of the modeled critical experimental configurations to define the range of applicability of the reported estimates of lower tolerance bounds for k{sub eff}. Inspection of trends in k{sub eff} versus the spectrum-related characteristics or design parameters has also been performed. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22039588
- Resource Relation:
- Conference: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation, Vancouver, BC (Canada), 10-14 Sep 2006; Other Information: Country of input: France; 12 refs.
- Country of Publication:
- United States
- Language:
- English
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