Comparisons of ENDF/B-V and ENDF/B-VI results for a variety of thermal-reactor lattices
- Los Alamos National Lab., NM (United States)
Calculations have been performed with the MCNP Monte Carlo code for 13 different thermal-reactor lattices at {open_quotes}cold{close_quotes} conditions, using continuous-energy libraries derived from ENDF/B-V and from Revision 2 of ENDF/B-VI. These calculations represent the first extensive comparisons of ENDF/B-V and ENDF/B-VI results for reactor lattices. The two sets of libraries produce similar results for cases with low-enriched uranium fuel. ENDF/B-VI cross sections tend to produce slightly higher values of k{sub {infinity}} for cases with highly enriched uranium fuel and slightly lower values for cases with natural, slightly enriched, or depleted uranium fuel. The differences for these cases result primarily from competition between a more positive reactivity contribution from {sup 235}U and a more negative contribution from {sup 238}U at thermal energies. The ENDF/B-VI cross sections for plutonium isotopes produce essentially the same reactivity effect for these lattices as do the corresponding ENDF/B-V cross sections.
- OSTI ID:
- 459219
- Report Number(s):
- CONF-950420--
- Country of Publication:
- United States
- Language:
- English
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