Development and benchmarking of higher energy neutron transport data libraries

Neutron cross-section evaluations covering the energy range from 10/sup /minus/11/ to 100 MeV have been prepared for several materials. The principal method used to generate this data base has employed statistical-preequilibrium nuclear models, sophisticated phase shift analyses, and R-matrix techniques. The library takes advantage of formats developed for Version 6 of the Evaluated Nuclear Data File, ENDF. Methods to efficiently utilize the ENDF/B-VI representation of this library in the MCNP Monte Carlo code have been developed. MCNP results using the new library have been compared with calculated results using codes or data based upon intranuclear cascade models. 7 refs., 8 figs.