Analytic Insights into the Neutronic Characteristics of Neutron Moderators from MCNP Calculations

摘要

Using the average number of collisions and the neutron lifetime produced by MCNP on infinite media verification test problems for thermal scattering data, and also using the averaged neutron velocity, very accurate estimates of the flux weighted average total, scattering, and capture cross sections can be made. Such estimates aid the understanding of the neutron characteristics of thermal scattering law materials - at any thermal temperature and for any scattering treatment. When these estimates are not very good, then there is either a problem in the nuclear data or in the MCNP test problem calculation.