Method of describing the neutron cross sections in the resonant energy range on solving neutron-transfer problems by the Monte Carlo method

Abstract

A method for computing the cross sections of neutron reactions and solving neutron transport theory using the Monte Carlo method is proposed in which cross sections are derived from a gamma distribution factor. This method is compared with and verified against a standard subgroup method in the following way. Neutrons are directed perpendicularly at a homogeneous iron barrier. The flux density of reflected radiation and radiation transmitted through the barrier in the resonant energy region are estimated as recalculated per single source neutron. It is assumed that the results obtained using the subgroup theory for neutron transport should be taken as the standard computational values. Then the relative difference for the given problem between the neutron flux density per source neutron in the subgroup description of the cross sections and the neutron flux density in the case where the method proposed here--using the gamma distribution factor--is no more than 10% on average. Aside from its accuracy, the method proposed in this paper has the advantage of requiring far less computer time than the subgroup method, both in initial calculations and in subsequent estimations of its accuracy.