Analytical Method for Processing Neutron Multigroup Transfer Cross Sections

Abstract

In this paper an approach is summarized for developing a full analytical method for the generation of laboratory (lab) coordinate system multigroup transfer cross sections of elastic and discrete-level inelastic scatterings of neutrons, where the angular distribution data of the scattered neutrons are given as coefficients of truncated Legendre polynomial expansions in the center-of-mass (c.m.) coordinate system. In the kernel form of the multigroup approximation, fluxes, cross sections, and angular data are left outside the integration signs of the transfer cross-section expression. Then, the integrand is a four-index kernel the source and sink energy groups and the Legendre polynomials in the c.m. and lab systems each contributing one index integrated over the source and sink groups. In the method introduced, the double integration on the neutron pre- and postscattering energies, in these two groups, is carried out analytically.