Analytical method for processing neutron multigroup transfer cross sections

Journal Article · Sat Aug 01 04:00:00 EDT 1992 · Nuclear Science and Engineering; (United States)

OSTI ID:7173092

Ofek, R; Segev, M ^[1]

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1. Ben-Gurion Univ. of the Negev, Dept. of Nuclear Engineering, P.O. Box 653, Beer-Sheva 84120 (IL)

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.

• OSTI ID: 7173092
• Journal Information: Nuclear Science and Engineering; (United States), Journal Name: Nuclear Science and Engineering; (United States) Vol. 111:4; ISSN 0029-5639; ISSN NSENA
• Country of Publication: United States
• Language: English

Related Subjects

663610* -- Neutron Physics-- (1992-)
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ANGULAR DISTRIBUTION
BARYONS
CALCULATION METHODS
CENTER-OF-MASS SYSTEM
COORDINATES
CROSS SECTIONS
DISTRIBUTION
ELASTIC SCATTERING
ELEMENTARY PARTICLES
FERMIONS
FUNCTIONS
HADRONS
INELASTIC SCATTERING
LEGENDRE POLYNOMIALS
NEUTRON FLUX
NEUTRONS
NUCLEONS
POLYNOMIALS
RADIATION FLUX
SCATTERING