Generation of fast neutron removal cross sections using a multi-layered spherical shell model

Abstract

A new method is explored for estimating the fast or fission neutron removal cross sections in single-element or multi-element media. A spherical Monte Carlo model conforming to the removal theory for fission neutrons was used. The MCNP5 code coupled with ENDF/B-VII.0 library was used to simulate the transport of U-235 fission neutrons within the model. Several removal cross sections were generated for abundant elements in construction and nuclear materials including H, Li, Be, B, C, O, F, Al, Cl, Fe, Ni, Cu, W, Pb, Bi, and U, and multi-element materials such as several compounds and composites. Validations were made using experimentally produced fast neutron removal cross section values, and those given by Phy-X/PSD and MRCsC software, and an empirical method. The comparison of the Monte Carlo generated removal cross sections showed good agreements with experimental values and those by theoretical methods. For multi-element media, the Monte Carlo generated removal cross sections were found to best fit the experimental values, in comparison with those produced by other available methods. Furthermore, a comparison of the thermal fluence trends obtained between the Monte Carlo simulations and several neutron removal experiments resulted in good agreement. The proposed method is highly useful in the generation of ENDF/B based removal cross sections, for point-kernel applications and fast neutron shielding materials comparisons.