Neutron Transport Calculations by Using Double-Differential Cross Sections

Abstract

Some test calculations were carried out to demonstrate the usefulness of double-differen-tial cross sections for neutron transport calculations including anisotropic scattering. A transport code system NITRAN was applied for the purpose. In NITRAN, the anisotropy of elastic and inelastic scattering can be treated in a general form by double-differential total neutron-emission cross sections, which are generated from single-differential and/or original double-differential cross section data base.
The test calculations were performed for neutron flux spectra in aluminum and lead slabs, and also f, or tritium production rates in a natural lithium sphere. Since the treat-ment free from collision kinematics is possible by using the double-differential cross sections in the S_n calculations, the discretization of secondary neutron energy distribution becomes independent of the segmentation of angular distribution. A significant improvement due to this independence can be seen in calculating the anisotropy of general inelastic scatter-ing and the extreme anisotropy of elastic scattering by heavy nuclei. For precise aniso-tropic transport calculations, it is therefore concluded that the nuclear data of double-dif-ferential type are more suitable than those of single-differential type.