Neutron Transport Calculations by Using Double-Differential Cross Sections

Some test calculations were carried out to demonstrate the usefulness of double-differential 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 for tritium production rates in a natural lithium sphere. Since the treatment free from collision kinematics is possible by using the double-differential cross sections in the Sⁿcalculations, 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 scattering and the extreme anisotropy of elastic scattering by heavy nuclei. For precise anisotropic transport calculations, it is therefore concluded that the nuclear data of double-differential type are more suitable than those of single-differential type.