Multicollision source method for discrete ordinates neutron transport calculation

Abstract

Discretization errors like ray effects and numerical diffusion are major shortcomings of SN calculation in shielding problems involving strong anisotropy. Especially in isolated point source shielding problems with cavities and ducts, the angular flux distribution in spatial-angular space may become unsmooth and discontinuous because of heterogeneity effects in the first few iterations. We have developed the multicollision sources (MCS) algorithm via spatial decomposition. The method semi-analytically calculates the scattering source for particles that have suffered several collisions in a selected region, and the SN transport sweeping calculates the scattering for particles in the remaining regions. The calculation parameters of the MCS method is selected according to the distribution of flux in different regions and the scattering effect of shielding materials. For practical complex multi-group problems, the MCS method can be carried out independently in several groups considering the specific problems. We implemented the MCS method in the 3D, parallel, multigroup discrete ordinates code ARES, and numerical results and evaluations are given for a self-designed problem and several general benchmarks. All these results can verify the ability of the MCS method to reduce discretization errors and demonstrate that the MCS method can improve calculation accuracy obviously with an acceptable additional computation cost.