Abstract: When solving neutron transport equation by deterministic method, it is necessary to estimate multi-group cross sections with accurate neutron flux spectrum. This flux spectrum can be obtained by MC neutron transport process. The existing program MCNPNG (Monte Carlo Multi-group Cross-section Statistical Analysis Software) for calculating multi-group cross-sections and scattering matrices by using the Monte Carlo method is very time-consuming when calculating models with various materials and complex geometric structures, which is not conducive to engineering application. In this paper, the reasons are analyzed firstly, and the optimization is carried out, which effectively improves the calculation efficiency; Then, the parallel method of multi-group cross section and scattering matrix calculation is studied, and the corresponding program module is developed, which significantly reduces the running time and can still maintain more than 85% parallel efficiency when hundreds of CPU are called. Finally, through the calculation and analysis of OKTAVIAN benchmark problem, the correctness and efficiency of the parallel calculation method of multi-group cross section and scattering matrix are verified.