Provided is a method for obtaining multiple groups of Pn scattering matrices in a multi-group library of a reactor. The method comprises following steps: obtaining an expression of continuous energy Pn scattering matrices at temperature TK based on a free gas scattering model through formula derivation and utilization of distribution data of elastic scattering angles at temperature of 0K provided by ENDF, wherein the expression is that sigma sn, T(E to E'); solving a neutron moderation formula based on a continuous energy Pn scattering matrix at the temperature TK to obtain neutron flux density of temperature TK and uniting groups and calculating a multi-group Pn scattering matrix at temperature TK. The method has following beneficial effects: the multi-group Pn scattering matrix after uniting groups and calculating is more accurate so that accuracy of resonance calculation in physical calculation of the reactor is finally increased.