Abstract: It outlines MCMGP3-A 3-D multigroup P₃ Monte Carlo neutron transport code, which is designed for computation of reactor critical safety. It is developed from the coupled neutron and photon transport Monte Carlo code MCNP with continual point cross sections and the continual point cross section module of MCNP code has been replaced by multigroup cross section module. MCMGP3 code is with the capablilities of MCNP geometry treatment, counting, deviation reducing techniques and plot. Either macroscopic or microscopic cross sections can be used.The neutron scattering angular distribution adopts P₃ approximation and generalized Gaussian quadrature scheme. The sample problem results prove the computational correctness of the code. MCMGP3 code can treat steady or unsteady porblems with exterior source and critical problems. It suits various cross section formats and the energy groups can be expanded easily from one group to multigroup. In addition, the MCMGP3 code has successfully realized the coupled with the reactor lattic code WIMS. It can be used to simulate the full reactor problems.

Key words: 3-D multigroup, P₃ approximation, multigroup cross section, Benchmarks