Abstract: Based on the continuous energy cross section Monte Carlo code MCNP a 3 D multigroup P 3 neutron transport Monte Carlo code MCMG is developed. The MCMG code is realized the coupled calculation of criticality and burnup with the lattice homogeneous code WIMS. It uses the 69 group resonance and self shield macroscopic neutron cross section which are produced by the WIMS code and BUGLE 80u47 group microscopic neutron cross section libraries to simulate the simple reactors and the critical test reactor. The calculation results are in good agreement with the results of other transport methods and experiments. The computational time of the MCMG code is less than that of MCNP code with the same precision.