Abstract: The 3-D multigroup P₃ neutron transport Monte-Carlo code MCMG was developed by adding functions and renewed edition. It can equip various microscopic or macroscopic cross-sections. The cross-section input was simplified. The edition was developed from Ⅰ to Ⅱ. The multigroup cross-section was produced by ENDF/B-Ⅶ library. The collision mechanism about the material was used and the developed MCMG-Ⅱ code has the parallel computational function. Twelve critical benchmarks and an external source problem were calculated, and the almost same results with experiments and point-wise cross-section MCNP-5 code were obtained. The speedup of MCMG-Ⅱ is a factor of 3-6 relative to the MCNP-5 in speed.