Abstract: Based on Monte Carlo simulation method, we apply the multigroup calculation code of the MCNP to simulate the neutron transport equation. A coupled code of the MCNP code and the lattice homogeneous code WIMS is proposed to realize the coupled calculation of criticality and burnup. The specific process is as follows: firstly, the energy group is extended to 69-group owing to the extension of the multigroup calculation character of the MCNP code; secondly, we utilize the interface program to generate 69-group resonance and to transform the self-shield macroscopic neutron cross-section produced by the WIMS code into the multigroup cross-section of ACE format; thirdly, the new cross-section is supplied to the MCNP to finish criticality and burnup calculation; lastly, we use the coupled code to simulate benchmarks and the opposite experiment. The results indicate that the coupled code is correct and rational.