Abstract: Based on the multigroup calculation feature of the MCNP code, the multigroup function is extended. The coupled code is used to calculate the burnup with the lattice homogenous code WIMS. The 69-group resonance and self-shield macroscopic neutron cross-section, is generated by the WIMS code to simulate the fuel cell and the fuel assembly and the comparative experiment is carried out. The calculation results are in good agreement with the results of other methods and experiments. It indicates that the coupled code is correct and rational. Finally, the paper uses the coupled code to compute and analyze the burnup of the Xi'an pulsed reactor.