Abstract: This paper presents the burnup calculation capability of RMC,which is a new Monte Carlo（MC） neutron transport code developed by Reactor Engineering Analysis Laboratory（REAL） in Tsinghua university of China.Unlike most of existing MC depletion codes which explicitly couple the depletion module,RMC incorporates ORIGEN 2.1 in an implicit way.Different burn step strategies,including the middle-of-step approximation and the predictor-corrector method,are adopted by RMC to assure the accuracy under large burnup step size.RMC employs a spectrum-based method of tallying one-group cross section,which can considerably saves computational time with negligible accuracy loss.According to the validation results of benchmarks and examples,it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.