Abstract: Molten salt reactor is the only one of liquid fuel reactor among the six candidate reactors chosen by the Generation Ⅳ International Forum (GIF), and the molten salt circled in the primary loop is nuclear fuel as well as coolant. This promises reactor concept features peculiar characteristics such as no manufacture of fuel assembles, online fuel reprocessing and refueling, and therefore the burnup calculation of molten salt reactor differs from conventional solid fuel nuclear reactors. MCORE is a reactor physics analysis code based the MCNP and ORIGEN used for conventional solid fuel nuclear reactors. A new burnup analysis code (MCORE-MS) which considered the online reprocessing and refueling was developed based on MCORE. The preliminary study indicates that under ²³³U-started mode, molten salt on-line processing can reduce fission products in the reactor core effectively, resulting in better neutron economy. And during MSFR operation, the temperature reactivity coefficient is always negative.