In order to reflect the behavior of the nuclear materials in the operating reactor veritably and roundly, and to supply reliable data for the safety review of the nuclear materials, the most effective solution is building a fuel test loop (FTL) based on the real pressurized water reactor in the research reactor. To ensure the safety of the fuel test loop and the research reactor, it is necessary to analyze the safety of the fuel test loop during accident transient. This article simulates and analyzes two typical kinds of accidents-small break loss-of-coolant accidents (LOCA) and loss-of-flow accidents (LOFA)-based on the origin design of the fuel test loop in China’s Mianyang Research Reactor (CMRR). The peak cladding temperature in cold leg SBLOCA and hot leg small break LOCA is 880.6 ℃ and 367.6 ℃ respectively, which is under the critical value (1204 ℃). During the total LOFA, the MDNBR is greater than 1.5, which means there is no departure from nucleate boiling. In the clamp shaft accident, the cladding temperature is 734.1 ℃，which is less than 1482 ℃. The results obtained are consistent with the safety criteria.