Abstract: The inherent safety of reactor is emphasized in the design of the sodium-cooled fast reactors (SFRs). For accident mitigation, inherent safety and passive measures are adopted to reduce the demand for power sources and enhance safety and economy. SFRs can effectively prevent the unprotected accidents. The probability of core meltdown is very small. However, in order to prevent the large radioactive release, SFRs still consider the mitigation measures for large-scale core meltdown. For pool-type sodium-cooled fast reactors, there is a large amount of sodium with high heat capacity in the reactor vessel. It is advantageous to install the core catcher in the reactor vessel for collecting and cooling the core melt. FRTAC is a liquid metal fast reactor safety analysis code developed by China Institute of Atomic Energy (CIAE). The Experimental Breeder Reactor-Ⅱ (EBR-Ⅱ) shutdown heat removal test 45R (SHRT-45R) is an unprotected loss-of-flow event conducted by Argonne National Laboratory (ANL). The primary circuit natural circulation test of Phenix Reactor is a main pump shut down with scram event conducted by French Alternative Energies and Atomic Energy Commission (CEA). The FRTAC code simulated the SHRT-45R test and the natural circulation test performed during the Phenix end-of-life. The results showed that the calculated value is in good agreement with the experiments. This indicated that FRTAC code can be used for thermal-hydraulic simulations under natural circulation of SFR. In this paper, FRTAC code was used to analyze the natural circulation heat transfer after the core meltdown. And the core melt in-vessel retention scheme was studied. The analysis object is a 1 500 MWt pool-type sodium-cooled fast reactor. The reactor has four independent decay heat removal systems (DHRSs). The DHRS consists of three loops, namely a primary loop, an intermediate loop and an air cooler loop. According to the research, the relocation time for the core melt to the core catcher is about 12 hours. At this time, the decay heat is about 9 MW. The natural circulation in the reactor vessel can effectively cool the core melt and transfer the decay heat into the sodium pool. The passive DHRSs can export the heat from the sodium pool to the atmosphere. The maximum temperature of the sodium pool does not exceed 450℃ during the cooling of the core melt. Therefore, the integrity of the core catcher and the reactor vessel will not be affected.