Abstract: Based on test facility named COPRA, the natural convection heat transfer characteristics in corium pool inside the reactor pressure vessel lower plenum during severe accident were studied. The test apparatus was a two-dimensional 1/4 circular slice structure with an inner radius of 2.2 m to simulate the lower plenum of reactor pressure vessel at 1∶1 scale for the Chinese independently-designed GEN-Ⅲ PWR. A non-eutectic binary mixture of 20%NaNO₃-80%KNO₃ (in mole fraction) compositions was selected as the simulant material. The Rayleigh number within the pool could reach up to 10¹⁶, matching those in the prototypical situation for PWR. The influences of relocation position, pool height, power density and relocation times on pool temperature and heat flux distribution were studied in the experiments. The result shows that the Nu of the downward heat transfer from COPRA experiment is lower than those from other experiments within the same range of Rayleigh number.