The potential of the RELAP5/MOD3.2 code was assessed for a loss-of-residual-heat-removal (RHR) event during midloop operation, and the predictability of major thermal-hydraulic phenomena was evaluated for the long-term transient. The calculations were compared for two cases of experiments conducted at the Rig of Safety Assessment-IV (ROSA-IV)/Large-Scale Test Facility (LSTF) in Japan: the cold-leg-opening and the pressurizer-manway-opening cases. In addition, the real plant responses to the event were evaluated for Yong Gwang nuclear power plant Units 3 and 4 (YGN 3/4) in Korea, especially concerning the mitigation capability to remove the decay heat through the steam generators (SGs). From the LSTF simulation, it was found that the RELAP5 code was capable of simulating the plant behavior following the loss-of-RHR event under a shutdown condition. As a result, the thermal-hydraulic transport process including noncondensable gas behavior was reasonably predicted with an appropriate time step and CPU time, and the major thermal-hydraulic phenomena agreed well with the experiment. However, there were some code deficiencies such as an estimation of large system mass errors for the long transient and severe flow oscillations in the core region. These should be improved for more accurate and reliable calculation. In the YGN 3/4 simulation, the water-filled SG case delayed the coolant discharge to containment by ~2 h and the core heatup by ~1.3 h, as compared to the emptied-SG case, because of reduction of the pressurization rate that resulted from condensation on the SG U-tube wall. For the water-filled SGs, the amount of heat transfer into the secondary side was estimated at more than 60% of the total core power throughout the transient.