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Two experiments for a small-break loss-of-coolant accident on a pressurizer top were conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST)facility to investigate the thermal-hydraulic behavior of a passive core cooling system (PCCS) in a Westinghouse pressurized water reactor (PWR). The test results are compared with previous IIST tests under the same initial and boundary conditions for a power operated relief valve (PORV) stuck-open incident. The objectives of this study are to understand the key thermal hydraulic phenomena associated with the PCCS and to compare the effectiveness of accident management with or without the PCCS. The break sizes are scaled down based on one and all three fully opened PORVs for a conventional PWR without the PCCS. This paper identifies the key phenomena commonly observed and the phenomena unique to a PWR with a PCCS. [ABSTRACT FROM AUTHOR]