The Japan Atomic Energy Research Institute (JAERI) and the United States Nuclear Regulatory Commission (USNRC) are jointly conducting confirmatory, integral testing on the Westinghouse AP600 reactor transient responses by using the ROSA-V Large Scale Test Facility of JAERI. This facility, built originally to simulate conventional 4-loop pressurized water reactors (PWRS) , has been modified by adding components specific to the AP600 design. The modified LSTF now provides a full-pressure, full-height, 1/30.5 volumetrically-scaled simulation of AP600. Five loss-of-coolant accident (LOCA) experiments were performed by August 1994, simulating transients initiated by cold leg breaks, a Pressure Balance Line (PBL) break, and inadvertently open Automatic Depressurization System (ADS) valves. These experiments indicated adequate core cooling and decay heat removal performance of the AP600 passive safety components.
The effectiveness of an operator-initiated steam generator (SG) secondary-side depressurization on the core cooling performance during small-break loss of coolant accidents (SBLOCAs) in a pressurized water reactor (PWR) with total failure of the high pressure injection (HPI) systems is studied. The study is based on experiments conducted in the ROSA-V Large Scale Test Facility (LSTF) and analyses with the RELAP5/Mod3 code. The sensitivity of the core minimum liquid level and peak cladding temperature (PCT) to the secondary-side depressurization rate and the initiation time of the depressurization is evaluated analytically for various break sizes. It is shown that the P C T takes a maximum value for break areas between 1.0% and 1.5% of the cold leg cross-sectional area. The conditions which the depressurization rate and the initiation time should satisfy to limit the maximum P C T are derived.
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