RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS

Abstract An experiment was conducted for the OECD/NEA ROSA-2 Project using the large scale test facility (LSTF), which simulated a cold leg intermediate-break loss-of-coolant accident with 17% break in a pressurized water reactor. Assumptions were made such as single-failure of high-pressure and low-pressure injection systems of emergency core cooling system. In the LSTF test, core dryout took place because of a rapid drop in the core collapsed liquid level before loop seal clearing (LSC). Liquid was accumulated in upper plenum, U-tube upflow-side and inlet plena of steam generators before the LSC because of counter-current flow limiting (CCFL) by high steam velocity, which caused further decrease in the core collapsed liquid level. The post-test analysis by RELAP5/MOD3.3 code revealed that peak cladding temperature (PCT) was overpredicted because of the underprediction of the core collapsed liquid level due to inadequate prediction of accumulator flow rate. We created the phenomena identification and ranking table for each component from the viewpoint of the importance of phenomena in determining the PCT. We found the combination of multiple uncertain parameters including slope m and intercept C of the Wallis CCFL correlation at the upper core plate, core decay power, and steam convective heat transfer coefficient in the core within the defined uncertain ranges largely affected the PCT.