Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

다국어 초록 (Multilingual Abstract)

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF),which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor(PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover legdownflow-side before loop seal clearing, and water remaining occurred on the upper core plate in theupper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences ofthe combination of multiple uncertain parameters on peak cladding temperature within the defineduncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomenaobserved in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project withthe Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF testsimulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-sidedepressurization as an accident management measure and nitrogen gas inflow. Some discrepanciesappeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level,and the cladding surface temperature probably due to effects of differences between the LSTF and thePKL in configuration, geometry, and volumetric size.

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