Assessment of RELAP5/MOD2 Using Semiscale Large Break Loss-of-Coolant Experiment S-06-3 (NUREG/IA-0046)
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Publication Information
Date Published: April 1992
Prepared by:
Kuo-Shing Iliang, Lainsu Kao, Jeng-Lang Chiou,
Lih-Yih Liao, Song-Feng Wang, Yi-Bin Chen
Institute of Nuclear Energy Research
P.O. Box 3, Lung-Tan 32500
Taiwan, Republic of China
Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Thermal-Hydraulic Code Assessment
and Application Program (ICAP)
Published by:
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555
Availability Notice
Abstract
This report presents the results of the-RELAP5/MOD2 posttest assessment utilizing a Semiscale large break loss-of-coolant experiment numbered S-06-3. Test S-06-3 is a 200% double ended cold leg break experiment performed in Semiscale Mod-i facility in 1978 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large LOCA in a pressurized water reactor CPWR) system and providing a data base for a U.S. Nuclear Regulatory Commission standard problem. Through extensive comparisons between data and best-estimate RELAP5 calculations, the capabilities of RELAPS to calculate the large LOCA accident were assessed. Emphasis was placed on the capability of the code to calculate break flow rates during system blowdown stage, emergency core cooling system (ECCS) injection bypass during refill stage, guenching during ref lood stage, and the peak cladding temperature (PCT) behavior throughout the whole experiment. Besides, effects of several different modelings which include radial connections between core hot and average channels, maximum number of heat slab axial interval for 2-D ref lod calculation, number of nodes representing the core, cross-flow junctions on vessel entrances, ref lood calculation etc., were all investigated.
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