Assessment of RELAP5/MOD2 using Semiscale large break loss-of-coolant experiment S-06-3

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-1 facility in 1978 for the purpose of investigating the thermal and hydraulic phenomena accompanying a hypothetical large LOCA in a pressurized water reactor (PWR) system and providing a data base for a US Nuclear Regulatory Commission standard problem. Through extensive comparisons between data and best-estimate RELAP5 calculations, the capabilities of RELAP5 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, quenching during reflood 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 reflood calculation, number of nodes representing the core, cross-flow junctions on vessel entrances, reflood calculation etc., were all investigated.