Subcooled Decompression Analysis in PWR LOCA

The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermohydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and omplexity of flow geometry. Good agreement is generally obtained.