Using an IIST SBLOCA Experiment to Assess RELAP5/MOD3.2

Abstract

The RELAP5/MOD3.2 code is used at the Institute of Nuclear Energy Research Integral System Test Facility to analyze a 2% cold-leg-break experiment that includes failure of the high-pressure injection system. The assessment code predictions include primary pressure, inventory distribution in the reactor coolant system (RCS), loop flow rate, break flow rate, and core thermal hydraulics. A comparison between the calculated results and the experimental data shows (a) a good match with the predictions of the RCS pressure and hot- and cold-leg fluid temperatures, (b) underprediction of the core and downcomer levels, (c) overprediction of the loop flow rates in single- and two-phase natural circulation, and (d) inadequate prediction of asymmetric coolant holdup in the three steam generators. Also presented are sensitivity studies of choked flow associated with the defaulted values of discharge coefficients in the simulation of the break flow, and of the core bypass area to evaluate the effect of core level depression.