Steam generator behavior during loss-of-feedwater and loss-of-offsite-power ATWS experiments in LOFT

Two Anticipated Transient Without Scram (ATWS) experiments, L9-3 and L9-4, were conducted in the Loss-of-Fluid Test (LOFT) facility. The LOFT facility is a volumetrically scaled (1/44) pressurized water reactor (PWR) system with a nuclear core and is designed for integral loss-of-coolant accident/emergency core cooling experiments. Experiment L9-3 simulated a loss-of-feedwater ATWS in a commercial PWR; Experiment L9-4 simulated a loss-of-offsite-power ATWS. The system transient behavior in each experiment was dominated by interaction between the primary-to-secondary heat removal rate in the steam generator and by reactor kinetics in the core. Comparisons of RELAP5/MOD1 calculational results to the measured test data show that the degradation of the primary-to-secondary heat transfer and the transient response of the primary coolant system in both experiments were calculated well.