Comparisons of TRAC-PF1 Calculations with Semiscale Mod-3 Small-Break Tests S-07-10D, S-SB-P1, and S-SB-P7

摘要

Semiscale Tests S-07-10D, S-SB-P1, and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Test S-07-10D simulated an equivalent pressurized-water-reactor (PWR) 10% communicative cold-leg break for an early pump trip with an emergency core coolant (ECC) injected only into the intact-loop cold leg. Tests S-SB-P1 and S-SB-P7 simulated 2.5% communicative cold-leg breaks for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics. (ERA citation 07:059592)