Comparisons of TRAC-PF1 calculations with semiscale Mod-3 small-break tests S-07-10D, S-SB-P1, and S-SB-P7. [PWR]
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.
- Research Organization:
- Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 5040552
- Report Number(s):
- LA-UR-82-2298; CONF-820802-14; ON: DE82021922
- Resource Relation:
- Conference: International meeting on thermal nuclear reactor safety, Chicago, IL, USA, 29 Aug 1982; Other Information: Portions of document are illegible
- Country of Publication:
- United States
- Language:
- English
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