Analysis of select Mod-1 semiscale blowdown heat transfer tests. Final report. [PWR]
The report contains the RELAP4 analysis and sensitivity studies of Semiscale Tests S-02-2 and S-02-7. The Semiscale System is an electrically heated experiment designed to produce data on system performance typical of PWR thermal-hydraulic behavior. The RELAP4 program used for these analyses is a digital computer program developed to predict the thermal-hydraulic behavior of experimental systems and water-cooled nuclear reactors subjected to postulated transients. The results of the analysis for Test S-02-2 were in very good agreement with the data. Two parameters which required improvement were identified. These were the lower plenum density and the mass flow on the vessel side of the break. Subsequently, before analyzing Test S-02-7, the lower plenum was renodalized and the critical flow model at the vessel side break was modified. The results of the analysis of Test S-02-7 compared more favorably with the data than those of S-02-2. Additional sensitivity studies included time step studies, steam generator and downcomer modeling, and core nodalization.
- Research Organization:
- Energy, Inc., Idaho Falls, ID (USA)
- OSTI ID:
- 7337166
- Report Number(s):
- EPRI-NP-206; TRN: 77-004310
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BLOWDOWN
HEAT TRANSFER
PWR TYPE REACTORS
COMPUTER CALCULATIONS
ECCS
LOSS OF COOLANT
MATHEMATICAL MODELS
MOCKUP
PERFORMANCE
REACTOR SAFETY
SIMULATION
ACCIDENTS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
STRUCTURAL MODELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled