Dispersed-flow heat transfer during reflood in a pressurized water reactor after a large-break loss-of-coolant accident
Conference
·
· Trans. Am. Nucl. Soc.; (United States)
OSTI ID:6907166
The postulated loss-of-coolant accident (LOCA) of a pressurized water reactor has been the subject of intensive experimental and analytical studies in light water reactor safety analysis. Many efforts are devoted to the investigation of the thermodynamic behavior of the reactor core and the effectiveness of the emergency-core-cooling system during reflood phase of a LOCA. In the initial period of reflood phase, the flow patterns at the core higher elevation are considered as the dispersed flow regime. The effect of liquid phase on the heat transfer cannot be neglected in this dispersed flow regime. It has been found experimentally that a steam-water-droplet flow is ultimately responsible for terminating the cladding temperature excursions in a LOCA. Recently, a study of reflood test predictions with RELAP5/MOD2 showed a tendency to predict lower cladding temperatures during the dispersed flow regime. The purpose of this study is to present a new dispersed flow film boiling, implemented in RELAP5/MOD2. In addition, the RELAP underestimated the vapor temperature. Therefore, a revised interfacial heat transfer coefficient between the droplets and steam was incorporated in the code. Comparison of the current predicted cladding/fuel temperature histories with the reflood data showed a fair agreement.
- Research Organization:
- Texas A and M, College Station
- OSTI ID:
- 6907166
- Report Number(s):
- CONF-861102-
- Conference Information:
- Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 53
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
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COMPUTER CODES
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ENERGY TRANSFER
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FILM BOILING
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FLUID FLOW
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HEAT TRANSFER
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LOSS OF COOLANT
MATHEMATICAL MODELS
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PHASE TRANSFORMATIONS
PIPES
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
RUPTURES
SAFETY
SIMULATION
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TWO-PHASE FLOW
WATER COOLED REACTORS
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210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BOILING
COMPUTER CODES
COMPUTERIZED SIMULATION
ENERGY TRANSFER
FAILURES
FILM BOILING
FLOW STRESS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
PHASE TRANSFORMATIONS
PIPES
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
RUPTURES
SAFETY
SIMULATION
STRESSES
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS