Dispersed-flow heat transfer during reflood in a pressurized water reactor after a large-break loss-of-coolant accident

Conference · Wed Jan 01 00:00:00 EST 1986 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:6907166

Hassan, Y A

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-
• Journal Information: Trans. Am. Nucl. Soc.; (United States), Vol. 53; Conference: American Nuclear Society and Atomic Industrial Forum joint meeting, Washington, DC, USA, 16 Nov 1986
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
PWR TYPE REACTORS
REACTOR SAFETY
COMPUTER CODES
COMPUTERIZED SIMULATION
FILM BOILING
FLOW STRESS
MATHEMATICAL MODELS
PIPES
R CODES
REACTOR ACCIDENTS
RUPTURES
TWO-PHASE FLOW
ACCIDENTS
BOILING
ENERGY TRANSFER
FAILURES
FLUID FLOW
FLUID MECHANICS
MECHANICS
PHASE TRANSFORMATIONS
REACTORS
SAFETY
SIMULATION
STRESSES
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled