Modeling the onset of flow instability for subcooled boiling in downflow

Conference · Mon Jan 01 04:00:00 EST 1990

OSTI ID:6397988

Qureshi, Z ^[1]; Barry, J J; Crowley, C J ^[2]

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1. Westinghouse Savannah River Co., Aiken, SC (USA)
2. Creare, Inc., Hanover, NH (USA)

A postulated loss-of-coolant accident (LOCA) scenario for the Savannah River Plant (SRP) production reactors involves a double-ended break of a reactor primary coolant pipe. The flow of coolant (D{sub 2}O) in the reactor may decrease in such an event. As the flow into the reactor decreases, boiling may occur, followed by dryout and failure of the fuel due to overheating. A typical SRP fuel assembly consists of multiple concentric tubes containing the fuel and target materials. Coolant passes through the annular passages in the assembly in downflow. Under normal operating conditions, the flow rate is maintained high enough to suppress or minimize subcooled boiling, i.e. the flow remains essentially single phase throughout. At high coolant flow rates, the flow is single phase or partially developed subcooled boiling, and the pressure drop decreases with decreasing flow rate. Here friction dominates the pressure gradient, and the flow is stable. Below a certain flow rate, however, pressure drop may increase with decreasing flow rate. This occurs when significant voids are produced by boiling, resulting in a large acceleration component to the pressure drop. The negative slope of the curve leads to an instability because the pressure drop cannot adjust to compensate -- the flow is driven to a lower value. Overheating of the channel may result. 15 refs., 14 figs.

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (USA)
• Sponsoring Organization: DOE/DP
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 6397988
• Report Number(s): WSRC-MS-90-267; CONF-9009219--4; ON: DE91005153
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
A CODES
ACCIDENTS
BOILING
COMPARATIVE EVALUATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
DRYOUT
ENERGY TRANSFER
FLOW RATE
FLUID MECHANICS
FUEL ASSEMBLIES
HEAT TRANSFER
HYDRAULICS
INSTABILITY
LOSS OF COOLANT
MECHANICS
NATIONAL ORGANIZATIONS
PHASE TRANSFORMATIONS
PIPES
PRESSURE DROP
PRODUCTION REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SAVANNAH RIVER PLANT
SIMULATION
STEADY-STATE CONDITIONS
SUBCOOLED BOILING
US AEC
US DOE
US ERDA
US ORGANIZATIONS