Flow excursion experiments with a production reactor assembly mockup

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

OSTI ID:6178252

Rush, G C; Blake, J E ^[1]; Nash, C A ^[2]

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1. Babcock and Wilcox Co., Alliance, OH (USA)
2. Westinghouse Savannah River Co., Aiken, SC (USA)

A series of power ramp and loss-of-coolant accidents were simulated with an electrically heated mockup of a Savannah River Site production reactor assembly. The one-to-one scale mockup had full multichannel annular geometry in its heated section in addition to prototypical inlet and outlet endfitting hardware. Power levels causing void generation and flow instability in the water coolant flowing through the mockup were found under different transient and quasi-steady state test conditions. A reasonably sharp boundary between initial operating powers leading to or not leading to flow instability were found: that being 0.2 MW or less on power levels of 4 to 6.3 MW. Void generation occurred before, but close to, the point of flow instability. The data were taken in support of the Savannah River reactor limits program and will be used in continuing code benchmarking efforts. 6 refs., 12 figs., 2 tabs.

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (USA)
• Sponsoring Organization: USDOE; USDOE, Washington, DC (USA)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 6178252
• Report Number(s): WSRC-RP-89-1390; CONF-900917-36; ON: DE91007959; TRN: 91-009389
• Resource Relation: Conference: American Nuclear Society (ANS) topical meeting on safety of commercial nuclear reactor research and irradiation facilities, Boise, ID (USA), 30 Sep - 3 Oct 1990
• 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
SPECIAL PRODUCTION REACTORS
REACTOR SAFETY EXPERIMENTS
ANNULAR FUEL ELEMENTS
BENCHMARKS
COMPUTER CODES
COMPUTERIZED SIMULATION
FIBER OPTICS
FLOW RATE
FLOWMETERS
FUEL ASSEMBLIES
FUEL CHANNELS
LOAD MANAGEMENT
PIPES
PRIMARY COOLANT CIRCUITS
REACTOR ACCIDENTS
REACTOR INSTRUMENTATION
REACTOR KINETICS
REACTOR OPERATION
REACTOR STABILITY
RUPTURES
SAVANNAH RIVER PLANT
SCRAM
SIMULATION
STEADY-STATE CONDITIONS
THERMOCOUPLES
TIME DEPENDENCE
TRANSIENTS
VOID FRACTION
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FAILURES
FLUID MECHANICS
FUEL ELEMENTS
KINETICS
MANAGEMENT
MEASURING INSTRUMENTS
MECHANICS
METERS
NATIONAL ORGANIZATIONS
OPERATION
PRODUCTION REACTORS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SHUTDOWN
REACTORS
SHUTDOWN
STABILITY
US AEC
US DOE
US ERDA
US ORGANIZATIONS
220900* - Nuclear Reactor Technology- Reactor Safety
220700 - Nuclear Reactor Technology- Plutonium & Isotope Production Reactors