Columbia University flow instability experimental program: Volume 6. Single annulus tests, transient test program

Technical Report · Tue Sep 01 04:00:00 EDT 1992

DOI:https://doi.org/10.2172/10168364· OSTI ID:10168364

Dougherty, T; Maciuca, C; McAssey, Jr, E V; Reddy, D G; Yang, B W

The coolant in the Savannah River Site (SRS) production nuclear reactor assemblies is circulated as a subcooled liquid under normal operating conditions. This coolant is evenly distributed throughout multiple annular flow channels with a uniform pressure profile across each coolant flow channel. During the postulated Loss of Coolant Accident (LOCA), which is initiated by a hypothetical guillotine pipe break, the coolant flow through the reactor assemblies is significantly reduced. The flow reduction and accompanying power reduction (after shutdown is initiated) occur in the first 1 to 2 seconds of the LOCA. This portion of the LOCA is referred to as the Flow Instability phase. This report presents the experimental results for the transient portion of the single annulus test program. The test program was designed to investigate the onset of flow instability in an annular geometry similar to the MARK 22 reactor. The test program involved testing of both a ribless heater and a ribbed heater under steady state as well as transient conditions. The ribbed heater testing is currently underway and will be reported separately. The steady state portion of this test program with ribless heater was completed and reported in report No. CU-HTRF-T3A. The present report presents transient test results obtained from a ribless, uniform annulus test section. A total of thirty five transients were conducted with six cases in which flow excursion occurred. No unstable conditions resulted for tests in which the steady state Q{sub ratio} OFI limit was not exceeded.

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 10168364
• Report Number(s): WSRC-TR--93-687; ON: DE94015499; IN: CU-HTRF-T3B
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220600
220900
ACCURACY
DATA ACQUISITION SYSTEMS
ERRORS
EXPERIMENTAL DATA
FLOWMETERS
FLUID FLOW
FUEL ASSEMBLIES
FUEL CHANNELS
HYDRODYNAMICS
LOSS OF COOLANT
MEASURING INSTRUMENTS
PRESSURE MEASUREMENT
REACTOR SAFETY
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
SAVANNAH RIVER PLANT
SPECIAL PRODUCTION REACTORS
TEMPERATURE MEASUREMENT
TEST FACILITIES
TRANSIENTS
UNSTEADY FLOW