Experimental study of downflow critical heat flux in multiannular SRS fuel assembly channels at low air-water flows
The problem addressed in this experimental study is the measurement of critical or dryout heat flux in multi-annular fuel assembly flow passages with low downward flows of air-water mixtures. These thermal hydraulic conditions pertain to specific conditions predicted for Savannah River Site reactors during hypothetical large loss-of-coolant accidents. Experimental data obtained on a full scale prototypic simulation of the multi-annular fuel assembly is important in establishing the safety margin of the reactor operating power. The SRS reactors, like some research reactors, utilize downwards flow of coolant through narrow parallel flow channels during normal operation. These channels are formed by concentric heated tubes of high thermal conductivity uranium-aluminum metal that are cooled on both sides. Ribs on the tubes subdivide the flow channels into curved subchannels which may be considered somewhat similar to the flat rectangular channels of research reactors. However, gaps between the ribs and the adjoining tube allow cross flows between subchannels. For this accident, preliminary analysis predict that downward flow of emergency coolant would entrain large amounts of air through the fuel assembly. Due to the above special conditions, no data has been found to be fully applicable to the SRS reactor. An experimental study was thus required to obtain prototypical data and investigate physical mechanisms to aid the development of analytical models in the code FLOWTRAN-TF. Comparison of the data with analysis will be reported in the future after code benchmarking. 5 refs.
- Research Organization:
- Westinghouse Savannah River Co., Aiken, SC (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- AC09-89SR18035
- OSTI ID:
- 5210430
- Report Number(s):
- WSRC-MS-91-411; CONF-920804-13; ON: DE92015894
- Resource Relation:
- Conference: American Society of Mechanical Engineers national heat transfer conference and exposition, San Diego, CA (United States), 9-12 Aug 1992
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL ASSEMBLIES
COOLING
REACTOR COOLING SYSTEMS
HEAT TRANSFER
SPECIAL PRODUCTION REACTORS
AIR
DESIGN BASIS ACCIDENTS
F CODES
HEAT FLUX
HYDRAULICS
LOSS OF COOLANT
REACTOR SAFETY
TESTING
TUBES
TWO-PHASE FLOW
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GASES
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OXYGEN COMPOUNDS
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
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220900* - Nuclear Reactor Technology- Reactor Safety
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Test & Experimental Reactors