Maximum wake temperature and Nusselt number behind blockages in sodium-cooled bundles
Conference
·
OSTI ID:5289681
Two important correlations have been obtained to calculate the maximum coolant temperature and the Nusselt number in the wake (the flow recirculating zone) downstream of blockages in 19-pin sodium-cooled bundles. These correlations can be applied to predict the maximum temperature rise and the average heat transfer coefficient behind small non-heat-generating blockages in the fuel assemblies of the FFTF and the CRBR for the wide range of flow and power conditions. Experiments with partial blockages in simulated LMFBR fuel assemblies have been performed at the THORS facility (formerly called FFM) of the Oak Ridge National Laboratory. Nineteen-pin sodium-cooled bundles were used, which had the same pin diameter and pitch as the CRBR and the FFTF fuel assemblies.
- Research Organization:
- Oak Ridge National Lab., Tenn. (USA)
- DOE Contract Number:
- W-7405-ENG-26
- OSTI ID:
- 5289681
- Report Number(s):
- CONF-771109-78
- Country of Publication:
- United States
- Language:
- English
Similar Records
Sodium boiling tests in a 19-pin internally guard heated simulated LMFBR fuel assembly with a six-channel internal blockage: record of experimental data for THORS bundle 3C
Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)
Effect of partial blockages in simulated LMFBR fuel assemblies
Technical Report
·
Tue May 01 00:00:00 EDT 1979
·
OSTI ID:6053419
Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)
Technical Report
·
Sat Sep 01 00:00:00 EDT 1979
·
OSTI ID:5916051
Effect of partial blockages in simulated LMFBR fuel assemblies
Technical Report
·
Fri Nov 30 23:00:00 EST 1973
·
OSTI ID:4357059
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500* -- Power Reactors
Breeding
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
BREEDER REACTORS
CLINCH RIVER BREEDER REACTOR
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FFTF REACTOR
FUEL ASSEMBLIES
HYDRAULICS
LIQUID METAL COOLED REACTORS
MOCKUP
NUSSELT NUMBER
POWER REACTORS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STRUCTURAL MODELS
TEMPERATURE GRADIENTS
TEST REACTORS
210500* -- Power Reactors
Breeding
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
BREEDER REACTORS
CLINCH RIVER BREEDER REACTOR
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FFTF REACTOR
FUEL ASSEMBLIES
HYDRAULICS
LIQUID METAL COOLED REACTORS
MOCKUP
NUSSELT NUMBER
POWER REACTORS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STRUCTURAL MODELS
TEMPERATURE GRADIENTS
TEST REACTORS