Comparison of COBRA III-C and SABRE-1 (wire-wrap version) computational results with steady-state data from a 19-pin internally guard heated sodium-cooled bundle with a six-channel central blockage (THORS Bundle 3C). [LMFBR]

Conference · Mon Jan 01 04:00:00 EST 1979

OSTI ID:5911606

Dearing, J F; Nelson, W R; Rose, S D

Computational thermal-hydraulic models of a 19-pin, electrically heated, wire-wrap liquid-metal fast breeder reactor test bundle were developed using two well-known subchannel analysis codes, COBRA III-C and SABRE-1 (wire-wrap version). These two codes use similar subchannel control volumes for the finite difference conservation equations but vary markedly in solution strategy and modeling capability. In particular, the empirical wire-wrap-forced diversion crossflow models are different. Surprisingly, however, crossflow velocity predictions of the two codes are very similar. Both codes show generally good agreement with experimental temperature data from a test in which a large radial temperature gradient was imposed. Differences between data and code results are probably caused by experimental pin bowing, which is presently the limiting factor in validating coded empirical models.

• Research Organization: Oak Ridge National Lab., TN (USA)
• Sponsoring Organization: Not Available; USDOE
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 5911606
• Report Number(s): CONF-790816-39
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500* -- Power Reactors
Breeding
BREEDER REACTORS
COMPUTER CALCULATIONS
CONFIGURATION
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL ELEMENTS
FUEL PINS
HEAT TRANSFER
HYDRAULICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MECHANICS
REACTOR COMPONENTS
REACTORS
TEMPERATURE DISTRIBUTION