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 00:00:00 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: USDOE
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 5911606
• Report Number(s): CONF-790816-39; TRN: 79-019841
• Resource Relation: Conference: International meeting on fast reactor safety technology, Seattle, WA, USA, 19 Aug 1979
• Country of Publication: United States
• Language: English

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