Critical heat flux experiments and correlation in a long, sodium-heated tube

Journal Article · Sun Feb 01 04:00:00 EST 1981 · J. Heat Transfer; (United States)

DOI:https://doi.org/10.1115/1.3244434· OSTI ID:6252145

France, D M; Carlson, R D; Chiang, T; Minkowycz, W J

Critical heat flux (CHF) experiments were performed in the Steam Generator Test Facility (SGTF) at Argonne National Laboratory for application to liquid metal fast breeder reactor steam generators. The test section consisted of a single, straight, vertical, fullscale LMFBR steam generator tube with force-circulated water boiling upwards inside the tube heated by sodium flowing countercurrent in a surrounding annulus. The test section tube parameters were as follows: 10.1 mm i.d., 15.9 mm o.d., material = 2 1/4 Cr-1 Mo of 7.0 to steel, and 13.1 m heated length. Experiments were performed in the water pressure range of 7.0 to 15.3 MPa and the water mass flux range of 720 to 3200 kg/m/sup 2/xs. The data exhibited two trends: heat flux independent and heat flux dependent. Empirical correlation equations were developed from over 400 CHF tests performed in the SGTF. The data and correlation equations were compared to the results of other CHF investigations.

• Research Organization: Argonne National Laboratory, Argonne, III.
• OSTI ID: 6252145
• Journal Information: J. Heat Transfer; (United States), Journal Name: J. Heat Transfer; (United States) Vol. 103:1; ISSN JHTRA
• Country of Publication: United States
• Language: English

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