Critical Heat Flux Experiments and Correlation in a Long, Sodium-Heated Tube

[+] Author and Article Information
D. M. France, R. D. Carlson, T. Chiang

Argonne National Laboratory, Argonne, Ill.

W. J. Minkowycz

Department of Energy Engineering, University of Illinois, Chicago, Ill. 60680

J. Heat Transfer 103(1), 74-80 (Feb 01, 1981) (7 pages) doi:10.1115/1.3244434 History: Received May 29, 1980; Online October 20, 2009


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, full-scale 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 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/m2 ̇s. 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.

Copyright © 1981 by ASME
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