Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Technical Report · Fri Apr 01 04:00:00 EST 1977

DOI:https://doi.org/10.2172/6440323· OSTI ID:6440323

Wolf, S.; Holmes, D. H.

An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

• Research Organization: General Electric Co., Sunnyvale, CA (USA). Fast Breeder Reactor Dept.
• DOE Contract Number: AT03-76SF70010
• OSTI ID: 6440323
• Report Number(s): DOE/SF/70010-T23; FBRD-00020
• Country of Publication: United States
• Language: English

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