Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube
- Institute of Advanced Energy, Kyoto Univ., Gokasho, Uji, Kyoto 611-0011 (Japan)
- Dept. of Energy Science and Technology, Kyoto University, Kyoto University, Kyoto 606-8501 (Japan)
- National Institute for Fusion Science, 322-6, Oroshi-cho, Toki, GIFU, 509-5292 (Japan)
The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcooling ({delta}T{sub sub,in} = 48.6 to 154.7 K), the inlet pressure (P{sub in} = 735.2 to 969.0 kPa) and the increasing heat input (Q{sub 0} exp(t/t), t = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, R{sub a} = 3.18 {mu}m), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with R{sub a} = 0.18 {mu}m and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d 22.2 and 11.6 respectively with R{sub a} = 0.45 {mu}m are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcooling. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (authors)
- Research Organization:
- The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)
- OSTI ID:
- 20997101
- Resource Relation:
- Conference: 14. International conference on nuclear engineering (ICONE 14), Miami - Florida (United States), 17-20 Jul 2006; Other Information: Country of input: France
- Country of Publication:
- United States
- Language:
- English
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