For a steam generator (SG) in a commercialized sodium-cooled fast breeder reactor (FBR), flow instability in the water side is one of the most important items needing research. As the first step of this research, thermal-hydraulic experiments using water as the test fluid were performed under high pressure conditions at the Japan Atomic Energy Agency (JAEA) by using a circular tube. Void fraction, pressure drop, and heat transfer coefficient data were obtained under 15, 17, and 18 MPa. This paper discusses the steam-water pressure drop and void fraction. Using the obtained data, we evaluated existing correlations for void fraction and two-phase flow multipliers under high pressure. As a result, the drift flux model implemented in the TRAC-BF1 code was confirmed to suitably predict the void fraction well under the present high pressure conditions. For the two-phase flow multiplier, the Chisholm correlation and the homogeneous model were confirmed to be the best under the present high-pressure conditions.
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Pressure Drop and Void Fraction in Steam-Water Two-Phase Flow at High Pressure
Kazuyuki Takase
Kazuyuki Takase
Japan Atomic Energy Agency (JAEA)
,2-4 Shirakata Tokai
,Ibaraki 319-1195
, Japan
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Wei Liu
e-mail: liu.wei@jaea.go.jp
Kazuyuki Takase
Japan Atomic Energy Agency (JAEA)
,2-4 Shirakata Tokai
,Ibaraki 319-1195
, Japan
Contributed by the Heat Transfer Division of ASME for publication in the Journal of Heat Transfer. Manuscript received February 10, 2011; final manuscript received January 23, 2013; published online July 11, 2013. Assoc. Editor: Wei Tong.
J. Heat Transfer. Aug 2013, 135(8): 081502 (13 pages)
Published Online: July 11, 2013
Article history
Received:
February 10, 2011
Revision Received:
January 23, 2013
Citation
Liu, W., Tamai, H., and Takase, K. (July 11, 2013). "Pressure Drop and Void Fraction in Steam-Water Two-Phase Flow at High Pressure." ASME. J. Heat Transfer. August 2013; 135(8): 081502. https://doi.org/10.1115/1.4023678
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