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International Heat Transfer Conference 15

ISSN: 2377-424X (online)
ISSN: 2377-4371 (flashdrive)

A Study on Post-CHF Heat Transfer at Near-Critical Pressure

Takashi Mawatari
Kyushu University

Hideo Mori
Department of Mechanical Engineering, Kyushu University, Motoka 744, Nishi-ku, Fukuoka, 819-0395, Japan

Keishi Kariya
Saga University, Honjo-machi, Saga-shi, Saga 840 – 0052, Japan

DOI: 10.1615/IHTC15.fbl.009867
pages 2709-2722

KEY WORDS: Boiling and evaporation, Nuclear energy, Post-CHF Heat Transfer, Near-Critical Pressure, Flow Boiling


Supercritical pressurized water cooled reactor (SCWR), which has an once-through water cooled reactor for supplying supercritical pressure steam at high temperature to a turbine system, is one of the next generation reactors for the purpose of improving economic efficiency and safety. In the SCWR system, the water pressure passes through the critical pressure during startup, shutdown and in case of loss of coolant accident (LOCA). In the pressure region slightly below the critical pressure, critical heat flux (CHF) phenomenon tends to occur at relatively low heat flux, and then there is a risk of serious damage to fuel rod due to surface temperature rise. Therefore, it is significant for safety design of the SCWR to clarify characteristics of post-CHF heat transfer in such near-critical pressure region. In this study, experiments on post-CHF heat transfer in vertical upward and downward flows with a circular tube of 4.4mm I.D. were carried out at near-critical pressure condition (reduced pressure range of 0.92 to 0.99) in order to evaluate their characteristics. HCFC22 and HFC134a were used as the test fluid instead of water because of easier handling. Based on obtained experimental data, influences of pressure, mass flux and heat flux conditions on the characteristic of post-CHF heat transfer were clarified, and then the characteristic was classified into two types mainly by mass flux of around 700 kg/(m2·s).

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