Critical heat flux under zero flow conditions in vertical annulus with uniformly and non-uniformly heated sections

被引:9
|
作者
Chun, SY
Moon, SK
Chung, HJ
Yang, SK
Chung, MK
Aritomi, M
机构
[1] Korea Atom Energy Res Inst, Taejon 305353, South Korea
[2] Tokyo Inst Technol, Nucl Reactors Res Lab, Meguro Ku, Tokyo 152, Japan
关键词
D O I
10.1016/S0029-5493(00)00377-0
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
The experimental study of water CHF (critical heat flux) under zero how conditions has been carried out in an annulus flow channel with uniformly and non-uniformly heated sections over a pressure range of 0.52-14.96 MPa. In the present boiling system, the CHFs occur in the upper region of the heated section, in contrast to the results in the experiments for boiling tubes conducted by several investigators. The general trend of the CHF with pressure is that the CHF increases up to a medium pressure of about 6-8 MPa and decreases as the pressure is further increased. A comparison of the present data with the existing flooding CHF correlations shows that the correlations depend greatly on the effect of the heat flux distribution. When the correction terms with the density ratio and the effect of the heat flux distribution proposed in the present work are used with the CHF correlation based on the Wallis flooding correlation, it predicts the measured flooding CHF within an RMS error of 9.0%. (C) 2001 Elsevier Science B.V. All rights reserved.
引用
收藏
页码:265 / 279
页数:15
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