Interfacial area transport of subcooled boiling flow in a vertical annulus

被引:19
作者
Brooks, Caleb S. [1 ]
Ozar, Basar [1 ]
Hibiki, Takashi [1 ]
Ishii, Mamoru [1 ]
机构
[1] Purdue Univ, Sch Nucl Engn, W Lafayette, IN 47907 USA
关键词
BUBBLE DEPARTURE FREQUENCY; 2-PHASE FLOW; HEAT-FLUX; EQUATION; WATER; MODEL; PRESSURES;
D O I
10.1016/j.nucengdes.2013.04.041
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
In an effort to improve the prediction of void fraction and heat transfer characteristics in two-phase systems, the two-group interfacial area transport equation has been developed for use with the two-group two-fluid model. The two-group approach treats spherical/distorted bubbles as Group-1 and cap/slug/churn-turbulent bubbles as Group-2. Therefore, the interfacial area transport of steam-water two-phase flow in a vertical annulus has been investigated experimentally, including bulk flow parameters and wall nucleation characteristics. The theoretical modeling of interfacial area transport equation with phase change terms is introduced and discussed along with the experimental results. Benchmark of the interfacial area transport equation is performed considering the effects of bubble interaction mechanisms such as bubble break-up and coalescence, as well as, effects of phase change mechanisms such as wall nucleation and condensation for subcooled boiling. From the benchmark, sensitivity in the constitutive relations for Group-1 phase change mechanisms, such as wall nucleation and condensation is clear. The Group-2 interfacial area transport is shown to be dominated by the interfacial heat transfer mechanism causing expansion of Group-1 bubbles into Group-2 bubbles in the boiling flow. (C) 2013 Elsevier B.V. All rights reserved.
引用
收藏
页码:152 / 163
页数:12
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