Effect of pressure on critical heat flux in uniformly heated vertical annulus under low flow conditions

被引:41
|
作者
Chun, SY
Chung, HJ
Moon, SK
Yang, SK
Chung, MK
Schoesse, T
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)00307-1
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
Critical heat flux (CHF) experiments have been carried out in a wide range of pressure for an internally heated vertical annulus. The experimental conditions covered a range of pressure from 0.57 to 15.01 MPa, mass fluxes of 0 kg m(-2) s(-1) and from 200 to 650 kg m(-2) s(-1), and inlet subcoolings from 85 to 413 kJ kg(-1). Most of the CHFs were identified to the dryout of the liquid him in the annular-mist flow. For the mass fluxes of 550 and 650 kg m(-2) s(-1), the CHFs had a maximum value at a pressure of 2-3 MPa, and the pressure at the maximum CHF values had a trend moving toward the pressure at the peak value of pool boiling CHF as the mass flux decreased. The CHF data under a zero mass flux condition indicate that both the effects of pressure and inlet subcooling on the CHF were smaller, compared with those for the CHF with a net water upflow. The Doerffer correlation using the 1995 CHF look-up table and the Bowring correlation show a good prediction capability for the present CHF data. (C) 2001 Elsevier Science B.V. All rights reserved.
引用
收藏
页码:159 / 174
页数:16
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