Critical Heat Flux of Flowing Water in Tube for Pressure Up to Near Critical Point-Experiment and Prediction

被引：5

作者：

Chen, Yuzhou ^{[1
]}

Zhao, Minfu ^{[1
]}

Bi, Keming ^{[1
]}

Yang, Bin ^{[1
]}

Zhang, Dongxu ^{[1
]}

Du, Kaiwen ^{[1
]}

机构：

[1] China Inst Atom Energy, Dept Reactor Engn Res & Design, POB 275 59, Beijing 102413, Peoples R China

来源：

JOURNAL OF NUCLEAR ENGINEERING AND RADIATION SCIENCE | 2018年 / 4卷 / 01期

关键词：

critical heat flux; near-critical pressure; subcooled and saturated condition;

D O I：

10.1115/1.4038215

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) experiment with uniform heating was performed in a tube of 8.2 mm in inner diameter and 2.4 m in heated length. The water flowed upward through the test section. The pressure covered the range from 8.6 to 20.8 MPa, mass flux 1157 to 3776 kg/m(2)s, inlet quality -2.79 to -0.08 (subcooling 19-337 degrees C), and local quality -0.97 to 0.53. For the pressure close to the near-critical point, the CHF decreased substantially with the pressure increasing. For the subcooling larger than a certain value, the CHF was related to the local condition. But for low subcooling and saturated condition, the CHF was related to the total power. The present results were in agreement with the previous experiment for the same local subcooled condition. Based on the present experimental results with subcooled and saturated conditions an empirical relation of the CHF was presented.

引用

页数：5

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