An experimental study on the critical heat flux for low pressure of water in a uniformly heated vertical tube
For pressurized-water nuclear reactor design, critical heat flux (CHF) is one of the most important thermohydrodynamic parameters for the security and economy of the reactor. This paper mainly presents the test section, measure of parameters, technique research of low pressure CHF experiment and analysis of experimental result etc. In the experimental study, the experimental methods and key experimental technique are explored by a uniformly heated vertical tube in low pressure. A number of different coordinate systems are used to present the experimental results, these include CHF as a function of the inlet subcooling, as a function of the pressure in outlet, as a function of the mass flow rate. The experiment totally obtains thirty-two data for a range of exit qualities (-0.05-0.20) and exit pressures ranging from 2 to 10 bar, and coordinate the empirical correlation for predicting CHF.
low pressure CHF experimental study uniformly heated vertical tube
Zhao Erlei Xiong Wanyu Lang Xuemei
National Key Laboratory of Bubble Physics and Natural Circulation, Nuclear Power Institute of China, National Key Laboratory of Bubble Physics and Natural Circulation,Nuclear Power Institute of China,
国际会议
上海
英文
1445-1453
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）