Steady thermal hydraulic characteristics of nuclear steam generators based on the drift flux code model  

作　　者：张小英 陈焕栋 白宁 朱元兵 任志豪 黄凯 

机构地区：[1]School of Electric Power, South China University of Technology [2]China Nuclear Power Technology Research Institute

出　　处：《Nuclear Science and Techniques》2014年第5期78-85,共8页核技术（英文）

基　　金：Supported by the National Natural Science Foundation of China(Nos.51376065 and 51176052)

摘　　要：To investigate the steady thermal hydraulic characteristics of U-tube steam generator(SG), a 1D simulation code based on the four-equation drift flux model is developed. The U-tube channels presumably consist mainly of the primary channel, secondary channel, and tube wall. In the sub-cooling regions of the primary and secondary channels, flow is simulated using the single-phase flow model, whereas that in the boiling regions of the secondary channels is simulated using the four-equation drift flux model. The first-order equations of upwind difference are derived based on the staggered grid. Steady-state thermal hydraulic parameters are obtained with a cross-iteration scheme of heat balance and natural circulation requirement. The developed code is applied to analyze the SG behavior of the Qinshan I Nuclear Power Plant under 100%, 75%, 50%, 30%, and 15% power conditions. Analysis results are then compared with the simulation results obtained using RELAP5.To investigate the steady thermal hydraulic characteristics of U-tube steam generator （SG）, a 1D simulation code based on the four-equation drift flux model is developed. The U-tube channels presumably consist mainly of the primary channel, secondary channel, and tube wall. In the sub-cooling regions of the primary and sec- ondary channels, flow is simulated using the single-phase flow model, whereas that in the boiling regions of the secondary channels is simulated using the four-equation drift flux model. The first-order equations of upwind difference are derived based on the staggered grid. Steady-state thermal hydraulic parameters are obtained with a cross-iteration scheme of heat balance and natural circulation requirement. The developed code is applied to analyze the SG behavior of the Qinshan I Nuclear Power Plant under 100%, 75%, 50%, 30%, and 15% power conditions. Analysis results are then compared with the simulation results obtained using RELAP5.

关 键 词：热工水力特性 漂移流模型 蒸汽发生器 秦山核电站 代码 基础 通量 一维模拟 

分 类 号：TL353.13[核科学技术—核技术及应用]