海洋核动力平台PRHR HX管内蒸汽冷凝换热特性分析

Heat transfer characteristic analysis of condensation for marine nuclear power platform PRHR HX

[目的]旨在探究非能动余热排出换热器(PRHR HX)管内饱和蒸汽冷凝传热特性,为海洋核动力平台非能动安全系统设计提供支撑。[方法]通过搭建的功率比1∶50的试验装置,使用分离热阻法处理试验结果,对PRHR HX管内饱和蒸汽冷凝换热特性进行分析。[结果]在试验参数范围内,PRHR HX管内主要以分层流或波状流-环状流-波状流流型存在,该管内凝液流动存在由层流到湍流的转捩过程;管内饱和蒸汽冷凝换热系数随压力升高而增大;在压力为0.52 MPa时,换热器内蒸汽流速最大值约为6.72 m/s,当压力大于0.52 MPa后,蒸汽流速反而逐渐减小;所提PRHR HX管内饱和蒸汽冷凝换热系数计算关系式与试验结果吻合良好,其计算值与试验值的相对误差在±8%以内。[结论]研究结果可为海洋核动力平台及类似应用对象非能动安全系统PRHR HX设计和优化提供参考。

[Objectives]The heat transfer characteristics of steam condensation in a passive residual heat re-moval system heat exchanger(PRHR HX)are among the most important factors affecting the operational cap-ability of marine nuclear power platform PRHR systems.This study aims to analyze the heat transfer charac-teristics of steam condensation in a PRHR HX and obtain the accurate heat transfer coefficient relationship.[Methods]The heat transfer characteristics of steam condensation in a PRHR HX are analyzed by building an experimental device with a power ratio of 1∶50.[Results]The experimental results show that the heat transfer coefficient increases with the increase in steam pressure.The internal flow pattern of the PRHR HX is a stratified flow or wave flow-annular flow-wave flow.The maximum flow rate of HX vapor is 6.72 m/s when the system pressure is 0.52 MPa,and when the pressure exceeds 0.52 MPa,the steam velocity decreases with the increase in heating power.The proposed new correlation shows good agreement,with an error of±8%between the calculated and experimental results.[Conclusions]This study can provide useful references for the design of reactor safety systems for marine nuclear power platforms.