核电厂SPRHR冷凝器传热特性研究  

Study on Heat Exchange Experiment Characteristics of SPRHR Condenser for Nuclear Power Plants

作　　者：杨锦春 

机构地区：[1]深圳中广核工程设计有限公司,广东深圳518026

出　　处：《核动力工程》2018年第1期61-63,共3页Nuclear Power Engineering

摘　　要：为保障三代核电厂在严重事故下能通过非能动系统排出堆芯余热的安全要求,对该系统中的二次侧非能动余热排出(SPRHR)冷凝器在不同蒸汽压力和流量下的传热特性进行了试验研究。研究结果表明:SPRHR冷凝器具有足够的安全余量,传热性能稳定;Shah和Foster-Zuber关系式可以很好地预测SPRHR冷凝器传热性能。In order to meet the safety requirement of nuclear power plants, the residual heat must be removed by the passive system under severe accidents. An experiment is conducted on the heat exchange characteristics of Secondary Passive Residual Heat Removal Condenser （SPRHR condenser） under different vapor pressure and temperature conditions. The results show that the heat exchange characteristics of SPRHR is stable and with enough safe margin. The heat exchange characteristics of SPRHR can be predicted well by Shah and Foster-Zuber correlations.

关 键 词：SPRHR冷凝器 冷凝换热 池式沸腾 

分 类 号：TL33[核科学技术—核技术及应用]