基于加速度效应的超临界水传热特性模型研究  

Study on Heat Transfer Model of Supercitical Water Based on Acceleration Effect

作　　者：曾小康[1] 李永亮[1] 闫晓[1] 黄志刚[1] 黄彦平[1] Zeng Xiaokang;Li Yongliang;Yan Xiao;Huang Zhigang;Huang Yanping(CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology,Nuclear Power Institute of China,Chengdu,610213,China)

机构地区：[1]中国核动力研究设计院中核核反应堆热工水力技术重点实验室

出　　处：《核动力工程》2018年第6期29-33,共5页Nuclear Power Engineering

摘　　要：超临界水冷堆(SCWR)运行在水的热力学临界点(22.1 MPa,374℃)之上,堆内冷却剂处于超临界状态,物性变化剧烈,与常规压水堆临界热流密度(CHF)导致包壳表面壁温飞升不同,超临界压力下的传热恶化是在变物性的影响下使得包壳表面温度相对缓慢上升,传统的热点判定方法和偏离泡核沸腾比(DNBR)限值等传热特性分析方法不再完全适用,因此,预测超临界水传热恶化时包壳壁温对SCWR的安全分析相当重要。本文基于边界层方程推导了超临界水传热关系式的加速度效应修正项,基于圆管实验数据,对加速度效应修正项的相关系数进行拟合获得超临界水传热特性半经验关系式,通过数据对比,该关系式在正常传热和传热恶化工况下均具有较好的适用性。本文获得的超临界水传热特性半经验关系式可为SCWR堆芯设计分析提供支持。Supercritical-water cooled reactor is above the thermal critical point(22.1 MPa,374℃). The coolant is supercritical water which physical property changes sharply with the change of temperature. The characteristic of heat transfer deterioration of supercritical water is that the wall temperature increases slowly via heat power, which is different from CHF of pressured-water reactors. So, forecasting the cladding wall temperature of the fuel rod for heat transfer deterioration is very important for the safety analysis of SCWR. The paper infers the modified term of acceleration effect based on the boundary layer equations. The paper achieves the fitting of the correlative factor of the modified term by the test data and the semi-empirical relationship is gained. By comparing the test data, the relationship is capable to forecast the wall temperature for normal heat transfer and heat transfer deterioration.

关 键 词：超临界水冷堆(SCWR) 传热特性模型 加速度效应 关系式 

分 类 号：TL33[核科学技术—核技术及应用]