超临界水堆子通道分析及核热耦合分析研究综述  

Review of Sub-channel Study and Coupled Neutronics/Thermal-hydraulic Study for Supercritical Water-Cooled Reactor

作　　者：赵冬建[1] 廖承奎[1] 史国宝[1] 

机构地区：[1]上海核工程研究设计院,200233

出　　处：《核电工程与技术》2008年第3期34-41,共8页Nuclear Power Engineering and Technology

摘　　要：超临界水堆核能系统作为国际上第四代核能系统长远开发的6种堆型之一，具有热效率高，系统简化，经济竞争能力强等优点。本文综述了国内外热中子谱超临界水堆子通道程序的开发及应用、超临界水堆核热耦合分析的相关研究成果，简要介绍了堆内冷却剂和慢化剂的温度、密度、焓、包壳瘟度沿流动方向的变化规律，以及不同子通道冷却剂质量流率的分布规律，探讨了当前研究尚存在的不足，并为下一步研究工作的开展提出了一些建议。The SCWR （Supercritical Water-cooled Reactor） is one of six development under the GIF （Generation Ⅳ International Forum）. The main concept design for further advantage of the SCWR is significant improvements in economy due to increased thermal efficiency and plant system simplification. The purpose of this paper is to review the development and application of sub-channel analysis and coupled neutronics/thermal-hydraulic analysis for the thermal neutron SCWR. The axial and radial distribution characteristics of the temperatures, densities, flow fluxes and enthalpies of the moderator and coolant are briefly summarized. Some shortage of current investigation is discussed and some suggestion is also proposed for the further research.

关 键 词：超临界水堆 子通道分析 核热耦合分析 

分 类 号：TL433.052[核科学技术—核技术及应用] V525[航空宇航科学与技术—人机与环境工程]