安全壳冷却系统设计方案综述
Summary on Design Solution of the Containment Cooling System
DOI: 10.12677/NST.2017.53015, PDF, HTML, XML, 下载: 1,517  浏览: 4,020 
作者: 代涛, 刘斯亮, 刘建阁:武汉第二船舶设计研究所,湖北 武汉
关键词: 安全壳冷却系统设计能动非能动Containment Cooling System Design Active Passive
摘要: 安全壳是核动力厂为防止放射性裂变产物释放的最后一道安全屏障,其主要功能是在事故工况期间和以后,限制放射性物质从堆芯和反应堆冷却剂系统释放到周围环境,此外,在运行工况和事故工况期间提供屏蔽,将运行工况下放射性物质的释放降至最低限度,并保护反应堆不受到外部事件的损害。为此,安全壳应设置执行功能的子系统和设施,主要有:安全壳构筑物及其延伸部分、安全壳隔离系统、能量控制设施(限制作用在安全壳内的压力、温度、机械载荷上)、放射性核素控制设施、可燃气体控制设施。其中,安全壳冷却系统被设计用来限制安全壳内的压力和温度,同时兼具放射性核素去除功能。本文调研分析了国内外核动力厂安全壳冷却系统设计方案和趋势,提出了安全壳冷却系统的设计方向:采用能动和非能动相结合的设计思潮,从而为新堆或小型堆的安全壳冷却系统方案设计提供参考价值。
Abstract: Containment is the last safety barrier for nuclear power plant in order to prevent radioactive fission products, its main function is to limit the release of radioactive material from the reactor core and the reactor coolant system to the surrounding environment in the accident conditions. In addition, during the operation and accident conditions, it provides shields, reduces the release of radioactive material to a minimum, and protects the reactor from the damage of external events. Therefore, executive function of the subsystem and facilities shall be set up in the containment, mainly includ-ing: containment structure and its extension, the containment isolation system, the energy control facilities (limited role in the containment vessel pressure, temperature, and mechanical load), ra-dioactive nuclide control facilities, combustible gas control facilities. Among them, the containment cooling system is designed to limit the pressure and temperature inside the containment; at the same time, it has the function of radionuclide removal. This research considered the design scheme of containment cooling system in the nuclear power plant and put forward the design direction of containment cooling system by adopting the combination of active and non active design to provide the reference value for the design of new heap or small reactor containment cooling system.
文章引用:代涛, 刘斯亮, 刘建阁. 安全壳冷却系统设计方案综述[J]. 核科学与技术, 2017, 5(3): 109-115. https://doi.org/10.12677/NST.2017.53015

参考文献

[1] 900 MW压水堆核电站系统与设备(上册)[M]. 北京: 原子能出版社.
[2] 林诚格. 非能动安全先进压水堆核电技术[M]. 北京: 原子能出版社, 2010.
[3] Iwamura, T., Murao, Y., Araya, F. and Okumura, K. (1995) A Concept and Safety Characteristics of JAERI Passive Safety Reactor (JPSR). Progress in Nuclear Energy, 29, 397-404.
https://doi.org/10.1016/0149-1970(95)00068-U
[4] Dmitriev, S.M., Dobrov, A.A., Legchanov, M.A. and Khrobostov, A.E. (2015) Modeling of Coolant Flow in the Fuel Assembly of the Reactor of a Floating Nuclear Power Plant Using the Logos CFD Pro-gram. Journal of Engineering Physics and Thermophysics, 88, 1297-1303.
https://doi.org/10.1007/s10891-015-1312-5
[5] 蒋孝蔚. 抑压式安全壳的抑压特性研究[J]. 核动力工程, 2014, 35(2): 114-117.
[6] 顾军扬, 陈连发. 先进型沸水堆核电厂[M]. 北京: 中国电力出版社, 2007.
[7] 郭志锋. AP1000的非能动安全系统[J]. 核电厂核反应堆, 2005(9): 14-20.
[8] 谭效时. 核电厂新型预应力混凝土安全壳及其非能动冷却系统设计与分析[J]. 原子能科学技术, 2014, 48(2): 271-276.