基于堆芯模型耦合过渡方法的严重事故模拟机开发研究
|
摘要
采取系统分析程序耦合过渡一体化严重事故(SA)分析程序的方法,对严重事故模拟机的开发进行研究。该方法首先使用系统分析程序计算事故早期响应,当满足耦合条件时,系统程序停止计算,切换至严重事故程序计算模拟事故中晚期。为实现切换时参数平滑过渡,以全范围模拟机常用程序RELAP5和严重事故程序MAAP4为例,主要分析了两程序热工水力模型重叠部分的堆芯区域的物理模型,选择传递了堆芯节点的芯块温度、包壳温度和堆芯功率。基于通用百万千瓦级压水堆小破口失水事故(SBLOCA)模型,使用该方法计算和SA程序单独计算进行对比验证。结果表明,过渡参数的选取是正确的,该系统分析程序耦合过渡SA程序的方法不仅能成功平滑地过渡参数,还保证了后续计算的准确性。
This paper presented a codes coupling method of system analysis code and integrated severe accident(SA) code to develop a new SA simulator. It uses system analysis code to calculate early stage of an accident, when satisfying coupling condition, system analysis code stops calculation and switches to SA code to simulate mid-late-term accident. In order to achieve smooth transition of parameters, this paper took full scope simulator general code RELAP5 and SA code MAAP4 as example to analyze their repetitive thermal-hydraulic models, especially the core model. Core fuel temperature, cladding temperature and core power were selected as transition parameters. Based on a same small break loss of coolant accident(SBLOCA) of generic million kilowatts of Pressurized Water Reactor(PWR), coupling transition calculation and SA code independent calculation were compared to test the method. The simulation results showed that this selection of transition parameters is correct. This method can not only smoothly transmit the parameters but also ensure accuracy of subsequent calculation.
This paper presented a codes coupling method of system analysis code and integrated severe accident(SA) code to develop a new SA simulator. It uses system analysis code to calculate early stage of an accident, when satisfying coupling condition, system analysis code stops calculation and switches to SA code to simulate mid-late-term accident. In order to achieve smooth transition of parameters, this paper took full scope simulator general code RELAP5 and SA code MAAP4 as example to analyze their repetitive thermal-hydraulic models, especially the core model. Core fuel temperature, cladding temperature and core power were selected as transition parameters. Based on a same small break loss of coolant accident(SBLOCA) of generic million kilowatts of Pressurized Water Reactor(PWR), coupling transition calculation and SA code independent calculation were compared to test the method. The simulation results showed that this selection of transition parameters is correct. This method can not only smoothly transmit the parameters but also ensure accuracy of subsequent calculation.
引文
[1]陈达.核能与核安全:日本福岛核事故分析与思考[J].南京航空航天大学学报,2012,44(05):597-602.
[2]环境保护局(国家核安全局),国家发展改革委,财政部,等.核安全与放射性污染防治“十二五规划及2020年远景目标[R].北京:环境保护局(国家核安全局),2012.
[3]国家核安全局.核电厂运行安全规定:HAF103[S].北京:中国法制出版社,2004,3:11-18.
[4]赵欣,刘东,王加昌,等.基于SAMG事故序列的严重事故模拟软件开发[J].核动力工程,2013,34(S1):210-213.
[5]郎明刚,高祖瑛.严重事故分析程序[J].核动力工程,2002,32(02):46-50.
[6]靖剑平,张春明,陈妍,等.浅谈核电领域中的热工水力分析程序[J].核安全,2012,(03):70-74.
[7]Fauske&Associates,LLC(FAI).MAAP4.0.7 user manual[M].USA:Electric Power Research Institute,1994.
[8]The RELAP5 Code Development Team.RELAP5/MOD3code manual[Z].USA:Idaho National Engineering Laboratory,1995.
[9]孙吉良,肖岷,黄辉章,等.大亚湾核电站严重事故管理导则[J].核动力工程,2003,24(S2):5-9.
[2]环境保护局(国家核安全局),国家发展改革委,财政部,等.核安全与放射性污染防治“十二五规划及2020年远景目标[R].北京:环境保护局(国家核安全局),2012.
[3]国家核安全局.核电厂运行安全规定:HAF103[S].北京:中国法制出版社,2004,3:11-18.
[4]赵欣,刘东,王加昌,等.基于SAMG事故序列的严重事故模拟软件开发[J].核动力工程,2013,34(S1):210-213.
[5]郎明刚,高祖瑛.严重事故分析程序[J].核动力工程,2002,32(02):46-50.
[6]靖剑平,张春明,陈妍,等.浅谈核电领域中的热工水力分析程序[J].核安全,2012,(03):70-74.
[7]Fauske&Associates,LLC(FAI).MAAP4.0.7 user manual[M].USA:Electric Power Research Institute,1994.
[8]The RELAP5 Code Development Team.RELAP5/MOD3code manual[Z].USA:Idaho National Engineering Laboratory,1995.
[9]孙吉良,肖岷,黄辉章,等.大亚湾核电站严重事故管理导则[J].核动力工程,2003,24(S2):5-9.