压水堆不同尺寸的破口失水事故分析
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摘要
基于一体化严重事故分析程序MAAP4.0.3(Modular Accident Analysis Program),本文建立了我国现役典型百万千瓦级压水堆(Pressurized-Water Reactor,PWR)核电机组模型,研究了热管段不同面积破口事故叠加安注失效的工况引起的严重事故过程,探讨了如何在恰当的时机采取有效的缓解措施对事故的进程进行干预。研究结果表明:在破口事故中随着破口面积而增大,压力容器会更早失效导致堆芯裸露;一旦压力容器失效,MCCI(Molten Corium Concrete Interaction)过程中氢气产量则会随着破口面积的增大而增大;在破口事故中尽早投入安全注射系统可以有效地缓解事故的进程,避免压力容器失效,并且安全注射系统越早投入对事故的缓解也就越有利。
[Background] Serious accidents refer to a series of processes in which the nuclear reactor core cooling loss and fuel widespread failure threaten or damage the integrity of the pressure vessel or containment of a nuclear power plant and may cause leakage of radioactive material. [Purpose] This study aims to analyze the influence of break size on the loss of coolant accidents(LOCA) in pressurized-water reactor(PWR). [Methods] A typical millionkilowatt PWR unit model in active serving nuclear power plant(NPP) in China was built based on MAAP4.0.3(Modular Accident Analysis Program). Comparison with different crevasse size LOCA were performed by MAAP code to analyze the accident process. Besides, LOCA with different time to inject cooling water was also simulated to evaluate the mitigation measures. [Results] Results show that in the process of LOCA without safety injection system, under the same other conditions, the larger the crack is, the faster the pressure vessel will fail and the more hydrogen will be produced. During the serious accidents, high head safe injection has little effect on the primary loop pressure. And input safety injection system manually or automatically can protect the integrity of the reactor vessel and prevent producing hydrogen. [Conclusion] The LOCA case with large crevasse can cause more severe rector core damaged. Meanwhile, it is important to inject water into the reactor core at an appropriate time to prevent the severe accident process effectively.
[Background] Serious accidents refer to a series of processes in which the nuclear reactor core cooling loss and fuel widespread failure threaten or damage the integrity of the pressure vessel or containment of a nuclear power plant and may cause leakage of radioactive material. [Purpose] This study aims to analyze the influence of break size on the loss of coolant accidents(LOCA) in pressurized-water reactor(PWR). [Methods] A typical millionkilowatt PWR unit model in active serving nuclear power plant(NPP) in China was built based on MAAP4.0.3(Modular Accident Analysis Program). Comparison with different crevasse size LOCA were performed by MAAP code to analyze the accident process. Besides, LOCA with different time to inject cooling water was also simulated to evaluate the mitigation measures. [Results] Results show that in the process of LOCA without safety injection system, under the same other conditions, the larger the crack is, the faster the pressure vessel will fail and the more hydrogen will be produced. During the serious accidents, high head safe injection has little effect on the primary loop pressure. And input safety injection system manually or automatically can protect the integrity of the reactor vessel and prevent producing hydrogen. [Conclusion] The LOCA case with large crevasse can cause more severe rector core damaged. Meanwhile, it is important to inject water into the reactor core at an appropriate time to prevent the severe accident process effectively.
引文
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8余建辉,张经瑜,郑利民.大破口LOCA事故ASTRUM最佳估算分析方法优化研究[J].核技术,2014,37(9):090606.DOI:10.11889/j.0253-3219.2014.hjs.37.090606.YU Jianhui,ZHANG Jingyu,ZHENG Limin.Study on ASTRUM optimization for large break LOCA analysis[J].Nuclear Techniques,2014,37(9):090606.DOI:10.11889/j.0253-3219.2014.hjs.37.090606.
9 Peter B,Stefan H.Comparative risk assessment of severe accidents in the energy sector[J].Energy Policy,2014,74:S45-S56.DOI:10.1016/j.enpol.2014.01.035.
10 Wang T C,Wang S J,Teng J T.Comparison of severe accident results among SCDAP/RELAP5,MAAP,and MELCOR codes[J].Nuclear Technology,2005,150(2):145-152.
11博金海,王飞.小破口失水事故研究综述[J].核科学与工程,1998,(2):172-179.BO Jinhai,WANG Fei.Review on research of small break loss of coolant accident[J].Chinese Journal of Nuclear Science and Engineering,1998,(2):172-179.
12胡海平,刘全友,王盟,等.基于MAAP5程序的秦山核电站严重事故分析[J].原子能科学技术,2018,52(4):641-645.DOI:10.7538/yzk.2017.youxian.0438.HU Haiping,LIU Quanyou,WANG Meng,et al.Severe accident analysis of Qinshan nuclear power plant based on MAAP5 code[J].Atomic Energy Science and Technology,2018,52(4):641-645.DOI:10.7538/yzk.2017.youxian.0438.
13黄雄,吕雪峰,陈彦霖,等.AP1000核电厂蒸汽发生器主管道发生小破口事故情况下氢气源项分析[J].核安全,2015,14(1):60-64.DOI:10.3969/j.issn.1672-5360.2015.01.010.HUANG Xiong,LYU Xuefeng,CHEN Yanlin,et al.The analysis of hydrogen source term under SB-LOCA in AP1000 nuclear power plant[J].Nuclear Safety,2015,14(1):60-64.DOI:10.3969/j.issn.1672-5360.2015.01.010.
14胡啸,黄挺,裴杰,等.百万千瓦级压水堆严重事故后再注水的有效性评价[J].原子能科学技术,2015,49(11):2069-2075.DOI:10.7538/yzk.2015.49.11.2069.HU Xiao,HUANG Ting,PEI Jie,et al.Water reflooding effectiveness assessment for 1 000 MWe PWR under severe accident condition[J].Atomic Energy Science and Technology,2015,49(11):2069-2075.DOI:10.7538/yzk.2015.49.11.2069.
2舒睿,许川.秦山核电二期工程严重事故研究[J].核动力工程,2003,(S1):36-39.DOI:10.3969/j.issn.0258-926.2003.z1.010.SHU Rui,XU Chuan.Study on serious accident in Qinshan nuclear power plantⅡproject[J].Nuclear Power Engineering,2003,(S1):36-39.DOI:10.3969/j.issn.0258-0926.2003.z1.010.
3丁亮.核电厂严重事故实时仿真两相流模型研究[D].哈尔滨:哈尔滨工程大学,2006.DING Liang.Study on real-time simulation two-phase flow model of severe accident in NPP[D].Harbin:Harbin Engineering University,2006.
4 Avramova M N,Ivanov K N.Verification,validation and uncertainty quantification in multi-physics modeling for nuclear reactor design and safety analysis[J].Progress in Nuclear Energy,2010,52(7):601-614.DOI:10.1016/j.pnucene.2010.03.009.
5 NRC.Severe accident risks:an assessment for five USnuclear power plants[S].Washington D C:NRC,NUREG-1150,1990.
6武小莉,李伟,邓坚,等.一体化严重事故程序与系统程序耦合研究[J].核动力工程,2018,39(2):157-161.DOI:10.13832/j.jnpe.2018.02.0157.WU Xiaoli,LI Wei,DENG Jian,et al.Study on coupling of integral severe accident analysis code with system code[J].Nuclear Power Engineering,2018,39(2):157-161.DOI:10.13832/j.jnpe.2018.02.0157.
7陈玉清,赵新文,杨磊.小型压水堆严重事故序列的筛选及模拟分析研究[J].核科学与工程,2016,36(3):335-340.DOI:10.3969/j.issn.0258-0918.2016.03.006.CHEN Yuqing,ZHAO Xinwen,YANG Lei.Filtration and imitate analysis of minitype PWR severe accident sequence[J].Nuclear Science and Engineering,2016,36(3):335-340.DOI:10.3969/j.issn.0258-0918.2016.03.006.
8余建辉,张经瑜,郑利民.大破口LOCA事故ASTRUM最佳估算分析方法优化研究[J].核技术,2014,37(9):090606.DOI:10.11889/j.0253-3219.2014.hjs.37.090606.YU Jianhui,ZHANG Jingyu,ZHENG Limin.Study on ASTRUM optimization for large break LOCA analysis[J].Nuclear Techniques,2014,37(9):090606.DOI:10.11889/j.0253-3219.2014.hjs.37.090606.
9 Peter B,Stefan H.Comparative risk assessment of severe accidents in the energy sector[J].Energy Policy,2014,74:S45-S56.DOI:10.1016/j.enpol.2014.01.035.
10 Wang T C,Wang S J,Teng J T.Comparison of severe accident results among SCDAP/RELAP5,MAAP,and MELCOR codes[J].Nuclear Technology,2005,150(2):145-152.
11博金海,王飞.小破口失水事故研究综述[J].核科学与工程,1998,(2):172-179.BO Jinhai,WANG Fei.Review on research of small break loss of coolant accident[J].Chinese Journal of Nuclear Science and Engineering,1998,(2):172-179.
12胡海平,刘全友,王盟,等.基于MAAP5程序的秦山核电站严重事故分析[J].原子能科学技术,2018,52(4):641-645.DOI:10.7538/yzk.2017.youxian.0438.HU Haiping,LIU Quanyou,WANG Meng,et al.Severe accident analysis of Qinshan nuclear power plant based on MAAP5 code[J].Atomic Energy Science and Technology,2018,52(4):641-645.DOI:10.7538/yzk.2017.youxian.0438.
13黄雄,吕雪峰,陈彦霖,等.AP1000核电厂蒸汽发生器主管道发生小破口事故情况下氢气源项分析[J].核安全,2015,14(1):60-64.DOI:10.3969/j.issn.1672-5360.2015.01.010.HUANG Xiong,LYU Xuefeng,CHEN Yanlin,et al.The analysis of hydrogen source term under SB-LOCA in AP1000 nuclear power plant[J].Nuclear Safety,2015,14(1):60-64.DOI:10.3969/j.issn.1672-5360.2015.01.010.
14胡啸,黄挺,裴杰,等.百万千瓦级压水堆严重事故后再注水的有效性评价[J].原子能科学技术,2015,49(11):2069-2075.DOI:10.7538/yzk.2015.49.11.2069.HU Xiao,HUANG Ting,PEI Jie,et al.Water reflooding effectiveness assessment for 1 000 MWe PWR under severe accident condition[J].Atomic Energy Science and Technology,2015,49(11):2069-2075.DOI:10.7538/yzk.2015.49.11.2069.