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论述了核电厂地震概率安全评价(PSA)定量化方法和工具的现状,指出了定量化工具面临的挑战和存在的问题。根据定量化的概率论本质,提出了计算方法。以我国某核电厂厂址多方案概率地震危险性分析(PSHA)结果和核电厂地震响应分析给出的最小割集为例,展示了计算方法的应用过程,分析了地震动参数和置信度参数对定量化计算结果的影响。结果表明,针对置信度参数进行拉丁超立方采样,采样次数较小时即可给出地震导致的核电厂堆芯损坏频率(SCDF)的稳定估计值;通常情况下,设备失效对SCDF的贡献最大,厂房失效的影响相对较小;地震动年发生率对SCDF的贡献需要根据工程场地的位置进行具体分析。 相似文献
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《核动力工程》2017,(6):66-71
先进核电厂设计中大量采用非能动安全系统提高反应堆安全性。但目前尚无系统性评价非能动系统的成熟方法,而且概率安全评价(PSA)也未考虑非能动系统自然循环现象不确定性导致的功能失效。在欧盟非能动系统可靠性评价研究项目(RMPS)研究成果的基础上,以压水堆二次侧非能动余热排出系统(PRS)为研究对象,基于统计学和热工水力计算确定了影响性能的参数重要度,进而利用蒙特卡罗抽样和响应面分析对全厂断电事故下的PRS自然循环失效概率进行了量化分析评价。初步评价结果表明:非能动系统功能失效概率为2.14×10-3,在PSA中应当充分考虑各种非能动系统的功能失效。本文的评价方法还可以为非能动安全系统设计优化提供支持。 相似文献
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地震导致丧失厂外电是核电厂地震情况下的典型始发事件。本研究使用地震概率安全分析方法,以高温气冷堆为研究对象,得到其在地震丧失厂外电事故下的风险水平。研究范围包括分析地震导致丧失厂外电的事故发展情景分析,筛选地震设备清单并结合现场巡访进行调整,建立地震导致丧失厂外电的风险评价模型,并对超过高温气冷堆风险接受准则剂量(概率安全目标)的放射性释放的频率结果进行了间隔分析、割集分析和重要度分析。本文工作可为高温气冷堆的地震概率安全分析在方法实施、建模假设、过程分析等方面提供有益的参考。 相似文献
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随着福岛事故的发生,核电厂外部事件概率安全评价工作的重要性被各国核安全当局所认同。而地震,作为核电厂最为主要的外部事件,其对应的概率安全评价工作便更为人们所重视。易损度计算是完成地震概率安全评价的关键技术环节,其结果将被使用作概率安全评价事故序列模型的输入条件。因此,易损度计算的准确性和正确性对地震概率安全评价工作最终结论的影响也就不言而喻了。本文首先总体性介绍了设备易损度计算的基础数学模型,随后详细描述了核电厂地震概率安全评价中电气设备易损度计算的操作步骤,并重点探讨了电气设备功能失效模式下对试验反应谱和要求反应谱的处理简化技巧,最后通过具体算例阐述了电气设备易损度计算过程中的注意事项和简化技巧应用。 相似文献
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结合三门和海阳核电厂厂用水系统的实际设计,考虑各运行工况下的运行要求,特别是系统布置相关性的影响,通过概率安全评价(PSA)建模确定了系统可用性情况及其相应的风险结果。基于风险分析的结果,为三门核电厂的设计缺陷提出了合理的改进建议方案,从而提高了该电厂的安全性和经济性。同时结合海阳核电厂的实际情况,分析结果认为基本无改进必要,充分体现了PSA对设计的支持和指导作用。 相似文献
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福岛核事故引发了全球范围内对核电厂地震风险的重新审视。我国是地震多发国家,同时在可以预期的未来多年内是世界上最大的核电建造国,因此应重视核电厂的地震风险。现有核电厂的抗震设计主要是基于确定论设计,难以全面评估核电厂地震风险的大小。核电厂地震概率安全评价是利用概率论方法评估核电厂地震风险的有效方法,对核电厂抗震薄弱环节识别和抗震安全改进具有重要意义。文章全面介绍了压水堆核电厂地震概率安全评价方法的开发流程和技术要素,指出了应在核电厂地震概率安全评价中考虑的重要因素和处理方法,为国内核电厂地震概率安全评价工作提供参考。文章建议尽快完善我国核电厂地震概率安全标准体系建设,指导国内核电厂广泛开展地震概率安全评价工作。 相似文献
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本文探索并研究了一种新的地震易损度算法,基于蒙特卡罗(MC)抽样和最大-最小法计算了单个设备和多个设备组合的最小割集的易损度。最小割集包括3种类型:纯地震失效最小割集、包含非事件的最小割集、地震失效和随机失效混合割集。对于仅包含地震失效的事故序列,可直接采用基于蒙特卡罗抽样和最大 最小法的易损度算法进行计算。涉及地震失效和随机失效混合的事故序列,可采用极限近似方法(MCUB)或其他割集定量化算法进行计算。经对比,基于蒙特卡罗抽样和最大 最小法的地震易损度算法计算结果与理论值一致,为工程应用中的地震易损度计算提供了另一种可行的算法。 相似文献
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Systems analysis is being used in conjunction with structural analysis to study the conservatisms and to provide insights into aspects of reactor seismic safety. An event-tree/fault-tree model of a commercial nuclear power plant is being constructed to determine the probability of release and probabilities of system and component failures caused by possible seismic events. The event-tree/fault-tree model is evaluated using failure data generated by applying the response a component sees to the component's fragility function. The responses are calculated by a structural analysis code using earthquake time histories as forcing functions. The quantification of the event-tree/fault-tree model is done conditional on a given seismic event and the conditional probabilities thus calculated unconditioned by integrating the results over the seismic hazard curve. In this way, most of the dependencies between event failures resulting from the seismic event itself are removed making known fault-tree analysis quentification techniques applicable. The outputs from the computations will be used in sensitivity studies to determine the key calculations and variables involved in seismic analyses of nuclear power plants. 相似文献
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I. Zentner 《Nuclear Engineering and Design》2010,240(6):1614-1621
The seismic probabilistic risk assessment (PRA) methodology is a popular approach for evaluating the risk of failure of engineering structures due to earthquake. In this framework, fragility curves express the conditional probability of failure of a structure or component for a given seismic input motion parameter A, such as peak ground acceleration (PGA) or spectral acceleration. The failure probability due to a seismic event is obtained by convolution of fragility curves with seismic hazard curves. In general, a log-normal model is used in order to estimate fragilities. In nuclear engineering practice, these fragilities are determined using safety factors with respect to design earthquake. This approach allows to determine fragility curves based on design study but largely draws on expert judgement and simplifying assumptions. When a more realistic assessment of seismic fragility is needed, simulation-based statistical estimation of fragility curves is more appropriate. In this paper, we will discuss statistical estimation of parameters of fragility curves and present results obtained for a reactor coolant system of nuclear power plant. We have performed non-linear dynamic response analyses using artificially generated strong motion time histories. Uncertainties due to seismic loads as well as model uncertainties are taken into account and propagated using Monte Carlo simulation. 相似文献
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Toshihiko Hirama Masashi Goto Hitoshi Kumagai Yukio Naito Atsushi Suzuki Hiroshi Abe Katsuki Takiguchi Hiroshi Akiyama 《Nuclear Engineering and Design》2007,237(11):1128-1139
In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses. 相似文献
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Toshihiko Hirama Masashi Goto Minoru Kanechika Tsutomu Mieda Katsuki Takiguchi 《Nuclear Engineering and Design》2005,235(13):1335-1348
In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses. 相似文献
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Toshihiko Hirama Masashi Goto Toshiyasu Hasegawa Minoru Kanechika Takahiro Kei Tsutomu Mieda Hiroshi Abe Katsuki Takiguchi Hiroshi Akiyama 《Nuclear Engineering and Design》2005,235(13):1128
In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses. 相似文献
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基于子集模拟法非能动系统功能故障概率评估 总被引:2,自引:2,他引:0
针对非能动系统多维不确定性参数和小功能故障概率问题,提出基于马尔可夫链蒙特卡罗子集模拟的可靠性分析方法。该方法通过引入适当的中间失效事件,将小功能故障概率表达为一系列较大的中间失效事件条件概率乘积的形式,进而利用马尔可夫链模拟的条件样本点来计算条件失效概率。以AP1000非能动余热排出系统为研究对象,考虑热工水力学模型和输入参数的不确定性,对其进行功能故障概率评估。结果表明:与其它概率评估方法相比,子集模拟法具有较高的计算效率,同时又能保证很高的计算精度;对非能动安全系统非线性功能函数有很强的适应性。 相似文献
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本文采用有限元软件ANSYS建立AP1000核电站堆芯补水箱(CMT)三维有限元模型,通过模态分析获得其结构特征,采用时程分析法较为真实地模拟CMT地震下响应。通过地震易损性数学模型,对CMT的各项易损性参数进行分析,获得了其抗震能力中值Am、随机性标准差βR以及不确定性标准差βU,计算出其高置信度低失效概率(HCLPF)值。结果表明:CMT的HCLPF值明显高于设计安全停堆地震强度0.3g,说明其具有较高的抗震能力,且HCLPF值略高于采用确定论方法得到的值。对易损性参量误差敏感性分析发现βR取值变化对CMT的条件失效概率和HCLPF值影响较小,可简化部分随机性误差的考虑,使得易损性分析更简洁。 相似文献