运行核电厂抗震裕度评价研究

抗震裕度评价是对核电厂应对超过设计基准地震能力的评价,特别是日本福岛核事故发生后,评价核电厂应对超过设计基准外部事件时的安全裕量、优化和落实改进措施、提高改进措施的有效性就显得尤为重要。本文通过研究国际上广泛采用的抗震裕度评价方法,最终选定EPRI SMA方法对秦山第二核电厂进行抗震裕量分析。分析结果表明：秦山第二核电厂满足1.5倍SSE的抗震裕度要求,具有较强的抗震能力。     

高温气冷堆蓄电池组地震易损性研究

为验证核电厂发生地震外部事件时的电力安全，需要对蓄电池组进行抗震鉴定试验。本文以高温气冷堆（HTR）核电厂安全级蓄电池组为研究对象、以安全级蓄电池组抗震鉴定试验数据和工程经验为基础，通过识别、量化蓄电池组的地震易损性变量，并应用基于试验的易损性分析法推导出地震易损性曲线和高置信度低失效概率（HCLPF）抗震能力。研究结果表明，安全级蓄电池组的抗震能力远高于核电厂设计基准地震动需求。      

核动力管道抗震完整性试验及分析研究

对于偶发性地震载荷过分保守的处理导致核电厂管道系统使用大量的阻尼器、支吊架而使管道系统刚性过大,使核电厂的制造、安装、在役检查及维修等费用增加。依托主管道先进设计技术和试验验证项目,完成了核动力管道系统的抗震极限承载能力试验。将试验和计算分析结果与现行规范对比,明确了当前管道设计标准的安全裕量,提出了管道系统阻尼比值与应力评价准则等参数的取值建议。     

压水堆核电厂地震概率安全评价开发方法研究

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。     

核电厂抗震设计取消OBE原因和措施浅析

通过分析对比美国核管理委员会(US NRC)和美国电力研究院(EPRI)出版的文件,结合核一级管道应力比的统计数据,对核电厂抗震设计中取消运行基准地震(OBE)的原因进行分析。结果表明,降低OBE对安全停堆地震(SSE)的比例,并且在抗震分析中不再把OBE作为直接的载荷因素,仅作为停堆检查的限值,能够达到优化核电厂抗震设计的目的。为了防止因此而降低核电厂抗震方面的安全性能,还需制定相关措施。     

我国核电厂抗震设计反应谱和RG 1.60设计反应谱的比较分析

设计反应谱对评价核电厂在地震作用下的安全性极为重要。本文从统计核电厂抗震设计标准反应谱时选取的强震数据及统计方法两个方面,分析比较了美国RG 1.60设计反应谱和我国核电厂抗震设计规范反应谱的异同。通过对比分析,深入理解核电厂抗震设计反应谱的提出需考虑的关键因素,为核电厂抗震设计和审评工作提供参考。     

三环路核电厂的抗震裕度评价

抗震裕度评价是核电厂中开展地震风险评价的方法之一。本论文采用PSA-based SMA方法对三环路核电厂进行了抗震裕度评价工作,并以福清核电厂一期工程为评价对象,通过对SMA各项技术要素,包括抗震裕度地震、地震设备清单、SSC的HCLPF值评价、SMA建模、电厂的HCLPF值评价等详细分析和评价,最终得到福清核电厂一期工程的抗震裕度评价结果。     

先进轻水反应堆业主要求文件（ALWR—URD）中对核电厂抗震设计要求

“先进轻水反应堆业主要求文件(ALWR—URD)”的宗旨是明确美国电力公司对先进轻水堆核电厂的要求。对URD中关于核电厂抗震设计，特别是对核电厂构筑物、系统和设备的抗震分类、取消OBE地震后的抗震设计要求等方面提出的修正意见．以及对核电厂抗震裕度和地震风险评价提出的初步要求，本文作了全面的总结，在关键处浅讨了看法。供新核电厂设计中参考。     

日本柏崎·刈羽核电厂新潟地震震害经验

通过对日本柏崎·刈羽核电厂在2007年7月16日新潟地震中的主要震害情况的描述,分析说明该核电厂的抗震设计基本情况,最后提出核电厂抗震设计与评价中应引起重视的问题。     

蒸汽发生器非线性支承系统的抗震能力分析

计算核电厂设备的高置信度低失效概率(HCLPF)抗震能力是地震概率安全评价、地震裕度评价的一个重要步骤。以蒸汽发生器支承为研究对象,建立其详细的非线性有限单元模型,通过逐步增大地面运动水平,反复计算系统的响应,最后得到蒸汽发生器支承的抗震能力,并与通过确定性失效裕度法得到的HCLPF进行比较。结果表明,两者的计算结果差别较大。本文建议对于非线性较强的设备需采用非线性时程分析方法计算设备的HCLPF。     

Seismic individual plant examination of external events of US nuclear power plants: insights and implications

As part of the implementation of the severe accident policy, nuclear power plants in the US are conducting the individual plant examination of external events (IPEEE). Seismic events are treated in these IPEEEs by either a seismic probabilistic risk assessment (PRA) or a seismic margin assessment. The major elements of a seismic PRA are the seismic hazard analysis, seismic fragility evaluation of structures and equipment and systems analysis using event tree and fault tree analysis techniques to develop accident sequences and calculate their frequencies of occurrence. The seismic margin assessment is a deterministic evaluation of the seismic margin of the plant beyond the design basis earthquake. A review level earthquake is selected and some of the components that are on the success paths are screened out as exceeding the review level earthquake; the remaining ones are evaluated for their seismic capacity using information from the original plant design criteria, test data and plant walkdown. The IPEEEs of over 100 operating nuclear power plants are nearing completion. This paper summarizes the lessons learned in conducting the IPEEEs and their applicability to nuclear power plants outside of the United States.     

地震下非能动堆芯冷却系统可靠性分析

地震情况下核电站非能动堆芯冷却系统（PXS）能否可靠运行对核电站的安全性有着重要影响。本文采用故障树方法分析计算了PXS各部件在峰值地面加速度（PGA）为0.5g、1.5g、2.5g情况下的失效概率以及各部件对系统失效的贡献，并与《AP1000概率安全分析报告》中的抗震裕量分析（SMA）方法的结果进行比较，分析部件的抗震能力。结果表明：本文方法计算的条件失效概率和各部件对系统失效的贡献与SMA方法的结果基本相符。本文方法可为AP1000等非能动核电站的安全分析提供参考。     

核电厂安全壳地震概率风险评估

地震概率风险评估可分别基于地震风险解析函数和风险卷积函数实现。本文推导了地震风险解析函数，分析了地震风险解析函数蕴含的两个基本假设和两个近似，分别基于地震风险解析函数和风险卷积函数计算了我国某核电厂安全壳地震风险。结果表明：采用幂指数函数近似地震危险性极值Ⅱ型分布对风险结果无影响；对于算例厂址，地震风险解析函数中KH和kⅠ为常数的近似会高估核电厂安全壳面临的地震风险；我国核电厂安全壳结构地震风险较低，具有较大安全裕量。建议采用地震风险解析函数初步评估我国核电厂安全壳地震风险。     

Seismic Probabilistic Risk Assessment of Nuclear Power Plant Containment

Seismic probabilistic risk assessment could be respectively conducted using analytical function of seismic risk and risk convolution function. In this paper, analytical function of seismic risk was conducted, two basic assumptions and two approximations of analytical function of seismic risk were analyzed, and seismic probabilistic risk analysis of a nuclear power plant containment of our country were respectively conducted using analytical function of seismic risk and risk convolution function. The results show that there is no influence on seismic risk results using a power exponent function approximating seismic hazard distribution following extreme value Ⅱ type distribution. For the case of this paper, seismic risk of a nuclear power plant containment is overestimated based on analytical function of seismic risk, which uses constant KH and kⅠ. Seismic risk of a containment is low in our country, which has a large safety margin. It is proposed that the preliminary seismic risk assessment of a nuclear power plant containment of our country using analytical function of seismic risk should be conducted.     

Development of the site-specific uniform hazard spectra for Korean nuclear power plant sites

A seismic risk analysis has been performed to evaluate the seismic safety of a nuclear power plant for strong earthquakes beyond a design earthquake level. A site-specific median spectrum has generally been used for a seismic fragility analysis of structures and equipment in Korean nuclear power plants as a part of a probabilistic seismic risk assessment. The site-specific response spectrum, however, does not represent the same probability of an exceedance over entire frequency range of interest. The site-specific uniform hazard spectrum (UHS) is more appropriate for use in a seismic probabilistic risk assessment (SPRA) than the site-specific spectrum, since there are only a few strong motion data and seismological information for the nuclear plant sites in Korea. In this study, the uniform hazard spectra were developed using the available seismic hazard data for four Korean NPP sites.     

PRA-Based SMA: the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS

International Reactor Innovative and Secure (IRIS) is an advanced, modular, medium-power PWR with an integral primary system layout. As part of the “safety-by-design_” philosophy that inspired the project from the very beginning, a risk-informed approach to its design phase is being adopted and a probabilistic risk assessment (PRA) is being used as an active tool in pursuing an advanced level of safety. Within this framework, a preliminary PRA-based seismic margin assessment (SMA) has been conducted to assess the ability of the IRIS standard design to respond to seismic events. A high confidence of low probability of failure at the core damage sequence level and then at the entire plant level is the primary result of the SMA model; in the end, it will have to ensure that IRIS can withstand the review-level earthquake of 0.5 g which is consistent with the upper bin level of the NUREG/CR-4334.1) In this preliminary phase of its development, in which the core of the quantitative data is critically extracted from the SMA of other PWR designs, the IRIS SMA model can be seen as a first step toward the development of an extensive seismic PRA model.     

Seismic probabilistic risk assessment and its impact on margin studies

In recent years a number of seismic probabilistic risk assessments of nuclear power plants have been conducted. These studies have highlighted the significance of seismic events to the overall plant risk and have identified several dominant contributors to the seismic risk. It has been learnt from the seismic PRAs that the uncertainty in the seismic hazard results contribute to the large uncertainty in the core damage and severe release frequencies. A procedure is needed to assess the seismic safety of a plant which is somewhat removed from the influence of the uncertainties in seismic hazard estimates. In the last two years, seismic margin review methodologies have been developed based on the results and insights from the seismic probabilistic risk assessments. They focus on the question of how much larger an earthquake should be beyond the plant design basis before it compromises the safety of the plant. An indicator of the plant seismic capacity called the High Confidence Low Probability of Failure (HCLPF) capacity, is defined as the level of earthquake for which one could state with high confidence that the plant will have a low probability of severe core damage. The seismic margin review methodologies draw from the seismic PRAs, experience in seismic analyses, testing and actual earthquakes in order to minimize the review effort. The salient steps in the review consists of preliminary screening of components and systems, performance of detailed seismic walkdowns and evaluation of seismic margins for components, systems and plant.     

多种影响因素下的设备易损性计算

抗震裕度评估是核电厂地震安全评估的方法之一，通过地震易损性分析计算高置信度低失效概率的抗震能力值是抗震裕度评估中很重要的一步。本文对于同时受到多种失效模式影响的设备易损性计算进行了研究，讨论了蒙特卡罗抽样方法和拉丁超立方分布抽样方法在设备易损性计算中的应用，对两种抽样方法的计算效率和准确度进行了评价。结果表明，在小样本抽样计算时拉丁超立方抽样方法有更好的计算效率和收敛速度，在1 000次样本数量时，两种抽样方法计算结果均可达到收敛。     

CDFM方法在秦山核电厂SMA中的应用

抗震裕度评价（SMA）是核电厂地震安全评价的方法之一,而计算构筑物、系统和部件（SSC）的高置信度低概率失效（HCLPF）值是开展抗震裕度评价的重要内容之一。本文介绍了HCLPF值的定义和计算HCLPF值的保守的确定论失效裕度（CDFM）方法。结合秦山核电厂抗震裕度评价,以应急柴油发电机组和主控制楼为例说明了CDFM方法在核电厂SMA中的应用。通过计算得到了大部分SSC的HCLPF值,为秦山核电厂SMA工作的顺利开展奠定了基础。