A method for quantification of nuclear reactor seismic risk due to design and construction errors based on deficiencies identified in the past

This paper proposes a plausible methodology for quantification of risk due to seismic related design and construction errors in nuclear power plants. This is based on information available on errors discovered in the past, as reported in construction deficiency reports pursuant to US NRC regulations. Deficiencies found during construction whose recurrence is considered possible are included. The possibility of deficiencies not being discovered is evaluated by comparison of data between operating plants and those under construction. Error consequences are also evaluated from construction deficiency reports; in particular, the impact of errors on the seismic capacity of the plants is quantified by an extension of seismic probabilistic risk assessment (PRA) methods.The application of the method is illustrated based on a limited review of data, showing its capabilities and limitations. Some tentative results are presented, but these by no means represent a formal assessment.     

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.     

Recent extensions to seismic margin review methodology

Two seismic margin review methodologies — one by USNRC and the other by EPRI — have been developed in the last four years. The focus is on assessing the capability of existing nuclear power plants to withstand earthquakes larger than the design basis earthquakes. The methods restrict the analysis to a selected few systems and components using the insights from past seismic PRAs, seismic analysis and qualification results, and earthquake experience data. The objective of this paper is to describe recent and on-going studies in extending the NRC seismic margin review methodology. Specifically, three topics are discussed: (1) extension of the HCLPF capacity to analyse radiological releases and importance of human factors and non-seismic failures; (2) importance of BWR plant systems and functions to seismic margins; and (3) extensions of seismic margin review results to obtain seismic risk estimates.     

福清核电厂1、2号机组抗震裕量评价

地震是核电厂主要外部灾害之一,地震风险评估对于核电厂的安全评价具有重要的价值。抗震裕量评价(SMA)是开展核电厂地震灾害风险分析的重要方法之一,其目的是为了判断核电厂的抗震设计能力相对于设计基准地震的抗震裕量,找出核电厂的抗震薄弱环节,提高核电厂的抗震能力。本文针对福建福清核电厂1、2号机组进行抗震裕量评价,分析表明电力支持系统和一回路辅助管道的抗震能力相对薄弱,是导致核电厂抗震能力薄弱的主要原因,电力支持系统和一回路辅助管道需进一步提高其抗震能力,且核电厂需考虑编制地震应急规程。     

Seismic fragilities for nuclear power plant risk studies

Seismic fragilities of critical structures and equipment are developed as families of conditional failure frequency curves plotted against peak ground acceleration. The procedure is based on available data combined with judicious extrapolation of design information on plant structures and equipment. Representative values of fragility parameters for typical modern nuclear power plants are provided. Based on the fragility evaluation for about a dozen nuclear power plants, it is proposed that unnecessary conservatism existing in current seismic design practice could be removed by properly accounting for inelastic energy absorption capabilities of structures. The paper discusses the key contributors to seismic risk and the significance of possible correlation between component failures and potential design and construction errors.     

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.     

Probabilistic cost-benefit seismic design criterion for a dedicated shutdown heat removal system

A probabilistic methodology is developed for assessing the risk reduction potential and cost-benefit tradeoff of a Dedicated Shutdown Heat Removal System (DSHRS) for a PWR as a function of its seismic design capability. The option of coping with a very small LOCA is included. The annual seismic risk of a plant and a similar hypothetical plant having a proposed DSHRS with various seismic strengths are computed. The difference in the annual seismic risks is the annual seismic risk reduction benefit for having the system. The present value of the future risk reduction benefit is then compared to the cost of building a DSHRS and the incremental seismic cost associated with building the system to withstand a stronger earthquake.A reactor like Zion was used for application of the method due to the availability of data. Studies were performed to investigate the sensitivity of the results to the assumed seismic hazard, probability of occurrence of seismic-induced accident initiating events, equipment seismic fragility, accident costs, and discount rate. The incremental seismic risk reduction benefit computed in these studies ranges from $207 million for a DSHRS with a median seismic capacity of 1.70g (i.e. 10 × SSE) in a new plant built at the site. The total cost of a DSHRS is crudely estimated to be $25 million or more, if it were built to withstand the current SSE of the plant (for which the system probably would have a median seismic capacity of 0.85g or more due to various design and construction conservatisms). The cost associated with the seismic design aspect of such a system is estimated to be approximately $2.5 million and it may be doubled if the seismic design capability of the system is tripled. The cost/benefit results and their inherent large uncertainties are not definitive but indicate that probabilistic seismic design of a DSHRS should be examined in further detail.     

弯头地震非线性响应及抗震设计规范保守性研究

通过对比研究1995版和2002版的ASME规范中管道抗震评价部分的相关内容,深入探讨了ASME规范在降低其保守性所做的努力。定义规范保守性因子,以便定量研究规范保守性,同时开展弯头在地震载荷作用下的线性与非线性响应研究,深入探讨了材料非线性、几何大变形和内压对弯头的地震响应的影响,并基于弯头的地震响应计算结果,分析了ASME规范中管道抗震评价部分的保守性。计算结果表明,ASME规范对管道系统抗震评价的保守性较大。     

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.     

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。     

Seismic instrumentation for nuclear power plants: An interpretative review of current practice and the related standard in Germany

The German nuclear safety standard KTA 2201: “Design of nuclear power plants against seismic events”, consists of the following parts: 1. basic principles; 2. characteristics of seismic excitation; 3. design of structural components; 4. design of mechanical and electrical parts; 5. seismic instrumentation; and 6. measures subsequent to earthquakes.While Part 1 was published in June 1975, Part 5 was approved by the Nuclear Safety Standards Commission — Kerntechnischer Ausschuss (KTA) — in June 1977. The other parts are still under development. The requirements of the safety standard KTA 2201.5 deal with

1. (a) number of location (number and location of acceleration recording systems for different sites, single-block plants and multi-block plants);

2. (b) characteristics of instruments (readiness and operation of instruments, margin or errors, dynamic and operation characteristics, duration of records, seismic switch);

3. (c) triggering and information (loss of electric power, start of the acceleration recording systems, threshold of acceleration for triggers and seismic switches, optical and acoustic information); and

4. (d) documentation (results of recordings, inspection and tests).

The purpose of this paper is to present some detailed requirements of the safety standard KTA 2201.5, with its philosophy, and compare these with corresponding requirements in the US. It will be shown that with relatively few instruments, which are very reliable in operation and in triggering, an optimum of data may be available after an earthquake.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant.The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load-deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.     

Lower bound earthquake magnitude for probabilistic seismic hazard evaluation

This paper presents the results of a study that develops an engineering and seismological basis for selecting a lower-bound magnitude (LBM) for use in seismic hazard assessment. As part of a seismic hazard analysis the range of earthquake magnitudes that are included in the assessment of the probability of exceedance of ground motion must be defined. The upper-bound magnitude is established by earth science experts based on their interpretation of the maximum size of earthquakes that can be generated by a seismic source. The lower-bound or smallest earthquake that is considered in the analysis must also be specified.The LBM limits the earthquakes that are considered in assessing the probability that specified ground motion levels are exceeded. In the past there has not been a direct consideration of the appropriate LBM value that should be used in a seismic hazard assessment. This study specifically looks at the selection of a LBM for use in seismic hazard analyses that are input to the evaluation/design of nuclear power plants (NPPs). Topics addressed in the evaluation of a LBM are earthquake experience data at heavy industrial facilities, engineering characteristics of ground motions associated with small-magnitude earthquakes, probabilistic seismic risk assessments (seismic PRAs), and seismic margin evaluations. The results of this study and the recommendations concerning a LBM for use in seismic hazard assessments are discussed.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

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.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

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.     

基于有限元模型验证的控制室盘台抗震鉴定

论文以国内某新建核电站控制室盘台抗震鉴定为例,阐述了基于有限元模型验证的盘台抗震鉴定方法。通过样机试验和模型验证分析,将盘台结构设计与有限元分析进行了有机结合,同时在盘台整体有限元分析验证过程中引入了修正因子,保证了盘台的抗震性能,并使其具有一定的安全裕度。     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

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.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant.The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load–deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.