An update of preliminary perspectives gained from individual plant examination of external events (IPEEE) submittal reviews

As a result of the US Nuclear Regulatory Commission (USNRC) initiated Individual Plant Examination of External Events (IPEEE) program, virtually every operating commercial nuclear power reactor in the United States has performed an assessment of severe accident risk due to external events. The USNRC staff has received all 70 IPEEE submittals. This paper is based on the information available for those 41 plants for which at least preliminary Technical Evaluation Reports have been prepared by the review teams. The goal of the review is to ascertain whether the licensee’s IPEEE process is capable of identifying external events-induced severe accident vulnerabilities and cost-effective safety improvements to either eliminate or reduce the impact of these vulnerabilities. The review does not, however, attempt to validate or verify the results of the licensee’s IPEEE. The primary objective of this paper is to provide an update on the preliminary perspectives and insights gained from the IPEEE process. For most licensees the principal objectives of the IPEEE program have been met, and the program has had some impact on improving plant safety.     

Preliminary perspectives gained from individual plant examination of external events (IPEEE) seismic and fire submittal review

As a result of the US Nuclear Regulatory Commission (USNRC) initiated Individual Plant Examination of External Events (IPEEE) program, every operating nuclear power reactor in the United States has performed an assessment of severe accident due to external events. This paper provides a summary of the preliminary insights gained through the review of 24 IPEEE submittals.     

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.     

Methodology and results of the seismic probabilistic safety assessment of Krško Nuclear Power Plant

A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Kr

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ko plant. The methodology adopted is the seismic PSA (probabilistic safety assessment). The Kr

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ko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of site hazard, calculation of plant structures response including soil–structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Relay chatter analysis and soil stability studies were also performed. The seismic PSA described here is limited to the analysis of CDF (level 1 PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Kr

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ko NPP but are not further described in this paper. The results of the seismic PSA study indicates that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable with most US and Western Europe NPPs located in high seismic areas.     

Seismic design of nuclear power plants — an assessment

This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the ‘nuclear power industry’. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation.Three methods were used in the study herein. The first involved the review of current literature, focusing primarily on publications dated later than 1970. This review included the results of numerous studies, of which those of Japanese origin and those presented in recent international conferences were predominant. The second method entailed a review of international experience in the dynamic testing of nuclear power plant structures and components, and related experience with scaled and model tests. Included in this experience, in addition to the questions of analysis, design, and measurement of dynamic parameters, are related efforts involving a review of responses obtained during measured earthquake response and investigations into appropriate methods for backfitting or upgrading older nuclear power plants to meet new seismic criteria.The third approach was to obtain the opinions and recommendations of technically knowledgeable individuals in the US ‘nuclear industry’; the survey results are shown in the Appendix.     

Resolution of flat bottom tank outlier

This paper contains the results of an outlier resolution evaluation for a large flat bottom tank, 40 ft 6 in. in diameter and 32 ft 8 in. in height. The tank was an outlier in both the USI A-46 and IPEEE programs due to insufficient strength of the bolt chair to transfer the bolt load to the side of the tank. The bolt chair evaluation resulting in the outlier was linear elastic. A more sophisticated non-linear analysis was performed of the bolt chair using the program ANSYS. The evaluation resulted in the conclusion that the bolt chair was able to transfer almost the entire yield strength of the bolt without excessive deformation that could ultimately cause overall tank failure. This evaluation tremendously increased the seismic capacity of the tank and resolved the outlier for both programs. The tank outlier evaluation also included a evaluation of soil–structure interaction (SSI) effects on the seismic demand on the tank. However, the formal consideration of SSI had a small effect on the overall seismic demand.     

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 risk analysis and decisions for nuclear power plants

Seismic risk analysis and associated sensitivity studies constitute a part of the Seismic Safety Margins Research Program being conducted by the Lawrence Livermore National Laboratory for the US Nuclear Regulatory Commission. Although seismic risks are an important contributor to the total nuclear risk, the occurrence of earthquake-related seismic phenomena is low. Safety decisions involving seismic hazards must be made, however. This paper briefly described several categories of decisions that can be made using seismic risk analysis. While risk analysis does not provide all the information required for these decisions, it is a useful tool in that it provides additional information for the decision-making process. We anticipate a growing interest in the use of seismic risk analysis in nuclear safety evaluations.     

Seismic safety of nuclear power plants in eastern Europe

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on "Benchmark study for the seismic analysis and testing of WWER type nuclear power plants". These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.     

Simplified seismic analysis methods in Belgium

Seismic design and analysis of nuclear plant systems, structures and components have requested huge effort and tremendous costs in the past two decades. The extended use of sophisticated, linear response type methods (modal analysis, spectral response) and the associated conservatism are responsible for the significant stiffening of the piping systems and the multiplication of supports and snubbers. The remedy used against the seismic risk seems worse than the pain itself, and safety might be impaired rather than improved. Indeed, system stiffening increases the average load level in normal operation (stresses, fatigue, nozzle loads, etc.); supports do not behave ideally as assumed (friction, rust, etc.) and snubbers are remarkably unreliable. On the other hand, experience with actual earthquakes shows that industrial facilities designed using very simplistic seismic techniques, or even no seismic requirement at all, suffer essentially no damage, even in the case of a large earthquake. This paradox challenges the traditional seismic design techniques, and appeals for revised seismic qualification methods of piping systems. When the assumption of the occurrence of an earthquake event is made in a plant in operation, which has not been designed against seismic criteria, the use of the standard seismic qualification techniques is still more questionable; simplified (quasi-static) techniques offer in this case a valuable and economically justified alternative. The paper describes the application of the quasi-static “modified load coefficient method” to the seismic assessment of the piping in a nuclear plant in operation, designed during the pre-seismic era.     

Dynamic analysis of a nuclear reactor with fluid–structure interaction: Part I: Seismic loading, fluid added mass and added stiffness effects

The present paper is related to the dynamic (seismic) analysis of a naval propulsion ground prototype (land-based) nuclear reactor with fluid–structure interaction modelling. Many numerical methods have been proposed over the past years to take fluid–structure phenomenon into account in various engineering domains, among which nuclear engineering in seismic analysis. The purpose of the present paper is to make a comparative study of these methods on an industrial case, namely the pressure vessel and internals of a nuclear reactor. A simplified model of the pressure vessel and the internal structure is presented; fluid–structure interaction is characterised by added mass, added stiffness and coupling effects. The basic principles of the mathematical techniques for fluid–structure modelling and dynamic methods used in the analysis are first presented and then applied to compute the eigenmodes and the dynamic response of the fluid–structure coupled system with various numerical procedures (quasi-static, spectral and temporal approaches). Numerical results are presented and discussed; fluid–structure interaction effects are highlighted. As a main conclusion, added mass effects are proved to have a significant influence on the dynamic response of the nuclear reactor.     

Department of energy’s annealing prototype demonstration projects

One of the challenges utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels. These vessels are exposed to neutrons during the operation of a reactor. For certain plants, this exposure can cause embrittlement of some of the vessel welds, which can shorten the useful life of the vessel. This reactor pressure vessel embrittlement issue has the potential to affect the continued operation of a number of US pressurized water reactor plants. However, the properties that are degraded by long-term irradiation can be recovered through a thermal annealing treatment of the vessel steel. Although a dozen Russian-designed and several US military vessels have been successfully annealed, US utilities concluded that an annealing demonstration using a US reactor pressure vessel was a prerequisite before annealing a licensed US nuclear power plant. In May 1995, the Department of Energy and Sandia National Laboratories initiated a program to evaluate the feasibility of annealing US licensed plants using two different heating technologies. One team completed its annealing prototype demonstration in July 1996, using an indirect gas-fired furnace at the uncompleted Public Service of Indiana’s Marble Hill nuclear power plant in southern Indiana. The second team’s annealing prototype demonstration using a direct heat electrical furnace at the uncompleted Consumers Power Company’s nuclear power plant at Midland, Michigan, was scheduled to be completed in early 1997, but has now been delayed indefinitely. This paper describes the Department of Energy’s annealing prototype demonstration program and the results to date for each project.     

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.     

An evaluation of structural failure modes for prestressed concrete containments

Prestressed Concrete Containment Vessels (PCCVs) refer to a popular type of containment used in the United States for Pressurized Water Reactors (PWRs).This paper presents analytically predicted ultimate pressures and seismic levels for PCCVs, considering various modes of failures. Results for six containments are presented, and correlated with the available test data.The analytical methods use either classical techniques or finite element analyses. The ultimate capacity calculations are based upon conservative deterministic estimates of strength of the structure, under both internal pressure and earthquake loads.The results indicate the following: internal pressure capacities of PCCVs built in the US are almost uniformly equal to 2.5 times the design pressure; seismic capacities are at least two times the design level, but they vary widely among the PCCVs depending on the foundation characteristics; seismic capacity of a PCCV decreases with internal pressure; and a PCCV is expected to contribute very little to the overall seismic risk of a nuclear power plant.     

基于随机模拟一致危险性谱的核电厂抗震性能分析

为改善概率地震危险性分析对震源传播特性考虑的不足,提出采用随机模拟与概率地震危险性分析结合的方法,充分考虑反应谱生成中震源机制、传播路径和场地效应等影响,生成更为精确的一致危险性谱。结合核电厂具体场地条件对场地近两千年的历史地震进行模拟,并生成同一超越概率下的一致危险性谱（UHS）。为了比较已有的厂址谱（SL-2）和安评报告中的UHS及美国RG1.60谱所生成的地震动对结构抗震性能的影响,以某核电结构为例,建立三维有限元模型,进行动力时程分析。结果表明：不同反应谱对结构的动力响应差别较大,UHS与SL-2对结构的响应较为接近,且略大于SL-2,但小于美国RG1.60谱。基于随机模拟方法生成的一致危险性谱可为核电厂抗震设计提供参考。     

Bell-ring vibration response of nuclear containment vessel with attached masses under earthquake motion

There are two types of vibrations, designated as ‘beam-type’ and ‘bell-ring type’ occurring with axisymmetric thin shell nuclear containment vessel. Up to this time, the seismic analysis for such thin axisymmetric shells has mostly been carried out only for the ‘beam-type vibration’ because the response participation factor for the ‘bell-ring type vibration’ under seismic motion is zero when the shell structure is perfectly axisymmetric. However, as with nuclear containment vessels, when the thin axisymmetric shell has several attached heavy masses such as the equipment hatch or the manholes, the resulting seismic response of bell-ring type vibration is unexpectedly large and becomes remarkably more important than the beam-type vibration.For the seismic analysis of bell-ring type vibration an approximate uncoupled analysis using the natural mode shapes of unweighted perfect axisymmetric shell has been advocated on the assumption that the effect of the attached mass on their natural modes might be very small. However, application of this method to some models showed that the response of bellring type vibration calculated was noticeably smaller than the experimental results.In this paper we show the seismic response analysis of the bell-ring type vibration coupled with the beam-type vibration through the attached masses with the new consideration. These results show good agreement between the theoretical calculation and the experiment.     

ENEA activities on seismic isolation of nuclear and non-nuclear structures

Work on seismic isolation of nuclear and non-nuclear structures was started by ENEA in cooperation with ISMES in 1988. The first activity consisted of a proposal for guidelines for seismically isolated nuclear plants using high-damping, steel-laminated elastomer bearings. This is being performed in the framework of an agreement with General Electric Company. Furthermore, research and development (R&D) work has been defined and recently initiated to support development of the seismic isolation guidelines as well as that of qualification procedures for seismic isolation systems in general. The present R&D work includes static and dynamic experiments on single bearings, shake table tests with multi-axial simultaneous excitations on reduced-scale mockups of isolated structures supported by multiple bearings, and dynamic tests on large-scale isolated structures with on-site test techniques. It also includes the development and validation of finite-element nonlinear models of the single bearings, as well as those of simplified design tools for the analysis of the isolated structures' dynamic behavior. Extension of this work is foreseen in a wider national frame.     

Reduced seismic loads for temporary conditions

A new methodology, developed under an EPRI Tailored Collaboration Project, to calculate and apply reduced seismic loads (RLSs) for evaluation of temporary conditions (TCs) in nuclear power plants using design-basis (DB) allowables is described. The methodology, which was submitted to Nuclear Regulatory Commission (NRC) through the Nuclear Energy Institute (NEI), calculates load reduction factors based on an allowed limit for time-averaged increase in seismic core damage frequency within the duration of a refueling cycle. For this allowable in the range 5×10⁻⁶ to 1×10⁻⁵ per reactor year, substantial reduction relative to DB seismic load is possible. The methodology is equally applicable to plants with and without seismic probabilistic risk analysis model.     

Seismic isolation development for the US advanced liquid-metal reactor program

GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US Department of Energy (DOE). The objective of this development is to provide, by the turn of the century, a reactor concept with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactors and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long-duration earthquakes. Seismic isolation is accomplished with high-damping natural-rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity.This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program, which is supported by the US Department of Energy (DOE), Argonne National Laboratory (ANL), Energy Technology Engineering Center (ETEC), the University of California at Berkeley (UC-Berkeley), General Electric (GE), and Bechtel National, Inc. (BNI), is described in this paper and selected results are presented. The initial testing indicated excellent performance of high-damping natural-rubber bearings. The development of seismic isolation guidelines is in progress as a joint activity between ENEA of Italy and the GE Team.     

US experience in seismic re-evaluation and verification programs

The purpose of this paper is to present a summary of the development of the seismic re-evaluation program for older nuclear power plants in the US. The principal focus of this re-evaluation is the use of actual strong motion earthquake response data for structures, mechanical and electrical systems and components. These data are supplemented by shake table test results. This type of seismic re-evaluation has lead to major cost reduction as compared to more conventional analytical and component specific testing procedures.