Testing to determine relay seismic ruggedness

The seismic qualification of equipment in operating nuclear plants has been identified as a potential safety concern in U.S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment in Operating Nuclear Power Plants”. In response to this concern, the Seismic Qualification Utility Group (SQUG), with support from the Electric Power Research Institute (EPRI), has undertaken a program to demonstrate the seismic adequacy of essential equipment by the use of actual experience with such equipment in plants which have undergone significant earthquakes and by the use of available test data for similar equipment. An important part of this program is the development of the methodology and test data for verifying the functionality of electrical relays used in essential circuits needed for plant shutdown during a seismic event. This paper describes the EPRI supported relay testing program to supplement existing relay test data. Many old relays which are used in safe shutdown systems of SQUG plants and for which seismic test data do not exist have been shake-table tested. The testing performed on these relays and the test results for two groups of relays are summarized in this paper.     

Large-scale seismic test program at Hualien Taiwan

The large scale seismic test (LSST) program at Hualien, Taiwan, is a follow-on to the soil-structure interaction (SSI) experiments at Lotung, Taiwan. The planned SSI studies are performed at a stiff soil site in Hualien, Taiwan, that historically has had slightly more destructive earthquakes in the past than Lotung. The objectives of the LSST program are as follows: to obtain earthquake-induced SSI data at a stiff soil site having similar prototypical nuclear power plant soil conditions; to confirm the findings and methodologies validated against the Lotung soft soil SSI data for prototypical plant condition applications; to validate further the technical basis of realistic SSI analysis approaches; to support further the resolution of USI A-40 “Seismic Design Criteria” issue. These objectives are accomplished through an integrated and carefully planned experimental program consisting of soil characterization, test model design and field construction, instrumentation layout and deployment, in situ geophysical information collection, forced vibration test, and synthesis of results and findings. The Hualien LSST is a joint effort among many interested parties. The Electric Power Research Institute (EPRI) and Taiwan Power Company (Taipower) are the organizers of the program and have the lead in planning and managing the program. Other organizations participating in the LSST program are the US Nuclear Regulatory Commission (NRC), Central Research Institute of Electric Power Industry (CRIEPI), Tokyo Electric Power Compamy (TEPCO), Commissariat A L'Energie Atomique (CEA), Electricité de France (EdF), Framatome, Korea Electric Power Corporation (KEPCO), Korea Institute of Nuclear Safety (KINS), and Korea Power Engineering Company (KOPEC). The LSST array started operation in June 1993, and is envisioned to be of five years duration.     

Regulatory perspective of seismic issues

Some of the current seismic issues facing the nuclear power industry, such as seismic design criteria (USI A-40), seismic qualification of equipment in operating nuclear power plants (USI A-46), eastern United States seismicity, operating basis earthquake (OBE) exceedance criteria, seismic instrumentation, post OBE inspection of nuclear power plants, anchor bolts too close to a free edge, seismic margins of plants, and the potential for external events to cause severe accidents, are presented and the Nuclear Regulatory Commission's perspective on the resolution of these issues are discussed.     

Zion SQUG pilot walkdown and resolution of outliers

A walkdown of Zion Unit 2 was performed in May 1987 to verify the seismic adequacy of mechanical and electrical equipment required for hot safe shutdown. This was the first pilot walkdown performed under the SQUG guidelines. Draft Generic Implementation Procedures developed by SQUG were used by the three Seismic Review Teams to walkdown 159 pieces of equipment. Based on the walkdown, 118 pieces of equipment were found to be adequate and the remaining 41 pieces were classified as outliers. Seven of these 41 outliers were further verified by subsequent walkdown and 17 were verified by additional calculations. Anchorage repairs/improvements and minor field adjustments due to seismic interaction concerns resolved the remaining 17 cases. A trend analysis of the outliers was also performed to see if other safety-related equipment, outside the scope of SQUG, should be addressed for any shortcomings. Two trends were identified, one for the anchorage of shop fabricated instrument racks, and another for bracing of HVAC duct system supplying intake air to the diesel generators. Appropriate repairs were issued for all safety-related items affected by these trends.     

Methodologies for rapid evaluation of seismic demand levels in nuclear power plant structures

A methodology for rapid assessment of both acceleration spectral peak and “zero period acceleration” (ZPA) values for virtually any major structure in a nuclear power plant is presented. The methodology is based on spectral peak and ZPA amplification factors, developed from regression analyses of an analytical database. The developed amplification factors are applied to the plant's design ground spectrum to obtain amplified response parameters. A practical application of the methodology is presented.This paper also presents a methodology for calculating acceleration response spectrum curves at any number of desired damping ratios directly from a single known damping ratio spectrum. The methodology presented is particularly useful and directly applicable to older vintage nuclear power plant facilities (i.e. such as those affected by USI A-46). The methodology is based on principles of random vibration theory. The methodology has been implemented in a computer program (SPECGEN). SPECGEN results are compared with results obtained from time history analyses.     

Seismic qualification of equipment — research needs

In general, industry has followed the Institute of Electric and Electronic Engineers' (IEEE) Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Stations (IEEE Standard 344-1975). However, the U.S. Nuclear Regulatory Commission Regulatory Guide 1.100 notes exceptions to a small part of the IEEE Standards.This paper describes research needed to reconcile the differencies between the IEEE Standard and the Regulatory Guide. In addition, the paper discusses the effects of shake table mass and stiffness on the dynamic response of equipment tested, and the effect attributable to the difference between methods of attaching to the shake table and the actual in-situ attachment method.     

Achieving cost savings through collaborative seismic testing

Seismic qualification of nuclear power plant equipment through testing has been perceived by utility personnel as a costly and complicated process. Certainly, some equipment types may only be qualified by test due to the complexity of the item and/or the inability to represent the item in a quantitative analysis. Other factors also influence the reluctance by some to resort to seismic testing. These include cost, dealing with test failures, lack of understanding of the testing process, and greater reliance on new analytical techniques. A group of utilities have allied themselves to address these issues and have formed the Seismic Qualification Reporting and Testing Standardization (SQURTS) group. SQURTS has implemented a program whereby testing is performed at a low cost, often lower than a comparable analytical solution. Testing is conducted at generic seismic levels using generic test procedures, which broadens the applicability of results. Test reports are published in a standard format which shortens the process of review and approval. Testing is no longer just a requirement, but a cost-effective option.     

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.     

Performance of a piping system and equipment in the HDR simulated seismic experiments

This paper describes a portion of the analysis and results of the United States Nuclear Regulatory Commission/Idaho National Engineering Laboratory (USNRC/INEL) participation in the SHAG (Shakergebaude) Seismic Research Program conducted by Kernforschungszentrum Karlsruhe (KfK) at the Heissdampfreaktor (HDR), a decommissioned nuclear reactor. The program analyzed the responses of a piping system and associated line-mounted equipment when subjected to various seismic and hydraulic loadings. Of interest was to evaluate the influence that piping support system flexibility has on piping system responses. The results of the studies will contribute to the technical basis for assessing the responses of light water reactor (LWR) piping and fine-mounted equipment to earthquakes.     

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.     

1978 National Symposium on Atomic Energy,Tokyo, Japan

Large components of LWR plants such as reactor containment vessel, primary recirculation system were proved the structual integrity of them against severe earthquake by means of a large-scale high-performance shaking table at Tadotsu Engineering Laboratory of Nuclear Power Engineering Test Center.

Seismic model tests clarified restoring force characteristics of reactor buildings and soil- reactor ? building interaction phenomena and also indicated that seismic behavior of the reactor buildings could be solved applying current analytical method.     

Development of seismic anchorage guidelines for nuclear plant equipment

Seismic anchorage guidelines have been developed for various classes of electrical and mechanical equipment typically found in nuclear power plants. The guidelines consist of tables and charts that give the seismic capacity of anchorage configurations as a function of the type, number, size, and location of fasteners and the dimensions of the equipment. Seismic anchorage capacities have been derived using a conventional static analysis but with more realistic assumptions and strength criteria than typically used for design. The guidelines also include simple and practical inspection procedures for verifying proper installation of existing equipment anchorages.     

核电站应急指挥中心基底隔震技术方案

日本福岛核电站事故之后,中国国家核安全局正式发布《福岛核事故后核电厂改进行动通用技术要求》,提高了核电站应急指挥中心的抗震安全要求。本文以防城港核电站应急指挥中心示范项目为分析对象,探讨了采用橡胶支座作为隔震装置的基底隔震方案,并进行了隔震效果评价。通过采用43个橡胶隔震支座,可以显著地削弱地震响应,提高应急指挥中心的抗震裕量。通过该技术方案可以保护应急指挥中心的人员、设备和系统的安全,从而提升核电站震后应急能力。分析研究成果和工程实践经验可在后续核电站建设中推广应用。     

基于经验-解析数据的某核电安全壳地震易损性分析

本文基于混合数据的地震易损性分析方法,对我国已运行核电厂地震易损性分析进行研究。首先基于地震危险性分析和分解结果,生成了我国华南地区某核电厂厂址条件谱;然后采用贪心优化算法,选取符合厂址危险性的地震动记录;基于增量动力分析方法,生成我国某核电厂安全壳地震易损性安全系数FS和FSA的解析数据;地震易损性其他参数采用经验数据,基于经验-解析数据,生成了我国某核电厂安全壳地震易损性曲线。建议将基于经验-解析数据的地震易损性分析方法应用于我国核电厂安全壳初步地震易损性分析中。     

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.     

利用HHT产生与多阻尼楼板谱匹配的地震波

美国原子能管理委员会（USNRC）规范规定了用于核电厂抗震分析和设计的地震波要求。在抗震分析和设计中，采用的地震波可与多阻尼目标反应谱匹配，也可与单阻尼目标反应谱匹配。然而，在对核电设备和部件进行动力时程分析时，则需要与多阻尼目标楼板谱匹配的地震波。基于此问题，提出利用希尔伯特-黄变换（HHT）方法，通过修改种子地震波的频率和振幅信息，使之与多阻尼目标楼板谱匹配，且完全符合USNRC规范的匹配标准，从而为核电设备和部件的地震安全评估提供合适的地震激励。      

Sharing product data of nuclear power plants across their lifecycles by utilizing a neutral model

Many public and private Korean organizations are involved during the lifecycle of a domestic nuclear power plant. Korea Plant Engineering Co. (KOPEC) participates in the design stage, Korea Hydraulic and Nuclear Power (KHNP) operates and manages all nuclear power plants in Korea, Doosan Heavy Industry and Construction Co. manufactures the main equipment, and a construction company constructs the plant. Even though each organization has its own digital data management system and obtains a certain level of automation, data sharing among organizations is poor. KHNP obtains drawings and technical specifications from KOPEC in the form of paper. This results in manual re-work of definitions, and errors can potentially occur in the process. In order to establish an information bridge between design and operation and maintenance (O&M) phases, a generic product model (GPM), a data model from Hitachi, is extended for constructing a neutral data warehouse and the Korean Nuclear Power Plant Information Sharing System (KNPISS) is implemented.     

Comparative forced vibration test of two BWR-type reactor buildings

The basis of this paper is comparative forced vibration testing of two GE 460 MW(e) BWR-type reactor buildings. The tested nuclear power plants are the Fukushima Nuclear Power Plant Unit No. 1 of the Tokyo Electric Power Company (hereinafter referred to as Fukushima) and the Shimane Nuclear Power Plant of the Chugoku Electric Power Company (Shimane). They are almost the same in both structure and function, but are built on rock of quite different rigidity. The shear wave velocity of Shimane is about three times that of Fukushima. The forced vibration tests were performed immediately after completion of each reactor building using a vibrator with a maximum exciting force of 3 t. The computer simulation analyses were carried out using vibration models possessing different internal viscous damping factors for each structural element. Both the resonance periods and damping factors of Fukushima were found to be larger than those of Shimane. Thus, site selection of nuclear power plants must be reviewed as a matter of utmost importance from the viewpoint of seismic design.