Japanese regulatory practices on licensing, inspection, operational supervision and related standards

Japanese view on the safety of nuclear power plants is based on the concept that the primary responsibility for securing safety lies on electric power companies, installers of reactors.Under this concept, the Ministry of International Trade and Industry (MITI), in the course of designing and construction, has been performed an examination of the basic design and the detailed design of nuclear power plants, and in each stage of construction, a pre-operational inspection process. In addition, MITI, in operating stage, has been made throughgoing investigations on the causes of troubles and incidents as well as accidents that may affect operation, forcing utilities to take measures to prevent recurrence, and implementing safety regulation based on the “preventive maintenance” including elaborate checkings and overhaulings at the periodical inspections conducted for a period of three to four months after every 12-month operation cycle under the laws and regulations.This paper discusses the current status of nuclear power development in Japan, safety regulatory systems, views on safety and future prospects of securing safety.     

A 10 MWe pressurized water reactor for safe and reliable power

A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve “walk-away” safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control.     

Development of aseismic design of piping, vessels and equipment in nuclear facilities

This report involves the development of aseismic design procedures of piping, vessels and equipment in Japan. These mechanical structures show their various characteristics of vibration. Pressure boundaries, a containment vessel and safety systems belong to such structures. The vital components of nuclear power plants are classified to “A” class according to the classification for the aseismic design in Japan. All components in “A” class are required to be based on dynamic earthquake-resistant design, of which level is decided in consideration of local seismisity.

For dynamic design purposes, the following processes are the most important: 1. estimating eigenfrequencies and modes of the system; 2. estimating its damping characteristics; 3. estimating the behavior of the system during strong earthquakes; 4. deciding the design criteria, especially the allowable stresses to earthquake loadings.     

Aseismic evaluation for aging degradation of nuclear power plant components

The commercial operation of light water reactor plants in Japan already has a history of nearly 30 years. Since the beginning of the 1990s, studies have been conducted on aging degradation of nuclear power plants in Japan and abroad and, earlier in 1999, the domestic program of plant life management (PLM) was settled on. The program is based on the results of the PLM Study, which started in 1997. The purpose of the study was to develop the preventive maintenance program with an evaluation of aging degradation for maintaining the functions of plant component equipment. Taking account of the need for proper management of aging degradation, meanwhile, the technical evaluation of aseismic capability of aged plants is also considered to be important. Based on this concept, we evaluated the impact of assumed aging degradation on the aseismic capability of the plant facilities and structures covered by the PLM Study. In the aseismic evaluation, aging degradation modes selected in the PLM Study were divided into two categories—the one is including some degradation modes which impact on the aseismic capability of the facilities and structures should be taken into account, the other is including those whose impact might be ignored. Then the aging degradation modes composing the former one were quantitatively evaluated primarily based on the Technical Guidelines for Aseismic Design of Nuclear Power Plants (JEAG-4601) (NUREG/CR-6241, 1987). The result of the evaluation indicated that no aging degradation mode to be reflected in the maintenance program was extracted from the viewpoint of securing the aseismic capability of the plant components. However, establishment of rational evaluation methods for aging degradation, e.g. aseismic capability evaluation of thinned piping systems, was made a future technical subject.     

Methods and requirements for seismic design in the United Kingdom

The United Kingdom is in an area of low but significant seismicity compared with the more active areas of the world where there are major active faults or tectonic plate boundaries. This paper presents the methods and requirements that are adopted to consider the extreme load in the design of nuclear facilities. In the United Kingdom, detailed procedures for demonstrating seismic adequacy are not specified by the nuclear licensing authority and as such the methods described in this paper are based on precedents arising from recent licensing applications. In presenting the method and requirements, the paper discusses the applicability of simplified methods for seismic qualification for both “new” and “existing” facilities. The paper concludes that simplified methods are applied to a significant extent for demonstrating the adequacy of existing plant. However, for new plant these methods have been limited in some cases to the evaluation of design loads and to the qualification of items where the required degree of assurance is less than that associated with formal qualification and for supporting studies which do not directly affect design. It is expected that as the body of experience in earthquake engineering develops in the United Kingdom, there will be a greater tendency to adopt more simplified procedures with a greater degree of confidence.     

On the basic research of design analysis and testing based on the failure rate for pipings and equipment under earthquake conditions

This paper deals with the evaluation method of the failure rate of pipings and equipment of nuclear power plants under destructive earthquakes and a new design concept in this stand point of view. These researches are supported by various studies related to this subject, which have been done by the author since 1966. In this paper, the history of the development, the summaries of these studies and their significances to the practice will be described briefly.The surveys on damages of industrial facilities caused by recent destructive earthquakes are the basical study for this subject. And the continuous response observation of model structures of a plant complex to natural earthquakes is another important basic study to know the stochastic nature and significance of response analysis for the anti-earthquake design of nuclear power plants.By having the exact knowledges on these subjects, the author has been developing the evaluation procedure of the failure rate of pipings and equipment under destructive earthquake conditions, a new design method ‘counter-input design’ and others. Now his effort is going towards to establish their practical procedure after finishing the basic researchers.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.     

Generating material strength standards of aluminum alloys for research reactors I. Yield strength values Sy and tensile strength values Su

Aluminum alloys are frequently used as structural materials for research reactors. The material strength standards, however, such as the yield strength values (S_y), the tensile strength values (S_u) and the design fatigue curve—which are needed to use aluminum alloys as structural materials in “design by analysis”—for those materials have not been determined yet. Hence, a series of material tests was performed and the results were statistically analyzed with the aim of generating these material strength standards. This paper, the first in a series on material strength standards of aluminum alloys, describes the aspects of the tensile properties of the standards. The draft standards were compared with MITI no. 501 as well as with the ASME codes, and the trend of the available data also was examined. It was revealed that the draft proposal could be adopted as the material strength standards, and that the values of the draft standards at and above 150°C for A6061-T6 and A6063-T6 could be applied only to the reactor operating conditions III and IV. Also the draft standards have already been adopted in the Science and Technology Agency regulatory guide (standards for structural design of nuclear research plants).     

基于TMSR-LF1的Ⅱ类研究堆抗震设计方法研究

抗震设计是核设施为满足安全与经济综合要求进行设计时的重要内容,目前研究堆的抗震设计缺乏相应的规范与研究,尚未发现较为完善的方法体系。本文推荐了一个匹配结构与设备的Ⅱ类研究堆抗震设计方法,以50 a超越概率2%地震动作为安全停堆地震（SSE）,并以2 MW液态燃料钍基熔盐实验堆（TMSR-LF1）为例,对比分析了采用该方法与采用其他相关规范方法得到的设计反应谱（DRS）,并将其应用于结构和设备的抗震设计计算中。结果表明：推荐方法在满足结构与设备的抗震设计匹配性的前提下,相比核电规范具有较好的经济性,相比民用规范具有较好的保守性,更加合理。     

The DEEPSSI project, design, testing and modeling of steam injectors

The DEEPSSI project is a steam injector research programme. Among thermal-hydraulic passive systems, the steam injectors (also called “condensing ejectors” or “steam jet pumps”) are very interesting apparatus with very specific characteristics (high velocity, very low pressure). The envisaged reactor application is the Steam Generator Emergency FeedWater System (EFWS) of Pressurised Water Reactors (PWRs). The heart of this project is the development and the testing of an innovative steam injector design. Three experimental facilities are involved: CLAUDIA in France, IETI in Italy and IMP-PAN in Poland. In these facilities, different design options have been tested and some significant improvements of the initial design have been obtained.In addition to the experimental studies, the development of a steam injector computational model has been undertaken in order to model industrial systems based on steam injectors. The one-dimensional module of the system code CATHARE2 has been chosen to be the basis of this model. The first results obtained have confirmed the capabilities of CATHARE2 to describe the steam injector thermal-hydraulics.     

The SQUG program for resolution of USI A-46 — Status and implementation plans

In response to U.S. Nuclear Regulatory Commission (NRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment and Operating Nuclear Power Plants”, the Seismic Qualification Utility Group (SQUG), with the support of the Electric Power Research Institute (EPRI), developed a comprehensive program to verify the seismic adequacy of equipment in operating nuclear power plants. The primary thrust of the program has been the development of procedures, criteria, and data to apply actual experience on the performance of equipment during earthquakes to the verification of seismic ruggedness of similar equipment in nuclear plants. While the use of such experience data continues to play a primary part in the SQUG program for resolution of USI A-46, the overall SQUG program includes a number of other significant elements which, taken together, provide a comprehensive approach for verification of the seismic adequacy of equipment in nuclear plants. These elements of the SQUG program include the assimilation and use of seismic shake table data in a generic way; the development of simplified analytical tools and criteria for evaluation of equipment anchorage, tanks, heat exchangers and cable trays; and the development of procedures for identifying and evaluating electrical relays, which are essential to plant shutdown in response to an earthquake. Procedures and data bases for performing and documenting the various seismic evaluations and plant walkdowns, and a program for training the large number of engineers who will be required to implement the SQUG methodology, have also been developed. This paper describes the main elements of the SQUG program for resolution of USI A-46 and provides a status report on the plans for their implementation in SQUG member plants.     

Design tools for computer-generated display of information to operators

More and more computers are being used to process and display information to operators who control nuclear power plants. Implementation of computer-generated displays in power plant control rooms represents a considerable design challenge for industry designers. Over the last several years, the Electric Power Research Institute has conducted research aimed at providing industry designers tools to meet this new design challenge. These tools provide guidance in defining more “intelligent” information for plant control and in developing effective displays to communicate this information to the operators.     

User effects on the thermal-hydraulic transient system code calculations

Large thermalhydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. During these comparisons of the code results, there has been a continuous debate on the way how the code user influences the predicted system behaviour. This rather subjective element might become a crucial point with respect to the quantitative evaluation of the code uncertainties which is essential if the “best estimate codes are used for licensing procedures”.The International Standard Problem Exercises (ISPs) proposed by the OECD/NEA-Committee for the Safety of Nuclear Installations (CSNI) and by IAEA (International Atomic Energy Agency) and thermalhydraulic code assessment activity undertaken by USNRC (US Nuclear Regulatory Commission) under International Code Assessment and Application Program (ICAP) demonstrate the large effort put in this framework by organizations all over the world. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermalhydraulic transient system codes will be presented and discussed on the basis of some case studies.The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them.     

RD & D for improving reliability of Japanese nuclear power plants

Research, development and demonstration of the technologies relating to improvement of nuclear power plant reliability in Japan are characterized by: (1) emphasis on public acceptance; (2) engineering philosophy of preventive maintenance; (3) execution of thorough countermeasures for improvement. Utilities and vendors are basically responsible for RD&D in this commercially established area. However, the government is also involved in safety related and other important items, and most of these activities are conducted under good coordination between government and private sectors. The government, more specifically, the Ministry of International Trade and Industry (MITI) is conducting various reliability proving and verification tests reaching the scale of 12 billion yen (50 million US dollars) in annual average.In close cooperation with industry MITI is also promoting sophistication of LWR technologies such as development of advanced LWR's and improvement of inspection, maintenance and operation, without being satisfied with the recent good records of nuclear power plant operations.     

On estimated modes of failure of nuclear power plants by potential earthquakes

This paper deals with the estimated ‘modes of failure’ of nuclear power plants during future violent earthquakes. The authors have been surveying the damage to industrial plants caused by several violent earthquakes since 1960. Some of them have already been reported in English, but here the authors try to rearrange them from the viewpoint of ‘modes of failures’ of nuclear power plant buildings, equipment, vessels and piping. The authors categorize the mechanisms of failure as follows: (i) damaged by the dynamic effect of acceleration waves, (ii) by resonance in displacement waves, (iii) by the static effect of seismic force, (iv) by external force from attached piping and others, or forced deformation, and (v) by liquefaction of soil.The authors try to determine the modes of failure of the following items in a matrix form of the mechanisms: (i) the reactor building, (ii) steel containment vessel, (iii) auxiliary building, (iv) reactor vessel, (v) core internals, (vi) primary and secondary coolant system, (vii) emergency power supply system, (viii) emergency gas treatment system and stack, (ix) fuel cooling pond and fuel rack, (x) refuel machine crane, (xi) auxiliary system and component, (xii) turbine and its pedestal, and (xiii) main power system and control instrumentation. They also examine them from another point of view, i.e. in ‘the classification of the important factor’ of items for their aseismic design.     

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.     

Use of power spectral density in deriving response spectral shapes for aseismic design

Mean and mean + σ response spectral shapes based on accelerograms normalized to their respective peak ground acceleration (PGA) values have been widely used in the aseismic design of critical structures such as nuclear power plants. The computation of response spectral shapes based on accelerograms normalized to their respective power spectral densities (PSDs) has been proposed. A comparison of response spectral shapes using the two alternative normalization parameters (PGA and PSD) has been presented to demonstrate that the ordinates of the PSD-based mean + σ spectral shapes, which are better representatives of earthquake vibrations, are considerably lower than those of the PGA-based spectral shapes. The mean spectral shapes in the two cases are very close to each other, suggesting that a large part of the scatter in the PGA-based shapes is attributable to the procedure adopted for normalization of the accelerogram. The PSD-based spectral shapes can play an effective role in determining the margins available in the aseismic design of structures using the PGA-based mean + σ spectral shapes. Smooth response spectral shapes based on normalization with respect to the PSD are given for rock and soil sites at the end of the paper.     

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.     

防火屏障有效性分析方法研究

该论文的目的是对当前的各类核电厂防火屏障有效性分析方法进行对比研究,为工程实践提供参考。文中对用于评估核电厂防火屏障耐火能力的等效火灾持续时间评估法(包括等效面积法、等效温度法)以及计算机模拟法分别进行了阐述,并利用这些方法对核电厂中两种比较典型的防火隔间(液压机组间和电气设备间)的防火屏障耐火能力进行了评估,通过比较各种方法的分析结果,对这些方法的优缺点及其在工程应用中的适用范围进行了评估,指出了在评估防火屏障有效性时应结合具体的情况,选取合适的方法开展评估,在实际工程应用中建议优先采用等效面积法对耐火极限进行估算,与此同时,还应尽可能地应用计算机模拟法,进行更现实的估算或进一步了解特定隔间的火灾特性,后续还应从法规、标准、规范以及方法等方面进一步完善数值评估法,推动数值评估法在国内核电厂防火设计中的使用。     

Seismic re-evaluation of nuclear facilities worldwide: overview and status

Existing nuclear facilities throughout the world are being subjected to severe scrutiny of their safety in the event of an earthquake. In the United States, there have been several licensing and safety review issues for which industry and regulatory agencies have cooperated to develop rational and economically feasible criteria for resolving the issues. Currently, all operating nuclear power plants in the United States are conducting an Individual Plant Examination of External Events, including earthquakes beyond the design basis. About two-thirds of the operating plants are conducting parallel programs for verifying the seismic adequacy of equipment for the design basis earthquake. The U.S. Department of Energy is also beginning to perform detailed evaluations of their facilities, many of which had little or no seismic design. Western European countries also have been re-evaluating their older nuclear power plants for seismic events often adapting the criteria developed in the United States. With the change in the political systems in Eastern Europe, there is a strong emphasis from their Western European neighbors to evaluate and upgrade the safety of their operating nuclear power plants. Finally, nuclear facilities in Asia are also being evaluated for seismic vulnerabilities. This paper focuses on the methodologies that have been developed for re-evaluation of existing nuclear power plants and presents examples of the application of these methodologies to nuclear facilities worldwide.