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.     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

Seismic analysis of ZPR-6 Reactor Facility

A safe shutdown earthquake analysis of ZPR 6 Reactor Facility was conducted through seismic risk analysis, soil-structure interaction analysis, reactor building dynamic time history analysis and equipment response spectrum analysis due to an assumed El Centro earthquake. Several ASME, AISC and ANSI design codes were used to demonstrate the adequacy of this facility and to design several equipment and piping supports.     

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.     

Piping research overview

The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses.     

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.     

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.     

Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the “force equivalent area” (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques.     

Response of piping system on friction support to bi-directional excitation

Installation of friction devices between a piping system and its supporting medium is an effective way of energy dissipation in the piping systems. In this paper, seismic effectiveness of friction type support for a piping system subjected to two horizontal components of earthquake motion is investigated. The interaction between the mobilized restoring forces of the friction support is duly considered. The non-linear behavior of the restoring forces of the support is modeled as an elastic-perfectly plastic system with a very high value of initial stiffness. Such an idealization avoids keeping track of transitional rules (as required in conventional modeling of friction systems) under arbitrary dynamic loading. The frictional forces mobilized at the friction support are assumed to be dependent on the sliding velocity and instantaneous normal force acting on the support. A detailed systematic procedure for analysis of piping systems supported on friction support considering the effects of bi-directional interaction of the frictional forces is presented. The proposed procedure is validated by comparing the analytical seismic responses of a spatial piping system supported on a friction support with the corresponding experimental results. The responses of the piping system and the frictional forces of the support are observed to be in close agreement with the experimental results validating the proposed analysis procedure. It was also observed that the friction supports are very effective in reducing the seismic response of piping systems. In order to investigate the effects of bi-directional interaction of the frictional forces, the seismic responses of the piping system are compared by considering and ignoring the interaction under few narrow-band and broad-band (real earthquake) ground motions. The bi-directional interaction of the frictional forces has significant effects on the response of piping system and should be included in the analysis of piping systems supported on friction supports. Further, it was also observed that the velocity dependence of the friction coefficient does not have noticeable effects on the peak responses of the piping system.     

塑性变形对管系地震动力响应的影响

在评述线弹性分析方法的基础上,阐明了在管系特别是核管系动力响应分析中考虑塑性变形影响的重要性,介绍了现有考虑塑性影响的方法及其存在的问题.指出要降低现行规范的保守性,提出合理的管系抗震设计方法,     

Ultimate strength of reinforce concrete shear walls under multi-axes seismic loads

“The model test on multi-axes loading on RC shear walls” had been carried out as for the 10-year project aiming at comprehension of the earthquake response behavior of three-dimensional (3D) reinforced concrete (RC) shear walls under the 3D of multi-axes earthquake loading condition. The motivation of the project building-up is that the current seismic design of nuclear power plant building is carried out by applying one-dimensional (1D) dynamic earthquake load to an analytical building model in each direction independently whereas actual earthquake jolts the building in the three directions simultaneously. Therefore, there were opinions requesting some testing confirm whether or not the current seismic design methodology is reliable for the input motions exceeding the design earthquake ground motion moreover for the input motions of the 3D. The project had completed with various valuable outcomes that can reply to the opinions. Moreover, the outcomes will play an important role in evaluating seismic margins of important structures in a nuclear power plant. In this paper, based on the published documents relating to this test project, the author describes a review of the whole testing and summarizes the major outcomes extracted by the test project.     

Seismic reevaluation and upgrading of nuclear power facilities outside the US using US developed methodologies

Several seismic licensing and safety issues have emerged over the past fifteen years for commercial US Nuclear Power Plants and US Government research reactors, production reactors and process facilities. The methodologies for the resolution of these issues have been developed in numerous government and utility sponsored research programs. The resolution criteria have included conservative deterministic design criteria, deterministic seismic margins assessments criteria (SMA) and seismic probabilistic risk assessment criteria (SPRA). The criteria for SMAs and SPRAs have been based realistically on considering the inelastic energy absorption capability of ductile structures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate the operability of complex mechanical and electrical equipment. Most of the applications to date have been confined to the US, however there have been several applications to Asian, Western and Eastern Europe reactors. This paper summarizes the major issues addressed, the development of reevaluation criteria and selected applications to non US reactors including VVER reactors of Soviet origin.     

Condition telemonitoring and diagnosis of power plants using web technology

The monitoring and diagnostic systems currently installed in power plants generally supply information for control room displays and for on-site personnel. Telemonitoring is also frequently used. In this case, relevant diagnostic data are transmitted remotely to a special laboratory for analysis using highly specialized equipment and software.

The appearance of the terms “Monitoring” and “Diagnosis” alongside the term “Web Technology” in the title of this paper does not mean that remote access to diagnostic systems over the Internet is being presented here as a simple extension of the existing situation.

Condition telemonitoring and diagnosis based on Web technology is a new departure in diagnostic system design philosophy. It is the technology used to integrate diagnostic systems into a customer's IT infrastructure (Intranet or Internet).

Siemens has started to use Web-based condition telemonitoring and diagnosis in some power plants (nuclear and fossil-fueled) to provide a global source of specialist support.     

Seismic resistant design of safety class structures and equipment

Some commonly encountered problems in the seismic resistant design of nuclear power plant facilities are discussed. The topics included here are ground input motions, local geology versus source mechanism and travel path, three components inputs, torsional responses, floor response spectra, seismic resistant design of heavy equipment, the application of component mode synthesis technique, seismic resistant design of piping systems, equipment qualification by testing, the effects of close modes, underground pipe design, and soil structure interaction.     

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.     

Positive use of damping devices for piping systems — Some experiences and new proposals

An increase of the damping ratio is known to be very effective for the seismic design of a piping system. It is reported that the energy dissipation in piping supports contributes to increase the damping ratio of the piping system. In this paper, with regard to increasing the damping and reducing the seismic response of the piping system, three application methods of damping devices used in other engineering fields are reviewed: (1) direct damper, (2) dynamic vibration absorber, and (3) connecting damper. Based on the results of this review, the following three types of damping devices for piping systems are introduced: (1) visco-elastic dampler (direct damper), (2) elasto-plastic damper (direct damper), and (3) compact dynamic absorber (dynamic vibration absorber). The dynamic characteristics of these damping devices are investigated by a component test and the applicability of them to the piping system was confirmed by the vibration test using a three-dimensional piping model. These damping devices are more effective than mechanical snubbers to suppress the vibration of the piping system.     

The effects of reactor coolant system pipe rupture motion on tributary piping and attached equipment

This paper presents a rigorous analysis of a pressurized water reactor coolant system (RCS) to determine time-history excitations of intact equipment and tributary piping attached to the RCS caused by a postulated guillotine rupture in the primary coolant piping. Reactor control rods and drive mechanisms, in core instrumentation guide tubes and reactor coolant pump motor appurtenances are examples of attached equipment which is excited by RCS LOCA induced motions. The surge line, mainsteam lines and emergency core cooling lines are examples of tributary piping similarly affected by RCS LOCA induced motions. The methods described herein include structural and dynamic modeling and analytical techniques used in the non-linear transient dynamic time-history analysis of a 3-D coupled model of the RCS. The results of this analysis are generated for the purpose of defining the excitation for subsequent analysis of intact tributary systems attached to the reactor coolant system in order to evaluate their response to those LOCA induced motions. This paper also presents results of analyses for intact tributary piping subjected to LOCA induced motions and assesses the severity of the response compared to typical seismic response.     

Safety performance of the 4S reactor on the ATWS events – statistical estimation of uncertainty

4S reactor is a sodium-cooled fast reactor developed as a small-decentralized power supply. The name of “4S” in this reactor stands for Super-Safe, Small and Simple, and they show representative features of the reactor.

The purpose of the present work is to evaluate quantitatively the super-safety of 4S reactor, and the safety performance is analyzed with ARGO-3, which is a plant dynamics code of a sodium-cooled fast reactor.

In this evaluation, some events, such as Unprotected Loss of Flow (ULOF) and Unprotected Transient Overpower (UTOP), are selected as typical cases from various transients and accidents. After metrics concerned with safety design is defined for each event, it is evaluated with statistical methods whether each metric satisfies acceptance criteria in a given criteria level.

Result about ULOF is as follows. The coolant temperature in the nominal hottest assembly outlet, “T_c” is selected as metric, and the upper side value of 95% confidential section in T_c is below 900 °C that is acceptance criteria. Also in UTOP, it is shown that the fuel maximum temperature in the nominal hottest assembly, “T_f” satisfies acceptance criteria. This result shows that 4S reactor has margin for safety acceptance criteria.     

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。     

Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value.