A simplified model for nonlinear seismic response analysis of equipment cabinets in nuclear power plants

An analytical model is required when the analysis method or the combined analysis and modal testing method is used as an aid to the seismic qualification (SQ) of equipment cabinets within nuclear power industry. This study proposes a simplified and computationally efficient model to represent the nonlinear dynamic behavior of the cabinet during earthquake. The presented model accounts for the softening behavior of the cabinets by incorporating the Duffing's type of restoring force. The characteristic of nonlinear restoring force for the finite element model (FEM) is based on the relationship of stress-strain of the element. Experiments have also been performed on an actual cabinet of nuclear power plant (NPP) to validate the model. The softening or reduction in dynamic stiffness of cabinets with increase in the excitation levels is observed in the experiments. It is also found that material yielding is not a significant source of the nonlinear behavior of the cabinet. The results obtained from the analysis using the proposed model are found to be in good agreement with the experimental results. The proposed model is expected to be useful for the prediction of seismic behavior of cabinets, particularly during the operation, owing to less computational effort required, accurate prediction of softening and no requirement of tests.     

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 analysis of nuclear power plants by using three-dimensional finite element models: a review

ABSTRACT

Seismic design of nuclear power plants (NPPs) is important for ensuring their integrity during earthquakes. Seismic analysis has been conducted using lumped mass beam models (LMBMs) for the design of plants in Japan, whereas three-dimensional (3D) finite element models (FEMs) have been used for novel plants outside Japan. The purposes of this study are to organize issues related to the development and application of 3D FEMs for seismic analysis of Japanese NPPs and to indicate future study directions. To organize these issues, the authors systematically investigated: (1) international guides and standards related to seismic analysis and (2) 3D FEMs of novel NPPs outside Japan. By considering other studies on the issues, the authors suggest directions for future studies. Resolving the issues will contribute to application of 3D FEMs for seismic analysis in the design of Japanese NPPs.     

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.     

Seismic protection technology for nuclear power plants: a systematic review

Seismic protection systems (SPS) have been developed and used successfully in conventional structures, but their applications in nuclear power plants (NPPs) are scarce. However, valuable research has been conducted worldwide to include SPS in nuclear engineering design. This study aims to provide a state-of-the-art review of SPS in nuclear engineering and to answer four significant research questions: (1) why are SPS not adopted in the nuclear industry and what issues have prevented their deployment? (2) what types of SPS are being considered in nuclear engineering research? (3) what are the strategies for location of SPS within NPPs? and (4) how may SPS provide improved structural performance and safety of NPPs under seismic actions? This review is conducted following the procedures of systematic reviews, where possible.

The issues concerning the use of SPS in NPPs are identified: cost, safety, licensing and scarcity of applications. NPPs demand full structural integrity and reactor's safe shutdown during earthquake actions. Therefore, horizontal isolation may be insufficient in active seismic zones and isolation in the vertical direction may be required. Based on the results in this review, it is likely that next generation reactors in seismic zones will include state-of-the-art SPS to achieve full standardised design.     

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.     

Seismic isolation of nuclear power plants — EDF's philosophy

The elastomer bearing pads used since 1963 as supports for prestressed concrete pressure vessels (PCPVs) was quickly chosen by Electricité de France (ED) to improve the capability of nuclear power plants (NPPs) to withstand strong earthquakes and to reduce the seismic loads on structures and equipment. The standardized units for 900 and 1300 MW(e) pressurizedwater reactor (PWR) plants have moderate seismic design loads of 0.2 and 0.15 g, respectively. These design loads were exceeded by the site dependent spectra of Cruas (France) and Koeberg (South Africa). To keep the plant design unchanged and to take the advantages of standardization, these units were put on laminated bearings with or without sliding plates. For the future French 1500 MW(e) fast breeder reactors (FBRs), which are more sensitive to seismic loads, the base isolation is considered by EDF at the beginning of the design, even for low ground motions of 0.1 g. The buildings are placed on laminated bearings while the reactor block is supported by springs and dampers. The isolated plant has identical costs as a conventional design such as SPX1 at Creys—Malville.     

Seismic response of reactor vessel internals for Korean standard nuclear power plant

This paper presents the development of seismic design criteria for the reactor vessel internals as a part of the standardization programme for the nuclear power plant in Korea. The seismic design loads of the reactor vessel internals are calculated using the reference input motions of reactor vessels taken from Yonggwang nuclear power plant units 3 and 4 which are being constructed in Korea. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components to the reactor vessel motions is carefully investigated.     

Priorities for using low-capacity nuclear power plants in Eastern Russia

Studies validating rational variants of the development of energy supplies for distributed consumers in the eastern regions of Russia on the basis of the availability of energy resources, transport access, and economic efficiency are indicated in this article. The efficiency and site conditions for the best expansion of a centralized electricity supply, siting of mini thermal and power plants using local fuel, renewable energy sources, low-capacity nuclear power plants, and gas in gas-diesel electricity plants are presented. Proposals are made for developing rational schemes for supplying energy to consumers supplied from the Chaun-Bilibino power system in the Chukotka Autonomous District.     

Seismic isolation for nuclear power plants: Technical and non-technical aspects in decision making

Technical aspects of seismic isolation systems show merit for their use in nuclear power plants. Less quantifiable non-technical aspects must be evaluated in the decision to employ a seismic isolation system.First, non-technical aspects are discussed. An historical and applications perspective is given, and it is suggested that the number of applications of seismic isolation systems is correlated with the amount of research activity in this area. For nuclear plants, it is suggested that application of seismic isolation systems is in part related to standardized plant designs in high seismic regions. Also, for nuclear plants, it is suggested that direct capital cost, enhanced seismic safety, regulatory licensing and unknown locations of nearby active faults are all factors which can weigh in favor and/or not in favor for seismic isolation application.Second, technical aspects are discussed. The technical results show that seismic isolation reduces building response, and reduces floor response spectra/equipment response. These results combine in application to reduce seismic risk and thus enhance safety for nuclear plants.     

Flood hazards for nuclear power plants

Flooding hazards for nuclear power plants may be caused by various external geophysical events. In this paper the hydrologic hazards from flash floods, river floods and heavy rain at the plant site are considered. Depending on the mode of analysis, two types of hazard evaluation are identified: (a) design hazard which is the probability of flooding over an expected service period, and (b) operational hazard which deals with real-time forecasting of the probability of flooding of an incoming event. Hazard evaluation techniques using flood frequency analysis can only be used for type (a) design hazard. Evaluation techniques using rainfall-runoff simulation or multi-station correlation can be used for both types of hazard prediction.     

Welding consumables for nuclear power plants

In nuclear power plants, submerged arc welding and covered arc welding have long been employed especially for main weld seams, including the core region of RPV.This paper investigates the mechanical properties of several welding consumables we have developed for industrial plants — that is, welding consumables which lower the phosphorus and copper content of the welded metal, those for plates possessing particularly high tensile strength and those for the narrow gap welding method.Recent data derived from irradiation embrittlement tests show that these welded metals using a non-copper coating are highly effective in minimizing shifts in the transition curve.Welding consumables for A533B C1.2, A543 C1.1 or A508 C1.4 steels have a higher tensile strength than those for A533B C1.1 or A508 C1.3.We have developed submerged arc and covered arc welding consumables to be used with these kinds of steels, and it was confirmed that these consumables possess excellent tensile strength and notch toughness.Our tests also confirmed that the narrow gap SAW and MIG welds are more efficient than the conventional ones. Moreover, the mechanical properties of the welded metals are also excellent.     

Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of a one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the non-boiling and boiling region) and the necessary connecting coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), and perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding ‘open’ loop considerations, it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established.From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using the digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method.     

Systematic methodology for diagnosis of water hammer in LWR power plants

The paper gives the dimensions of the knowledge base that is necessary to carry out a diagnosis of water hammer susceptibility/root cause analyses for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plant systems. After introducing some fundamentals, water hammer phenomena are described. Situations where each phenomenon is encountered are given and analytical models capable of simulating the phenomena are referenced. Water hammer events in operating plants and their inclusion in the knowledge base is discussed. The diagnostic methodology is presented through an application on a system in a typical light water reactor plant. The methodology presented serves as a possible foundation for the creation of an expert water hammer diagnosis system.     

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components.This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.     

The real-time functional test facility for advanced instrumentationand control in nuclear power plants

Testing and validation of the functions and performance of the digital instrumentation and control (I&C) system should be done prior to installation in nuclear power plants. The objective of the I&C Functional Test Facility (FTF) is to test and validate the functions of developed digital control and various monitoring systems. The FTF provides the simulated testing environment as an experimental test bed. The FTF software consists of a mathematical modeling program which simulates a three-loop 993 MWe pressurized water reactor and a supervisory program that comprises all the instructions necessary to run the FTF. The hardware equipment provides an interface between a host computer and a simple test panel or the developed target systems to be tested. The graphical user interface supports an easy and friendly interface between the FTF and users. It is implemented through a Picasso-3 graphic tool developed by the Halden Reactor Project. The FTF is applied to an advanced I&C system prototype, such as an automatic start-up intelligent control system, dynamic alarm system, accident identification system, and intelligent logic tracking system, to test its algorithm or performance. The results of the test show good operational performance of the FTF in normal and transient conditions     

Earthquake response spectra for seismic design of nuclear power plants in the UK

Earthquake actions for the seismic design of nuclear power plants in the United Kingdom are generally based on spectral shapes anchored to peak ground acceleration (PGA) values obtained from a single predictive equation. Both the spectra and the PGA prediction equation were derived in the 1980s. The technical bases for these formulations of seismic loading are now very dated if compared with the state-of-the-art in this field. Alternative spectral shapes are explored and the options, and the associated benefits and challenges, for generating uniform hazard response spectra instead of fixed shapes anchored to PGA are discussed.