Analysis of failed nuclear plant components

Science.gov (United States)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

An estimation method of system failure frequency using both structure and component failure data

International Nuclear Information System (INIS)

Takaragi, Kazuo; Sasaki, Ryoichi; Shingai, Sadanori; Tominaga, Kenji

1981-01-01

In recent years, the importance of reliability analysis is appreciated for large systems such as nuclear power plants. A reliability analysis method is described for a whole system, using structure failure data for its main working subsystem and component failure data for its safety protection subsystem. The subsystem named main working system operates normally, and the subsystem named safety protection system acts as standby or protection. Thus the main and the protection systems are given mutually different failure data; then, between the subsystems, there exists common mode failure, i.e. the component failure affecting the reliability of both two. A calculation formula for sytem failure frequency is first derived. Then, a calculation method with digraphs is proposed for conditional system failure probability. Finally the results of numerical calculation are given for the purpose of explanation. (J.P.N.)

GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

Energy Technology Data Exchange (ETDEWEB)

S. A. Eide; S. V. Chmielewski; T. D. Swantz

1990-02-01

A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

Seismically induced common cause failures in PSA of nuclear power plants

International Nuclear Information System (INIS)

Ravindra, M.K.; Johnson, J.J.

1991-01-01

In this paper, a research project on the seismically induced common cause failures in nuclear power plants performed for Toshiba Corp. is described. The objective of this research was to develop the procedure for estimating the common cause failure probabilities of different nuclear power plant components using the combination of seismic experience data, the review of sources of dependency, sensitivity studies and engineering judgement. The research project consisted of three tasks: the investigation of damage instances in past earthquakes, the analysis of multiple failures and their root causes, and the development of the methodology for assessing seismically induced common cause failures. The details of these tasks are explained. In this paper, the works carried out in the third task are described. A methodology for treating common cause failures and the correlation between component failures is formulated; it highlights the modeling of event trees taking into account common cause failures and the development of fault trees considering the correlation between component failures. The overview of seismic PSA, the quantification methods for dependent failures and Latin Hypercube sampling method are described. (K.I.)

Results of an aging-related failure survey of light water safety systems and components

International Nuclear Information System (INIS)

Meale, B.M.; Satterwhite, D.G.; MacDonald, P.E.

1988-01-01

The collection and evaluation of operating experience data are necessary in determining the effects of aging on the safety of operating nuclear plants. This paper presents the final results of a two-year research effort evaluating aging impacts on components in light water reactor systems. This research was performed as a part of the Nuclear Plant Aging Research program, sponsored by the US Nuclear Regulatory Commission. Two unique types of data analyses were performed. In the first, an aging-survey study, aging-related failure data for fifteen light water reactor systems were obtained from the Nuclear Plant Reliability Data System (NPRDS). These included safety, support, and power conversion systems. A computerized sort of these records classified each record into one of five generic categories, based on the utility's choice of the failure's NPRDS cause category. Systems and components within the systems that were most affected by aging were identified. In the second analysis, information on aging-related reported causes of failures was evaluated for component failures reported to NPRDS for auxiliary feedwater, high pressure injection, service water, and Class 1E electrical power distribution systems. 3 refs., 13 figs., 4 tabs

Failures of austenitic stainless steel components during storage: Case studies

International Nuclear Information System (INIS)

Shah, B.K.; Rastogi, P.K.; Sinha, A.K.; Kulkarni, P.G.

1993-01-01

Three studies of failures of austenitic stainless steel components during storage are described. In all cases, stress corrosion cracking was the failure mode by the action of residual stress alone. However, the source of residual stress was different for each case. Case 1 was the failure of a sample tube header for a pressurized heavy water reactor (PHWR). In Case 2, a heat exchanger shell failed during a hydrotest in a fertilizer plant. Cases concerned the cracking of type 304L plates used for spent fuel pool lining of a nuclear power station

Component failure data base of TRIGA reactors

International Nuclear Information System (INIS)

Djuricic, M.

2004-10-01

This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

Sensor Failure Detection of FASSIP System using Principal Component Analysis

Science.gov (United States)

Sudarno; Juarsa, Mulya; Santosa, Kussigit; Deswandri; Sunaryo, Geni Rina

2018-02-01

In the nuclear reactor accident of Fukushima Daiichi in Japan, the damages of core and pressure vessel were caused by the failure of its active cooling system (diesel generator was inundated by tsunami). Thus researches on passive cooling system for Nuclear Power Plant are performed to improve the safety aspects of nuclear reactors. The FASSIP system (Passive System Simulation Facility) is an installation used to study the characteristics of passive cooling systems at nuclear power plants. The accuracy of sensor measurement of FASSIP system is essential, because as the basis for determining the characteristics of a passive cooling system. In this research, a sensor failure detection method for FASSIP system is developed, so the indication of sensor failures can be detected early. The method used is Principal Component Analysis (PCA) to reduce the dimension of the sensor, with the Squarred Prediction Error (SPE) and statistic Hotteling criteria for detecting sensor failure indication. The results shows that PCA method is capable to detect the occurrence of a failure at any sensor.

Study of a simplified method of evaluating the economic maintenance importance of components in nuclear power plant system

International Nuclear Information System (INIS)

Aoki, Takayuki; Takagi, Toshiyuki; Kodama, Noriko

2014-01-01

Safety risk importance of components in nuclear power plants has been evaluated based on the probabilistic risk assessment and used for the decisions in various plant managements. But economic risk importance of the components has not been discussed very much. Therefore, this paper discusses risk importance of the components from the viewpoint of plant economic efficiency and proposes a simplified evaluation method of the economic risk importance (or economic maintenance importance). As a result of consideration, the followings were obtained. (1) A unit cost of power generation is selected as a performance indicator and can be related to a failure rate of components in nuclear power plant which is a result of maintenance. (2) The economic maintenance importance has to major factors, i.e. repair cost at component failure and production loss associated with plant outage due to component failure. (3) The developed method enables easy understanding of economic impacts of plant shutdown or power reduction due to component failures on the plane which adopts the repair cost in vertical axis and the production loss in horizontal axis. (author)

Failure data collection from a Swedish nuclear power plant

International Nuclear Information System (INIS)

Andersson, B.; Bhattacharyya, A.; Hilding, S.

1975-01-01

The Swedish nuclear utilities have formed a joint working group in the field of reliability data of thermal power plants, nuclear and fossil fuelled. The primary task of the working group is to create a standard procedure of collecting failure data from the Swedish nuclear power plants in operation. The failure data will be stored in a joint data bank. A first test collection of such data has been implemented on Oskarshamn I, and the experience with this work is discussed in this report. Reliability analysis of an engineering system is based on the availability of pertinent information on the system components. Right from the beginning within the Swedish nuclear industry the consensus has been that such data can be suitably obtained by monitoring the operating power stations. This has led to a co-operative arrangement between the vendor, ASEA-ATOM and a utility, Oskarshamnsverkets Kraftgrupp AB (OKG) to utilize information from component malfunctions in the reliability analysis. The utility prepares component failure reports which are sent to the vendor for further treatment. Experience gathered to date indicates that this arrangement is effective although many persons are involved in this process of information transmittal. The present set-up is flexible enough to accommodate necessary changes in view of problems which arise now and then in monitoring a complex system like a nuclear power station. This report briefly describes the structure of the failure data collection system. The way in which the raw data collection is done in the station by the owner and the subsequent data processing by the vendor is discussed. A brief status report of the information collected since 1971 is given. It can be concluded that valuable reliability data can be obtained by monitoring component failure reports from an operating power plant. Two requirements are, however, that all the parties involved in the arrangement follow given instructions carefully and that the assumed

Common cause failures of reactor pressure components

International Nuclear Information System (INIS)

Mankamo, T.

1978-01-01

The common cause failure is defined as a multiple failure event due to a common cause. The existence of common failure causes may ruin the potential advantages of applying redundancy for reliability improvement. Examples relevant to large mechanical components are presented. Preventive measures against common cause failures, such as physical separation, equipment diversity, quality assurance, and feedback from experience are discussed. Despite the large number of potential interdependencies, the analysis of common cause failures can be done within the framework of conventional reliability analysis, utilizing, for example, the method of deriving minimal cut sets from a system fault tree. Tools for the description and evaluation of dependencies between components are discussed: these include the model of conditional failure causes that are common to many components, and evaluation of the reliability of redundant components subjected to a common load. (author)

Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

International Nuclear Information System (INIS)

Shimada, Yoshio

2000-01-01

It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

An analysis of human maintenance failures of a nuclear power plant

International Nuclear Information System (INIS)

Pyy, P.

2000-01-01

In the report, a study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components appeared to be especially prone to human failures. Many human failures were found in safety related systems. Several failures also remained latent from outages to power operation. However, the safety significance of failures was generally small. Modifications were an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more in the future. (orig.)

Epistemic uncertainties when estimating component failure rate

International Nuclear Information System (INIS)

Jordan Cizelj, R.; Mavko, B.; Kljenak, I.

2000-01-01

A method for specific estimation of a component failure rate, based on specific quantitative and qualitative data other than component failures, was developed and is described in the proposed paper. The basis of the method is the Bayesian updating procedure. A prior distribution is selected from a generic database, whereas likelihood is built using fuzzy logic theory. With the proposed method, the component failure rate estimation is based on a much larger quantity of information compared to the presently used classical methods. Consequently, epistemic uncertainties, which are caused by lack of knowledge about a component or phenomenon are reduced. (author)

Statistical analysis of human maintenance failures of a nuclear power plant

International Nuclear Information System (INIS)

Pyy, P.

2000-01-01

In this paper, a statistical study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components are especially prone to human failures. Many human failures were found in safety related systems. Similarly, several failures remained latent from outages to power operation. The safety significance was generally small. Modifications are an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more, in future. (orig.)

Trend and pattern analysis of failures of main feedwater system components in United States commercial nuclear power plants

International Nuclear Information System (INIS)

Gentillon, C.D.; Meachum, T.R.; Brady, B.M.

1987-01-01

The goal of the trend and pattern analysis of MFW (main feedwater) component failure data is to identify component attributes that are associated with relatively high incidences of failure. Manufacturer, valve type, and pump rotational speed are examples of component attributes under study; in addition, the pattern of failures among NPP units is studied. A series of statistical methods is applied to identify trends and patterns in failures and trends in occurrences in time with regard to these component attributes or variables. This process is followed by an engineering evaluation of the statistical results. In the remainder of this paper, the characteristics of the NPRDS that facilitate its use in reliability and risk studies are highlighted, the analysis methods are briefly described, and the lessons learned thus far for improving MFW system availability and reliability are summarized (orig./GL)

Nuclear reactor component populations, reliability data bases, and their relationship to failure rate estimation and uncertainty analysis

International Nuclear Information System (INIS)

Martz, H.F.; Beckman, R.J.

1981-12-01

Probabilistic risk analyses are used to assess the risks inherent in the operation of existing and proposed nuclear power reactors. In performing such risk analyses the failure rates of various components which are used in a variety of reactor systems must be estimated. These failure rate estimates serve as input to fault trees and event trees used in the analyses. Component failure rate estimation is often based on relevant field failure data from different reliability data sources such as LERs, NPRDS, and the In-Plant Data Program. Various statistical data analysis and estimation methods have been proposed over the years to provide the required estimates of the component failure rates. This report discusses the basis and extent to which statistical methods can be used to obtain component failure rate estimates. The report is expository in nature and focuses on the general philosophical basis for such statistical methods. Various terms and concepts are defined and illustrated by means of numerous simple examples

A multi-component and multi-failure mode inspection model based on the delay time concept

International Nuclear Information System (INIS)

Wang Wenbin; Banjevic, Dragan; Pecht, Michael

2010-01-01

The delay time concept and the techniques developed for modelling and optimising plant inspection practices have been reported in many papers and case studies. For a system comprised of many components and subject to many different failure modes, one of the most convenient ways to model the inspection and failure processes is to use a stochastic point process for defect arrivals and a common delay time distribution for the duration between defect the arrival and failure of all defects. This is an approximation, but has been proven to be valid when the number of components is large. However, for a system with just a few key components and subject to few major failure modes, the approximation may be poor. In this paper, a model is developed to address this situation, where each component and failure mode is modelled individually and then pooled together to form the system inspection model. Since inspections are usually scheduled for the whole system rather than individual components, we then formulate the inspection model when the time to the next inspection from the point of a component failure renewal is random. This imposes some complication to the model, and an asymptotic solution was found. Simulation algorithms have also been proposed as a comparison to the analytical results. A numerical example is presented to demonstrate the model.

Aging effects in PWR power plants components

Energy Technology Data Exchange (ETDEWEB)

Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2013-07-01

This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

Failure Rate Prediction of Active Component Using PM Basis Database

International Nuclear Information System (INIS)

Kim, J. S.; Kim, H. W.; Park, J. S.; Jung, S. G.

2011-01-01

The safety security and efficient management of NPPs (Nuclear Power Plants) have been increased after the accident of TEPCO's Fukushima nuclear power stations. The needs for the safety and efficiency are becoming more important because about 90 percent of the NPPs all over the world are more than 20 operation years old. The preventive maintenance criteria need to be flexible, considering long-term development of the equipment performance and preventive maintenance. The PMBD (Preventive Maintenance Basis Database) program was widely used for optimization of the preventive maintenance criteria. PMBD program contains all kinds of failure mechanisms for each equipment that may occur in the power plant based on RCM(Reliability-Centered Maintenance) and numerically calculate the variation of reliability and failure rate based on PM interval changes. In this study, propriety evaluation of preventive maintenance task, cycle, technical basis for cost effective preventive maintenance strategy and an appropriate evaluation were suggested by the case application of PMBD for major components in the NPPs

Probabilistic methods in nuclear power plant component ageing analysis

International Nuclear Information System (INIS)

Simola, K.

1992-03-01

The nuclear power plant ageing research is aimed to ensure that the plant safety and reliability are maintained at a desired level through the designed, and possibly extended lifetime. In ageing studies, the reliability of components, systems and structures is evaluated taking into account the possible time- dependent decrease in reliability. The results of analyses can be used in the evaluation of the remaining lifetime of components and in the development of preventive maintenance, testing and replacement programmes. The report discusses the use of probabilistic models in the evaluations of the ageing of nuclear power plant components. The principles of nuclear power plant ageing studies are described and examples of ageing management programmes in foreign countries are given. The use of time-dependent probabilistic models to evaluate the ageing of various components and structures is described and the application of models is demonstrated with two case studies. In the case study of motor- operated closing valves the analysis are based on failure data obtained from a power plant. In the second example, the environmentally assisted crack growth is modelled with a computer code developed in United States, and the applicability of the model is evaluated on the basis of operating experience

Predictive based monitoring of nuclear plant component degradation using support vector regression

International Nuclear Information System (INIS)

Agarwal, Vivek; Alamaniotis, Miltiadis; Tsoukalas, Lefteri H.

2015-01-01

Nuclear power plants (NPPs) are large installations comprised of many active and passive assets. Degradation monitoring of all these assets is expensive (labor cost) and highly demanding task. In this paper a framework based on Support Vector Regression (SVR) for online surveillance of critical parameter degradation of NPP components is proposed. In this case, on time replacement or maintenance of components will prevent potential plant malfunctions, and reduce the overall operational cost. In the current work, we apply SVR equipped with a Gaussian kernel function to monitor components. Monitoring includes the one-step-ahead prediction of the component's respective operational quantity using the SVR model, while the SVR model is trained using a set of previous recorded degradation histories of similar components. Predictive capability of the model is evaluated upon arrival of a sensor measurement, which is compared to the component failure threshold. A maintenance decision is based on a fuzzy inference system that utilizes three parameters: (i) prediction evaluation in the previous steps, (ii) predicted value of the current step, (iii) and difference of current predicted value with components failure thresholds. The proposed framework will be tested on turbine blade degradation data.

Summary of component reliability data for probabilistic safety analysis of Korean standard nuclear power plant

International Nuclear Information System (INIS)

Choi, S. Y.; Han, S. H.

2004-01-01

The reliability data of Korean NPP that reflects the plant specific characteristics is necessary for PSA of Korean nuclear power plants. We have performed a study to develop the component reliability DB and S/W for component reliability analysis. Based on the system, we had have collected the component operation data and failure/repair data during plant operation data to 1998/2000 for YGN 3,4/UCN 3,4 respectively. Recently, we have upgraded the database by collecting additional data by 2002 for Korean standard nuclear power plants and performed component reliability analysis and Bayesian analysis again. In this paper, we supply the summary of component reliability data for probabilistic safety analysis of Korean standard nuclear power plant and describe the plant specific characteristics compared to the generic data

Failure characteristic analysis of a component on standby state

International Nuclear Information System (INIS)

Shin, Sungmin; Kang, Hyungook

2013-01-01

Periodic operations for a specific type of component, however, can accelerate aging effects which increase component unavailability. For the other type of components, the aging effect caused by operation can be ignored. Therefore frequent operations can decrease component unavailability. Thus, to get optimum unavailability proper operation period and method should be studied considering the failure characteristics of each component. The information of component failure is given according to the main causes of failure depending on time flow. However, to get the optimal unavailability, proper interval of operation for inspection should be decided considering the time dependent and independent causes together. According to this study, gradually shorter operation interval for inspection is better to get the optimal component unavailability than that of specific period

Effect of the addition of mixture of plant components on the mechanical properties of wheat bread

Science.gov (United States)

Wójcik, Monika; Dziki, Dariusz; Biernacka, Beata; Różyło, Renata; Miś, Antoni; Hassoon, Waleed H.

2017-10-01

Instrumental methods of measuring the mechanical properties of bread can be used to determine changes in the properties of it during storage, as well as to determine the effect of various additives on the bread texture. The aim of this study was to investigate the effect of the mixture of plant components on the physical properties of wheat bread. In particular, the mechanical properties of the crumb and crust were studied. A sensory evaluation of the end product was also performed. The mixture of plant components included: carob fiber, milled grain red quinoa and black oat (1:2:2) - added at 0, 5, 10, 15, 20, 25 % - into wheat flour. The results showed that the increase of the addition of the proposed additive significantly increased the water absorption of flour mixtures. Moreover, the use of the mixture of plant components above 5% resulted in the increase of bread volume and decrease of crumb density. Furthermore, the addition of the mixture of plant components significantly affected the mechanical properties of bread crumb. The hardness of crumb also decreased as a result of the mixture of plant components addition. The highest cohesiveness was obtained for bread with 10% of additive and the lowest for bread with 25% of mixture of plant components. Most importantly, the enrichment of wheat flour with the mixture of plant components significantly reduced the crust failure force and crust failure work. The results of sensory evaluation showed that the addition of the mixture of plant components of up to 10% had little effect on bread quality.

Reliability Evaluation of Machine Center Components Based on Cascading Failure Analysis

Science.gov (United States)

Zhang, Ying-Zhi; Liu, Jin-Tong; Shen, Gui-Xiang; Long, Zhe; Sun, Shu-Guang

2017-07-01

In order to rectify the problems that the component reliability model exhibits deviation, and the evaluation result is low due to the overlook of failure propagation in traditional reliability evaluation of machine center components, a new reliability evaluation method based on cascading failure analysis and the failure influenced degree assessment is proposed. A direct graph model of cascading failure among components is established according to cascading failure mechanism analysis and graph theory. The failure influenced degrees of the system components are assessed by the adjacency matrix and its transposition, combined with the Pagerank algorithm. Based on the comprehensive failure probability function and total probability formula, the inherent failure probability function is determined to realize the reliability evaluation of the system components. Finally, the method is applied to a machine center, it shows the following: 1) The reliability evaluation values of the proposed method are at least 2.5% higher than those of the traditional method; 2) The difference between the comprehensive and inherent reliability of the system component presents a positive correlation with the failure influenced degree of the system component, which provides a theoretical basis for reliability allocation of machine center system.

Assessment of ALWR passive safety system reliability. Phase 1: Methodology development and component failure quantification

International Nuclear Information System (INIS)

Hake, T.M.; Heger, A.S.

1995-04-01

Many advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive systems to perform safety functions, rather than active systems as in current reactor designs. These passive systems depend to a great extent on physical processes such as natural circulation for their driving force, and not on active components, such as pumps. An NRC-sponsored study was begun at Sandia National Laboratories to develop and implement a methodology for evaluating ALWR passive system reliability in the context of probabilistic risk assessment (PRA). This report documents the first of three phases of this study, including methodology development, system-level qualitative analysis, and sequence-level component failure quantification. The methodology developed addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. Traditional PRA methods, such as fault and event tree modeling, are applied to the component failure aspect. Thermal-hydraulic calculations are incorporated into a formal expert judgment process to address uncertainties in selected natural processes and success criteria. The first phase of the program has emphasized the component failure element of passive system reliability, rather than the natural process uncertainties. Although cursory evaluation of the natural processes has been performed as part of Phase 1, detailed assessment of these processes will take place during Phases 2 and 3 of the program

Trial application of the candidate root cause categorization scheme and preliminary assessment of selected data bases for the root causes of component failures program

International Nuclear Information System (INIS)

Bruske, S.Z.; Cadwallader, L.C.; Stepina, P.L.

1985-04-01

The objective of the Nuclear Regulatory Commission's (NRC) Root Causes of Component Failures Program is to develop and apply a categorization scheme for identifying root causes of failures for components that comprise safety and safety support systems of nuclear power plants. Results from this program will provide valuable input in the areas of probabilistic risk assessment, reliability assurance, and application of risk assessments in the inspection program. This report presents the trial application and assessment of the candidate root cause categorization scheme to three failure data bases: the In-Plant Reliability Data System (IPRDS), the Licensee Event Report (LER) data base, and the Nuclear Plant Reliability Data System (NPRDS). Results of the trial application/assessment show that significant root cause information can be obtained from these failure data bases

Failures on stainless steel components

International Nuclear Information System (INIS)

Haenninen, H.

1994-01-01

Economic losses due to failure mainly by corrosion in process and nuclear industries are considered. In these industries the characteristics of different forms of corrosion and their economic effects are fairly well known and, especially, in nuclear industry the assessment of corrosion related costs has been comprehensive. In both industries the economic losses resulting from environmentally enhanced cracking of stainless steel components and the accompanying failures and outages have been considerable, owing as much to the frequency as the unpredictability of such occurrences. (orig.)

An approach to integrating surveillance and maintenance tasks to prevent the dominant failure causes of critical components

International Nuclear Information System (INIS)

Martorell, S.; Munoz, A.; Serradell, V.

1995-01-01

Surveillance requirements and maintenance activities in a nuclear power plant aim to preserve components' inherent reliability. Up to now, predictive and preventive maintenance mainly concerned plant staff, but the US Nuclear Regulatory Commission Maintenance Rule released in July 1991 will have significant impact on how nuclear power plants perform and document this maintenance. Reliability Centered Maintenance (RCM) is a systematic methodology to establish maintenance tasks for critical components in plant with a high degree of compliance with the goals of the Rule. RCM pursues the identification of applicable and efficient tasks to prevent these components from developing their dominant failure causes, and, in turn, towards achieving proper levels of components availability with low cost. In this paper, we present an approach for identifying the most suitable set of tasks to achieve this goal, which involves the integration of maintenance activities and surveillance requirements for each critical component based on the unavailability and cost associated with each individual task which is performed on it

Recent Operating Experience involving Power Electronics Failure in Korea Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Jaedo

2015-01-01

Recently, modern power electronics devices for electrical component were steadily increased in electrical systems which used for main power control and protection. To upgrade the system reliability we recommended the redundancy for electrical equipment trip system. The past several years, Korean Nuclear power plants have changed the electrical control and protection systems (Auto Voltage Regulator, Power Protection Relay) for main generator and main power protection relay systems. In this paper we deal with operating experience involving modern solid state power electronics failure in Korean nuclear power plants. One of the failures we will discuss the degraded phenomenon of power electronics device for CEDMCS(Control Element Drive Mechanism Control System). As the result of the failure we concerned about the modification for trip source of main generator excitation systems and others. We present an interesting issue for modern solid state devices (IGBT, Thyristors). (authors)

Component aging and reliability trends in Loviisa Nuclear Power Plant

International Nuclear Information System (INIS)

Jankala, K.E.; Vaurio, J.K.

1989-01-01

A plant-specific reliability data collection and analysis system has been developed at the Loviisa Nuclear Power Plant to perform tests for component aging and analysis of reliability trends. The system yields both mean values an uncertainty distribution information for reliability parameters to be used in the PSA project underway and in living-PSA applications. Several different trend models are included in the reliability analysis system. Simple analytical expressions have been derived from the parameters of these models, and their variances have been obtained using the information matrix. This paper is focused on the details of the learning/aging models and the estimation of their parameters and statistical accuracies. Applications to the historical data of the Loviisa plant are presented. The results indicate both up- and down-trends in failure rates as well as individuality between nominally identical components

Operating experiences with passive systems and components in German nuclear power plants

International Nuclear Information System (INIS)

Maqua, M.

1996-01-01

Operating experience with passive systems and components is limited to the equipment installed in existing NPPs. In German power plants, this experience is available for equipment of the IAEA categories A, C and D. The presentation is focused on typical examples out of these three categories. An overview is given on the number of reported events and typical failure modes. Selected failures are discussed in detail. 1 ref., 6 figs, 7 tabs

Operating experiences with passive systems and components in German nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Maqua, M [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

1996-12-01

Operating experience with passive systems and components is limited to the equipment installed in existing NPPs. In German power plants, this experience is available for equipment of the IAEA categories A, C and D. The presentation is focused on typical examples out of these three categories. An overview is given on the number of reported events and typical failure modes. Selected failures are discussed in detail. 1 ref., 6 figs, 7 tabs.

Development of component failure data for seismic risk analysis

International Nuclear Information System (INIS)

Fray, R.R.; Moulia, T.A.

1981-01-01

This paper describes the quantification and utilization of seismic failure data used in the Diablo Canyon Seismic Risk Study. A single variable representation of earthquake severity that uses peak horizontal ground acceleration to characterize earthquake severity was employed. The use of a multiple variable representation would allow direct consideration of vertical accelerations and the spectral nature of earthquakes but would have added such complexity that the study would not have been feasible. Vertical accelerations and spectral nature were indirectly considered because component failure data were derived from design analyses, qualification tests and engineering judgment that did include such considerations. Two types of functions were used to describe component failure probabilities. Ramp functions were used for components, such as piping and structures, qualified by stress analysis. 'Anchor points' for ramp functions were selected by assuming a zero probability of failure at code allowable stress levels and unity probability of failure at ultimate stress levels. The accelerations corresponding to allowable and ultimate stress levels were determined by conservatively assuming a linear relationship between seismic stress and ground acceleration. Step functions were used for components, such as mechanical and electrical equipment, qualified by testing. Anchor points for step functions were selected by assuming a unity probability of failure above the qualification acceleration. (orig./HP)

Seismic fragility of nuclear power plant components. Phase I

International Nuclear Information System (INIS)

Bandyopadhyay, K.K.; Hofmayer, C.H.

1986-06-01

As part of the Component Fragility Research Program, sponsored by the US Nuclear Regulatory Commission, BNL is involved in establishing seismic fragility levels for various nuclear power plant equipment by identifying, collecting and analyzing existing test data from various sources. In Phase I of this program, BNL has reviewed approximately seventy test reports to collect fragility or high level test data for switchgears, motor control centers and similar electrical cabinets, valve actuators and numerous electrical devices of various manufacturers and models. This report provides an assessment and evaluation of the data collected in Phase I. The fragility data for medium voltage and low voltage switchgears and motor control centers are analyzed using the test response spectra (TRS) as a measure of the fragility level. The analysis reveals that fragility levels can best be described by a group of TRS curves corresponding to various failure modes. The lower-bound curve indicates the initiation of malfunctioning or structural damage; whereas, the upper-bound curve corresponds to overall failure of the equipment based on known failure modes. High level test data for some components are included in the report. These data indicate that some components are inherently strong and do not exhibit any failure mode even when tested at the vibration limit of a shake table. The common failure modes are identified in the report. The fragility levels determined in this report have been compared with those used in the PRA and Seismic Margin Studies. It appears that the BNL data better correlate with the HCLPF (High Confidence of a Low Probability of Failure) level used in Seismic Margin Studies and can improve this level as high as 60% for certain applications. Specific recommendations are provided for proper application of BNL fragility data to other studies

Nuclear plant aging research - an overview (electrical and mechanical components)

International Nuclear Information System (INIS)

Vora, J.P.

1985-01-01

As the operating nuclear power plants advance in age there must be a conscious national and international effort to understand the influence and safety implications of aging and service wear of components and structures in nuclear power plants and develop measures which are practical and cost effective for timely mitigation of aging degradation that could significantly affect plant safety. The Office of Nuclear Regulatory Research has, therefore, initiated a multi-year, multi-disciplinary program on Nuclear Plant Aging Research (NPAR). The overall goals identified for the program are as follows: 1) to identify and characterize aging and service wear effects associated with electrical and mechanical components, interfaces, and systems whose failure could impair plant safety; 2) to identify and recommend methods of inspection, surveillance and condition monitoring of electrical and mechanical components and systems which will be effective in detecting significant aging effects prior to loss of safety function so that timely maintenance and repair or replacement can be implemented; and, 3) to identify and recommend acceptable maintenance practices which can be undertaken to mitigate the effects of aging and to diminish the rate and extent of degradation caused by aging and service wear. The specific research activities to be implemented to achieve these goals are described

Estimation of component failure probability from masked binomial system testing data

International Nuclear Information System (INIS)

Tan Zhibin

2005-01-01

The component failure probability estimates from analysis of binomial system testing data are very useful because they reflect the operational failure probability of components in the field which is similar to the test environment. In practice, this type of analysis is often confounded by the problem of data masking: the status of tested components is unknown. Methods in considering this type of uncertainty are usually computationally intensive and not practical to solve the problem for complex systems. In this paper, we consider masked binomial system testing data and develop a probabilistic model to efficiently estimate component failure probabilities. In the model, all system tests are classified into test categories based on component coverage. Component coverage of test categories is modeled by a bipartite graph. Test category failure probabilities conditional on the status of covered components are defined. An EM algorithm to estimate component failure probabilities is developed based on a simple but powerful concept: equivalent failures and tests. By simulation we not only demonstrate the convergence and accuracy of the algorithm but also show that the probabilistic model is capable of analyzing systems in series, parallel and any other user defined structures. A case study illustrates an application in test case prioritization

Parameters governing the failure of steel components

International Nuclear Information System (INIS)

Schmitt, W.

1977-01-01

The most important feature of any component is the ability to carry safely the load it is designed for. The strength of the component is influenced mainly by three groups of parameters: 1. The loading of the structure; Here the possible loading cases are: normal operation, testing, emergency and faulted conditions; the kinds of loading can be divided into: internal pressure, external forces and moments, temperature loading. 2. The defects in the structure: cavities and inclusions, pores, flaws or cracks. 3. The material properties: Young's modulus, Yield - and ultimate strength, absorbed charpy energy, fracture toughness, etc. For different failure modes one has to take into account different material properties, the loading and the defects are assumed to be within certain deterministic bounds, from which deterministic safety factors can be determined with respect to any failure mode and failure criterion. However, since all parameters have a certain scatter about a mean value, there is a probability to exceed the given bounds. From the extrapolation of the distribution a value for the failure probability can be estimated. (orig.) [de

Statistical investigations of the failure behaviour of components in the AVR-experimental nuclear power plant. Vol. 1

International Nuclear Information System (INIS)

Hennings, W.

1989-08-01

From operational reports of the years 1970 to 1984, failure rates of valves in gas circuits of the AVR experimental power plant were determined. Also, potential influences of environmental and operational conditions were investigated. The resulting failure rates are for manual valves app. 0,1.10 -6 /h, for pneumatic valves between 3 and 9.10 -6 /h, for solenoid valves between 1,5 and 4.10 -6 /h and for control valves between 12 and 41.10 -6 /h. (orig.) [de

Performance Based Failure Criteria of the Base Isolation System for Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Jung Han; Kim, Min Kyu; Choi, In Kil

2013-01-01

The realistic approach to evaluate the failure state of the base isolation system is necessary. From this point of view, several concerns are reviewed and discussed in this study. This is the preliminary study for the performance based risk assessment of a base isolated nuclear power plant. The items to evaluate the capacity and response of an individual base isolator and a base isolation system were briefly outlined. However, the methodology to evaluate the realistic fragility of a base isolation system still needs to be specified. For the quantification of the seismic risk for a nuclear power plant structure, the failure probabilities of the structural component for the various seismic intensity levels need to be calculated. The failure probability is evaluated as the probability when the seismic response of a structure exceeds the failure criteria. Accordingly, the failure mode of the structural system caused by an earthquake vibration should be defined first. The type of a base isolator appropriate for a nuclear power plant structure is regarded as an elastometric rubber bearing with a lead core. The failure limit of the lead-rubber bearing (LRB) is not easy to be predicted because of its high nonlinearity and a complex loading condition by an earthquake excitation. Furthermore, the failure mode of the LRB system installed below the nuclear island cannot be simply determined because the basemat can be sufficiently supported if the number of damaged isolator is not much

A new approach for estimation of component failure rate

International Nuclear Information System (INIS)

Jordan Cizelj, R.; Kljenak, I.

1999-01-01

In the paper, a formal method for component failure rate estimation is described, which is proposed to be used for components, for which no specific numerical data necessary for probabilistic estimation exist. The framework of the method is the Bayesian updating procedure. A prior distribution is selected from a generic database, whereas the likelihood distribution is assessed from specific data on component state using principles of fuzzy logic theory. With the proposed method the component failure rate estimation is based on a much larger quantity of information compared to presently used classical methods.(author)

Common Cause Failure Analysis for the Digital Plant Protection System

International Nuclear Information System (INIS)

Kagn, Hyun Gook; Jang, Seung Cheol

2005-01-01

Safety-critical systems such as nuclear power plants adopt the multiple-redundancy design in order to reduce the risk from the single component failure. The digitalized safety-signal generation system is also designed based on the multiple-redundancy strategy which consists of more redundant components. The level of the redundant design of digital systems is usually higher than those of conventional mechanical systems. This higher redundancy would clearly reduce the risk from the single failure of components, but raise the importance of the common cause failure (CCF) analysis. This research aims to develop the practical and realistic method for modeling the CCF in digital safety-critical systems. We propose a simple and practical framework for assessing the CCF probability of digital equipment. Higher level of redundancy causes the difficulty of CCF analysis because it results in impractically large number of CCF events in the fault tree model when we use conventional CCF modeling methods. We apply the simplified alpha-factor (SAF) method to the digital system CCF analysis. The precedent study has shown that SAF method is quite realistic but simple when we consider carefully system success criteria. The first step for using the SAF method is the analysis of target system for determining the function failure cases. That is, the success criteria of the system could be derived from the target system's function and configuration. Based on this analysis, we can calculate the probability of single CCF event which represents the CCF events resulting in the system failure. In addition to the application of SAF method, in order to accommodate the other characteristics of digital technology, we develop a simple concept and several equations for practical use

Development of life evaluation technology for nuclear power plant components

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

2001-03-15

This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

Reliability Analysis of Fatigue Failure of Cast Components for Wind Turbines

OpenAIRE

Hesam Mirzaei Rafsanjani; John Dalsgaard Sørensen

2015-01-01

Fatigue failure is one of the main failure modes for wind turbine drivetrain components made of cast iron. The wind turbine drivetrain consists of a variety of heavily loaded components, like the main shaft, the main bearings, the gearbox and the generator. The failure of each component will lead to substantial economic losses such as cost of lost energy production and cost of repairs. During the design lifetime, the drivetrain components are exposed to variable loads from winds and waves an...

In-plant reliability data base for nuclear plant components: a feasibility study on human error information

International Nuclear Information System (INIS)

Borkowski, R.J.; Fragola, J.R.; Schurman, D.L.; Johnson, J.W.

1984-03-01

This report documents the procedure and final results of a feasibility study which examined the usefulness of nuclear plant maintenance work requests in the IPRDS as tools for understanding human error and its influence on component failure and repair. Developed in this study were (1) a set of criteria for judging the quality of a plant maintenance record set for studying human error; (2) a scheme for identifying human errors in the maintenance records; and (3) two taxonomies (engineering-based and psychology-based) for categorizing and coding human error-related events

Dependent failure analysis of NPP data bases

International Nuclear Information System (INIS)

Cooper, S.E.; Lofgren, E.V.; Samanta, P.K.; Wong Seemeng

1993-01-01

A technical approach for analyzing plant-specific data bases for vulnerabilities to dependent failures has been developed and applied. Since the focus of this work is to aid in the formulation of defenses to dependent failures, rather than to quantify dependent failure probabilities, the approach of this analysis is critically different. For instance, the determination of component failure dependencies has been based upon identical failure mechanisms related to component piecepart failures, rather than failure modes. Also, component failures involving all types of component function loss (e.g., catastrophic, degraded, incipient) are equally important to the predictive purposes of dependent failure defense development. Consequently, dependent component failures are identified with a different dependent failure definition which uses a component failure mechanism categorization scheme in this study. In this context, clusters of component failures which satisfy the revised dependent failure definition are termed common failure mechanism (CFM) events. Motor-operated valves (MOVs) in two nuclear power plant data bases have been analyzed with this approach. The analysis results include seven different failure mechanism categories; identified potential CFM events; an assessment of the risk-significance of the potential CFM events using existing probabilistic risk assessments (PRAs); and postulated defenses to the identified potential CFM events. (orig.)

Corrosion-related failures in power plant condensers. Final report

International Nuclear Information System (INIS)

Beavers, J.A.; Agrawal, A.K.; Berry, W.E.

1980-08-01

A survey of the literature has been conducted for the Electric Power Research Institute on corrosion failures in surface condensers. The survey was directed toward condenser failures in pressurized water reactor (PWR) power plants but includes pertinent literature related to fossil and to other nuclear power plants. It includes literature on reported service failures and on experimental studies that impact on these failures

Estimation of the common cause failure probabilities on the component group with mixed testing scheme

International Nuclear Information System (INIS)

Hwang, Meejeong; Kang, Dae Il

2011-01-01

Highlights: ► This paper presents a method to estimate the common cause failure probabilities on the common cause component group with mixed testing schemes. ► The CCF probabilities are dependent on the testing schemes such as staggered testing or non-staggered testing. ► There are many CCCGs with specific mixed testing schemes in real plant operation. ► Therefore, a general formula which is applicable to both alternate periodic testing scheme and train level mixed testing scheme was derived. - Abstract: This paper presents a method to estimate the common cause failure (CCF) probabilities on the common cause component group (CCCG) with mixed testing schemes such as the train level mixed testing scheme or the alternate periodic testing scheme. In the train level mixed testing scheme, the components are tested in a non-staggered way within the same train, but the components are tested in a staggered way between the trains. The alternate periodic testing scheme indicates that all components in the same CCCG are tested in a non-staggered way during the planned maintenance period, but they are tested in a staggered way during normal plant operation. Since the CCF probabilities are dependent on the testing schemes such as staggered testing or non-staggered testing, CCF estimators have two kinds of formulas in accordance with the testing schemes. Thus, there are general formulas to estimate the CCF probability on the staggered testing scheme and non-staggered testing scheme. However, in real plant operation, there are many CCCGs with specific mixed testing schemes. Recently, Barros () and Kang () proposed a CCF factor estimation method to reflect the alternate periodic testing scheme and the train level mixed testing scheme. In this paper, a general formula which is applicable to both the alternate periodic testing scheme and the train level mixed testing scheme was derived.

Incorporation of passive components aging into PRAs

International Nuclear Information System (INIS)

Phillips, J.H.; Roesener, W.S.; Magleby, H.L.; Geidl, V.

1993-01-01

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants. (orig.)

Regression to fuzziness method for estimation of remaining useful life in power plant components

Science.gov (United States)

Alamaniotis, Miltiadis; Grelle, Austin; Tsoukalas, Lefteri H.

2014-10-01

Mitigation of severe accidents in power plants requires the reliable operation of all systems and the on-time replacement of mechanical components. Therefore, the continuous surveillance of power systems is a crucial concern for the overall safety, cost control, and on-time maintenance of a power plant. In this paper a methodology called regression to fuzziness is presented that estimates the remaining useful life (RUL) of power plant components. The RUL is defined as the difference between the time that a measurement was taken and the estimated failure time of that component. The methodology aims to compensate for a potential lack of historical data by modeling an expert's operational experience and expertise applied to the system. It initially identifies critical degradation parameters and their associated value range. Once completed, the operator's experience is modeled through fuzzy sets which span the entire parameter range. This model is then synergistically used with linear regression and a component's failure point to estimate the RUL. The proposed methodology is tested on estimating the RUL of a turbine (the basic electrical generating component of a power plant) in three different cases. Results demonstrate the benefits of the methodology for components for which operational data is not readily available and emphasize the significance of the selection of fuzzy sets and the effect of knowledge representation on the predicted output. To verify the effectiveness of the methodology, it was benchmarked against the data-based simple linear regression model used for predictions which was shown to perform equal or worse than the presented methodology. Furthermore, methodology comparison highlighted the improvement in estimation offered by the adoption of appropriate of fuzzy sets for parameter representation.

Data book of the component failure rate stored in the RECORD

International Nuclear Information System (INIS)

Oikawa, Testukuni; Sasaki, Shinobu; Hikawa, Michihiro; Higuchi, Suminori.

1989-04-01

The Japan Atomic Energy Research Insitute (JAERI) has developed a computerized component reliability data base and its retrieval system, RECORD, on collected failure rates from published literatures in order to promote convenience and efficiency of systems reliability analysis in the PSA (Probabilistic Safety Assessment). In order to represent collected failure rates in a uniform format, codes are defined for component category, failure mode, data source, unit of failure rate and statistocal parameter. Up to now, approximately 11,500 pieces of component failure rate data from about 35 open literatures have been stored in the RECORD. This report provides the failure rate stored in the RECORD data base for the usage by systems analysts, as well as brief descriptions about the data base structure and how to use this data book. (author)

Operating Experience Insights into Pipe Failures for Electro-Hydraulic Control and Instrument Air Systems in Nuclear Power Plant. A Topical Report from the Component Operational Experience, Degradation and Ageing Programme

International Nuclear Information System (INIS)

2015-01-01

Structural integrity of piping systems is important for plant safety and operability. In recognition of this, information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organisations (e.g. OECD/NEA and IAEA) and industry organisations worldwide to provide systematic feedback for example to reactor regulation and research and development programmes associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programmes, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. Several OECD member countries have agreed to establish the OECD/NEA 'Component Operational Experience, Degradation and Ageing Programme' (CODAP) to encourage multilateral co-operation in the collection and analysis of data relating to degradation and failure of metallic piping and non-piping metallic passive components in commercial nuclear power plants. The scope of the data collection includes service-induced wall thinning, part through-wall cracks, through-wall cracks with and without active leakage, and instances of significant degradation of metallic passive components, including piping pressure boundary integrity. The OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) acts as an umbrella committee of the Project. CODAP is the continuation of the 2002-2011 'OECD/NEA Pipe Failure Data Exchange Project' (OPDE) and the Stress Corrosion Cracking Working Group of the 2006-2010 'OECD/NEA Stress Corrosion Cracking and Cable Ageing Project' (SCAP). OPDE was formally launched in May 2002. Upon completion of the third term (May 2011), the OPDE project was officially closed to be succeeded by CODAP. SCAP was enabled by a voluntary contribution from Japan. It was formally launched in June 2006 and officially closed with an international workshop held in Tokyo in May

Operational failure at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Szatmary, Z.

2003-01-01

NPP failures are ranked according to the International Nuclear Event Scale. To rank the failure first a presentation of the pressurized water plant is given, including fuel change, maintenance cleaning and decontamination process. The failure has been produced with fuel bars in the cleaning container. Consequences of the failure are small, negligible environmental pollution with radioactive material and significant financial outfall due to inactivity of block 2. Among the causes of the failure are design errors of the cleaning container, the pure chemical approach to cleaning, unknown risk factors for some of the cleaning staff, cleaning container has not been verified and approved by responsible authorities, the prevalence of economic and quantitative indicators of the plant on the detriment of safety. Organisational factors also contribute to the possibility of nuclear failures. Specialist training in Germany (where the container has been produced) is significantly reduced, while in Hungary the political tide has caused a permanent change in the higher echelons of the plant management, where nuclear specialists were not included. (Gy.M.)

Development of the DQFM method to consider the effect of correlation of component failures in seismic PSA of nuclear power plant

International Nuclear Information System (INIS)

Watanabe, Yuichi; Oikawa, Tetsukuni; Muramatsu, Ken

2003-01-01

This paper presents a new calculation method for considering the effect of correlation of component failures in seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs) by direct quantification of Fault Tree (FT) using the Monte Carlo simulation (DQFM) and discusses the effect of correlation on core damage frequency (CDF). In the DQFM method, occurrence probability of a top event is calculated as follows: (1) Response and capacity of each component are generated according to their probability distribution. In this step, the response and capacity can be made correlated according to a set of arbitrarily given correlation data. (2) For each component whether the component is failed or not is judged by comparing the response and the capacity. (3) The status of each component, failure or success, is assigned as either TRUE or FALSE in a Truth Table, which represents the logical structure of the FT to judge the occurrence of the top event. After this trial is iterated sufficient times, the occurrence probability of the top event is obtained as the ratio of the occurrence number of the top event to the number of total iterations. The DQFM method has the following features compared with the minimal cut set (MCS) method used in the well known Seismic Safety Margins Research Program (SSMRP). While the MCS method gives the upper bound approximation for occurrence probability of an union of MCSs, the DQFM method gives more exact results than the upper bound approximation. Further, the DQFM method considers the effect of correlation on the union and intersection of component failures while the MCS method considers only the effect on the latter. The importance of these features in seismic PSA of NPPs are demonstrated by an example calculation and a calculation of CDF in a seismic PSA. The effect of correlation on CDF was evaluated by the DQFM method and was compared with that evaluated in the application study of the SSMRP methodology. In the application

Automated derivation of failure symptoms for diagnosis of nuclear plant

International Nuclear Information System (INIS)

Washio, T.; Kitamura, M.; Kotajima, K.; Sugiyama, K.

1986-01-01

A method of automated derivation of failure symptoms was developed as an approach to computer-aided failure diagnosis in a nuclear power plant. The automated derivation is realized using a knowledge representation called the semantic network (S-net). The purpose of this paper is to demonstrate the applicability of the S-net representation as a basic tool for deriving failure symptoms. If one can generate symptoms automatically, the computer-aided plant safety analysis and diagnosis can be performed easily by evaluating the influence of the failures on the whole plant. A specific description format called a 'network list' was introduced to implement the knowledge of the structure of the plant. The failure symptoms are derived automatically, based on the knowledge of the structure of the plant, using a PROLOG-based database handling system. This approach allows us to derive the failure symptoms of the plant without using conventional event-chain models (e.g. a cause-consequence tree) which are subject to human errors in their design and implementation. Applicability of this method was evaluated with a simulation model of the dynamics of the secondary system of a PWR. (author)

Reliability Analysis of Fatigue Failure of Cast Components for Wind Turbines

Directory of Open Access Journals (Sweden)

Hesam Mirzaei Rafsanjani

2015-04-01

Full Text Available Fatigue failure is one of the main failure modes for wind turbine drivetrain components made of cast iron. The wind turbine drivetrain consists of a variety of heavily loaded components, like the main shaft, the main bearings, the gearbox and the generator. The failure of each component will lead to substantial economic losses such as cost of lost energy production and cost of repairs. During the design lifetime, the drivetrain components are exposed to variable loads from winds and waves and other sources of loads that are uncertain and have to be modeled as stochastic variables. The types of loads are different for offshore and onshore wind turbines. Moreover, uncertainties about the fatigue strength play an important role in modeling and assessment of the reliability of the components. In this paper, a generic stochastic model for fatigue failure of cast iron components based on fatigue test data and a limit state equation for fatigue failure based on the SN-curve approach and Miner’s rule is presented. The statistical analysis of the fatigue data is performed using the Maximum Likelihood Method which also gives an estimate of the statistical uncertainties. Finally, illustrative examples are presented with reliability analyses depending on various stochastic models and partial safety factors.

The right maintenance on the right components, at the right time, with the right parts: maintaining high plant reliability through an effective maintenance program

International Nuclear Information System (INIS)

Von Hatten, P.

2008-01-01

The objective of the maintenance program at a Nuclear Power Plant is to be proactive and prevent unexpected failures of equipment that can impact on Nuclear or Conventional Safety and Plant Production. This does not mean that all equipment failures will be prevented; in a number of cases the most cost effective solution is to allow equipment to run to failure. Deciding what components are critical to the plant is the first step. The industry uses guidance from INPO Advanced Process, AP913, to classify components as Critical, Non Critical or Run to Failure based on the consequence of the failure. Once this is complete, then the right maintenance program needs to be specified. This is done through utilization of experience from the industry based on the type of component. Maintenance strategies and templates have been produced for most power plant components. Each station or fleet needs then to apply the criteria, with exceptions as required, to determine the maintenance requirements and frequency for their components. This includes predictive and preventative maintenance. The more critical the component is the more rigorous the maintenance requirements. Once the maintenance program is defined it can be implemented. This requires that the Preventative Maintenance (PM's) are updated to ensure the correct tasks are in place and the frequency is correct. Work Management will group the PM's so they can scheduled efficiently and to minimize equipment down time. The last element is to ensure that the required parts are specified and are stocked or readily available for the maintenance when it is scheduled. This is an ongoing effort since components become obsolete or suppliers go out of business or change hands. (author)

Future needs for inelastic analysis in design of high-temperature nuclear plant components

International Nuclear Information System (INIS)

Corum, J.M.

1980-01-01

The role that inelastic analyses play in the design of high-temperature nuclear plant components is described. The design methodology, which explicitly accounts for nonlinear material deformation and time-dependent failure modes, requires a significant level of realism in the prediction of structural response. Thus, material deformation and failure modeling are, along with computational procedures, key parts of the methodology. Each of these is briefly discussed along with validation by comparisons with benchmark structural tests, and problem areas and needs are discussed for each

Development of IPRO-ZONE to Determine Component Failure Modes Affected by a Fire Event

International Nuclear Information System (INIS)

Kang, Dae Il; Han, Sang Hoon

2010-01-01

A Fire PSA requires a PSA analyst to select internal initiating events and to determine component failure modes for fire occurrence event of each fire compartment. The component failure modes caused by a fire depend on the several factors. These factors are whether components and their relating equipment and cables are located at fire initiation and propagation compartments or not, fire effects on control and power cables for components and their relating equipment, designed failure modes of component, success criteria in a PSA model, etc. Up to the present, a PSA analyst has been manually determining component failure modes based on criteria mentioned above. This task is one of the difficult works required for fire PSA expertise. In addition, since it requires much information, a fire PSA analyst may have difficulty in maintaining consistency for determining the component failure modes and documentation for them. After determining the component failure modes, internal PSA basic events corresponding to the component failure modes are selected and fire events are modeled for the selected basic events if required. KAERI has been developing the IPRO-ZONE (interface program for constructing zone effect table) to determine component failure modes affected by a fire, to select the internal PSA basic events, and to generate fire events to be modeled. In this paper, we introduce the overview of the IPRO-ZONE and approaches for determining component failure modes implemented in the IPRO-ZONE

Development of a web-based fatigue life evaluation system for primary components in a nuclear power plant

International Nuclear Information System (INIS)

Seo, Hyong Won; Lee, Sang Min; Choi, Jae Boong; Kim, Young Jin; Choi, Sung Nam; Jang, Ki Sang; Hong, Sung Yull

2004-01-01

A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including regular in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage the integrity issues on a nuclear power plant. In this paper, a web-based fatigue life evaluation system for primary components in nuclear power plant is proposed. This system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant

Analysis of reactor trips involving balance-of-plant failures

International Nuclear Information System (INIS)

Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

1986-01-01

The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

Analysis and prevention of human failure in nuclear power plants

International Nuclear Information System (INIS)

Liu Xinshuan

2001-01-01

Based on the performances in Daya Bay Nuclear Power Plant and the common experience from the world nuclear industry, the features and usual kinds of human failures in nuclear power plants are highlighted and the prominent factors on the personal, external and decision problems which might cause the human failures are analyzed. Effective preventive measures have been proposed respectively. Some successful human-failure-prevention practices applied in the Daya Bay Nuclear Power Plant are illustrated specifically

Trend analysis of cables failure events at nuclear power plants

International Nuclear Information System (INIS)

Fushimi, Yasuyuki

2007-01-01

In this study, 152 failure events related with cables at overseas nuclear power plants are selected from Nuclear Information Database, which is owned by The Institute of Nuclear Safety System, and these events are analyzed in view of occurrence, causal factor, and so on. And 15 failure events related with cables at domestic nuclear power plants are selected from Nuclear Information Archives, which is owned by JANTI, and these events are analyzed by the same manner. As a result of comparing both trends, it is revealed following; 1) A cable insulator failure rate is lower at domestic nuclear power plants than at foreign ones. It is thought that a deterioration diagnosis is performed broadly in Japan. 2) Many buried cables failure events have been occupied a significant portion of cables failure events during work activity at overseas plants, however none has been occurred at domestic plants. It is thought that sufficient survey is conducted before excavating activity in Japan. 3) A domestic age related cables failure rate in service is lower than the overseas one and domestic improper maintenance rate is higher than the overseas one. Maintenance worker' a skill improvement is expected in order to reduce improper maintenance. (author)

Development status of component reliability database for Korean NPPs and a case study

International Nuclear Information System (INIS)

Choi, S. Y.; Yang, S. H.; Lee, S. C.; Kim, S. H.; Han, S. H.

1999-01-01

We have applied a generic database to the PSA (Probabilistic Safety Assessment) for the Korean Standard NPPs (Nuclear Power Plant) since there is no specific component reliability database. However generic data is not enough to reflect the specific characteristics of domestic plants since it is collected by foreign plants. Therefore we are developing the plant-specific component reliability database for domestic NPPs. In this paper, we describe the development status of the component reliability database and the approach method of data collection and component failure analysis. We also summarize a case study of component failure analysis. We first collect the failure and repair data from the TR (Trouble Report) electronic database and the daily operation report sheet. Now we add a data collection method that checks the original TR sheet to improve the data quality. We input the component failure and repair data of principal components of about 30 systems into the component reliability database. Now, we are analyzing the component failure data of 11 safety systems among the systems to calculate component failure rate and unavailability etc

Failure analysis a practical guide for manufacturers of electronic components and systems

CERN Document Server

Bâzu, Marius

2011-01-01

Failure analysis is the preferred method to investigate product or process reliability and to ensure optimum performance of electrical components and systems. The physics-of-failure approach is the only internationally accepted solution for continuously improving the reliability of materials, devices and processes. The models have been developed from the physical and chemical phenomena that are responsible for degradation or failure of electronic components and materials and now replace popular distribution models for failure mechanisms such as Weibull or lognormal. Reliability engineers nee

The maintenance optimization of structural components in nuclear power plants

International Nuclear Information System (INIS)

Bryla, P.; Ardorino, F.; Aufort, P.; Jacquot, J.P.; Magne, L.; Pitner, P.; Verite, B.; Villain, B.; Monnier, B.

1997-10-01

An optimization process, called 'OMF-Structures', is developed by Electricite de France (EDF) in order to extend the current 'OMF' Reliability Centered Maintenance to piping structural components. The Auxiliary Feedwater System of a 900 MW French nuclear plant has been studied in order to lay the foundations of the method. This paper presents the currently proposed principles of the process. The principles of the OMF-Structures process include 'Risk-Based Inspection' concepts within an RCM process. Two main phases are identified: The purpose of the first phase is to select the risk-significant failure modes and associated elements. This phase consists of two major steps: potential consequences evaluation and reliability performance evaluation. The second phase consists of the definition of preventive maintenance programs for piping elements that are associated with risk-significant failure modes. (author)

Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

International Nuclear Information System (INIS)

Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O.

2016-01-01

During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O., E-mail: vasconv@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2016-07-01

During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

Energy Technology Data Exchange (ETDEWEB)

Fok, Alex

2013-10-30

The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the model to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.

Tritium Waste Treatment System component failure data analysis from June 18, 1984--December 31, 1989

International Nuclear Information System (INIS)

Cadwallader, L.C.; Stolpe Gavett, M.A.

1990-09-01

This document gives the failure rates for the major tritium-bearing components in the Tritium Waste Treatment System at the Tritium Systems Test Assembly, which is a fusion research and technology facility at the Los Alamos National Laboratory. The failure reports, component populations, and operating demands/hours are given in this report, and sample calculations for binomial demand failure rates and poisson hourly failure rates are given in the appendices. The failure rates for tritium-bearing components were on the order of the screening failure rate values suggested for fusion reliability and risk analyses. More effort should be directed toward collecting and analyzing fusion component failure data, since accurate failure rates are necessary to refine reliability and risk analyses. 15 refs., 4 figs., 4 tabs

Reliability of piping system components. Volume 4: The pipe failure event database

Energy Technology Data Exchange (ETDEWEB)

Nyman, R; Erixon, S [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B [RSA Technologies, Visat, CA (United States)

1996-07-01

Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A `data driven and systems oriented` analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs.

Reliability of piping system components. Volume 4: The pipe failure event database

International Nuclear Information System (INIS)

Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

1996-07-01

Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A 'data driven and systems oriented' analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs

Parametric Study on Ultimate Failure Criteria of Elbow Piping Components in Seismically Isolated NPP

Energy Technology Data Exchange (ETDEWEB)

Hahm, Dae Gi; Ki, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

2016-05-15

It is well known that the interface pipes between isolated and non-isolated structures will become the most critical in the seismically isolated NPPs. Therefore, seismic performance of such interface pipes should be evaluated comprehensively especially in terms of the seismic fragility capacity. To evaluate the seismic capacity of interface pipes in the isolated NPP, firstly, we should define the failure mode and failure criteria of critical pipe components. Hence, in this study, we performed the dynamic tests of elbow components which were installed in a seismically isolated NPP, and evaluated the ultimate failure mode and failure criteria by using the test results. To do this, we manufactured 25 critical elbow component specimens and performed cyclic loading tests under the internal pressure condition. The failure mode and failure criteria of a pipe component will be varied by the design parameters such as the internal pressure, pipe diameter, loading type, and loading amplitude. From the tests, we assessed the effects of the variation parameters onto the failure criteria. For the tests, we generated the seismic input protocol of relative displacement between the ends of elbow component. In this paper, elbow in piping system was defined as a fragile element and numerical model was updated by component test. Failure mode of piping component under seismic load was defined by the dynamic tests of ultimate pipe capacity. For the interface piping system, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system. In this study, the dynamic tests were performed for the elbow components which were installed in an actual NPPs, and the ultimate failure mode and failure criteria were also evaluated by using the test results.

Failure cause analysis and improvement for magnetic component cabinet

International Nuclear Information System (INIS)

Ge Bing

1999-01-01

The magnetic component cabinet is an important thermal control device fitted on the nuclear power. Because it used a self-saturation amplifier as a primary component, the magnetic component cabinet has some boundness. For increasing the operation safety on the nuclear power, the author describes a new scheme. In order that the magnetic component cabinet can be replaced, the new type component cabinet is developed. Integrate circuit will replace the magnetic components of every function parts. The author has analyzed overall failure cause for magnetic component cabinet and adopted some measures

Ranking of risk significant components for the Davis-Besse Component Cooling Water System

International Nuclear Information System (INIS)

Seniuk, P.J.

1994-01-01

Utilities that run nuclear power plants are responsible for testing pumps and valves, as specified by the American Society of Mechanical Engineers (ASME) that are required for safe shutdown, mitigating the consequences of an accident, and maintaining the plant in a safe condition. These inservice components are tested according to ASME Codes, either the earlier requirements of the ASME Boiler and Pressure Vessel Code, Section XI, or the more recent requirements of the ASME Operation and Maintenance Code, Section IST. These codes dictate test techniques and frequencies regardless of the component failure rate or significance of failure consequences. A probabilistic risk assessment or probabilistic safety assessment may be used to evaluate the component importance for inservice test (IST) risk ranking, which is a combination of failure rate and failure consequences. Resources for component testing during the normal quarterly verification test or postmaintenance test are expensive. Normal quarterly testing may cause component unavailability. Outage testing may increase outage cost with no real benefit. This paper identifies the importance ranking of risk significant components in the Davis-Besse component cooling water system. Identifying the ranking of these risk significant IST components adds technical insight for developing the appropriate test technique and test frequency

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. NDI is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI methods are reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components. Among these can be highlighted Failure Modes and Effects Analysis (FMEA) and THERP (Technique for Human Error Rate Prediction). The application of these techniques is illustrated in an example of qualitative and quantitative studies to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues. (author)

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano

2017-01-01

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. NDI is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI methods are reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components. Among these can be highlighted Failure Modes and Effects Analysis (FMEA) and THERP (Technique for Human Error Rate Prediction). The application of these techniques is illustrated in an example of qualitative and quantitative studies to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues. (author)

Evaluation of Risk Metrics for KHNP Reference Plants Using the Latest Plant Specific Data

International Nuclear Information System (INIS)

Jeon, Ho Jun; Hwang, Seok Won; Ghi, Moon Goo

2010-01-01

As Risk-Informed Applications (RIAs) are actively implemented in the nuclear industry, an issue associated with the technical adequacy of the Probabilistic Safety Assessment (PSA) arises in its data sources. The American Society of Mechanical Engineers (ASME) PRA standard suggests the use of component failure data that represent the as-built and as-operated plant conditions. Furthermore, the peer reviews for the KHNP reference plants stated that the component failure data should be updated to reflect the latest plant specific data available. For ensuring the technical adequacy in PSA data elements, we try to update component failure data to reflect the as-operated plant conditions, and a trend analysis of the failure data is implemented. In addition, by applying the updated failure data to the PSA models of the KHNP reference plants, the risk metrics of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) are evaluated

Component reliability analysis for development of component reliability DB of Korean standard NPPs

International Nuclear Information System (INIS)

Choi, S. Y.; Han, S. H.; Kim, S. H.

2002-01-01

The reliability data of Korean NPP that reflects the plant specific characteristics is necessary for PSA and Risk Informed Application. We have performed a project to develop the component reliability DB and calculate the component reliability such as failure rate and unavailability. We have collected the component operation data and failure/repair data of Korean standard NPPs. We have analyzed failure data by developing a data analysis method which incorporates the domestic data situation. And then we have compared the reliability results with the generic data for the foreign NPPs

Components selection for ageing management

International Nuclear Information System (INIS)

Mingiuc, C.; Vidican, D.

2002-01-01

Full text: The paper presents a synthesis of methods and activities realized for the selection of critical components to assure plant safety and availability (as electricity supplier). There are presented main criteria for selection, screening process. For the resulted categories of components shall be applied different category of maintenance (condition oriented, scheduled or corrective), function of the importance and financial effort necessary to fulfil the task. 1. Systems and components screening for plant safety assurance For the systems selection, from Safety point of view, was necessary first, to define systems which are dangerous in case of failure (mainly by rupture/ release of radioactivity) and the safety systems which have to mitigate the effects. This is realized based on accident analysis (from Safety Report). Also where taken in to account the 4 basic Safety Principles: 'Reactor shut down; Residual heat removal; Radioactivity products confinement; NPP status monitoring in normal and accident conditions'. Following step is to establish safety support systems, which have to action to assure main safety systems operation. This could be realized based on engineering judgement, or on PSA Level I analysis. Finally shall be realized chains of the support systems, which have to work, till primary systems. For the critical components selection, was realized a Failure Mode and Effect Analysis (FMEA), considering the components effects of failures, on system safety function. 2. Systems and components screening for plant availability assurance The work was realized in two steps: Systems screening; Components screening The systems screening, included: General, analyze of the plant systems list and the definition of those which clearly have to run continue to assure the nominal power; Realization of a complex diagram to define interdependence between the systems (e.g. PHT and auxiliaries, moderator and auxiliaries, plant electrical diagram); Fill of special

Failure diagnosis aiding device for plant equipment

International Nuclear Information System (INIS)

Uhara, Yoshihiko.

1990-01-01

The present invention intends to improve the efficiency of trouble shooting for equipments of industrial plants such as nuclear power plants. The device of the present invention comprises an intelligence base and an inference mechanism base. The intelligence base comprises a rule base, an information storing section having a part frame and a working frame and a user's frame. The parts frame contains the failure rate on every parts and data on related operations. The working frame contains the importance and frequency of working. The user's frame contains parameters showing the extent of user's skills. The rule base, the parts frame and the working frame can be selected in accordance with the extent of the user's skill in the inference mechanism. With such a constitution, failures can be checked with the intelligence base in accordance with the knowledges for the failures of the equipments and the extent of user's skill by way of the inference mechanism. (I.S.)

Case study on the use of PSA methods: Determining safety importance of systems and components at nuclear power plants

International Nuclear Information System (INIS)

1991-04-01

This case study emphasizes the step of probabilistic safety assessment (PSA) regarding identification of systems and components important to nuclear plant safety. An importance analysis involves combining information that is both qualitative and probabilistic in nature to generate a numerical ranking to determine the system and/or component failures that dominate the risk. Such a ranking can suggest where hardware, software, human factors and component design changes can be implemented to improve plant safety. Examples of using ranking methodology are described. A qualitative ranking criteria is discussed for components and systems that are not included in a PSA. 18 refs, 7 figs, 18 tabs

Statistical analysis of nuclear power plant pump failure rate variability: some preliminary results

International Nuclear Information System (INIS)

Martz, H.F.; Whiteman, D.E.

1984-02-01

In-Plant Reliability Data System (IPRDS) pump failure data on over 60 selected pumps in four nuclear power plants are statistically analyzed using the Failure Rate Analysis Code (FRAC). A major purpose of the analysis is to determine which environmental, system, and operating factors adequately explain the variability in the failure data. Catastrophic, degraded, and incipient failure severity categories are considered for both demand-related and time-dependent failures. For catastrophic demand-related pump failures, the variability is explained by the following factors listed in their order of importance: system application, pump driver, operating mode, reactor type, pump type, and unidentified plant-specific influences. Quantitative failure rate adjustments are provided for the effects of these factors. In the case of catastrophic time-dependent pump failures, the failure rate variability is explained by three factors: reactor type, pump driver, and unidentified plant-specific influences. Finally, point and confidence interval failure rate estimates are provided for each selected pump by considering the influential factors. Both types of estimates represent an improvement over the estimates computed exclusively from the data on each pump

FRAC (failure rate analysis code): a computer program for analysis of variance of failure rates. An application user's guide

International Nuclear Information System (INIS)

Martz, H.F.; Beckman, R.J.; McInteer, C.R.

1982-03-01

Probabilistic risk assessments (PRAs) require estimates of the failure rates of various components whose failure modes appear in the event and fault trees used to quantify accident sequences. Several reliability data bases have been designed for use in providing the necessary reliability data to be used in constructing these estimates. In the nuclear industry, the Nuclear Plant Reliability Data System (NPRDS) and the In-Plant Reliability Data System (IRPDS), among others, were designed for this purpose. An important characteristic of such data bases is the selection and identification of numerous factors used to classify each component that is reported and the subsequent failures of each component. However, the presence of such factors often complicates the analysis of reliability data in the sense that it is inappropriate to group (that is, pool) data for those combinations of factors that yield significantly different failure rate values. These types of data can be analyzed by analysis of variance. FRAC (Failure Rate Analysis Code) is a computer code that performs an analysis of variance of failure rates. In addition, FRAC provides failure rate estimates

Determination of Component Failure Modes for a Fire PSA by Using Decision Trees

International Nuclear Information System (INIS)

Kang, Dae Il; Han, Sang Hoon; Lim, Jae Won

2007-01-01

KAERI developed the method, called a mapping technique, for the quantification of external events PSA models with one top model for an internal events PSA. The mapping technique can be implemented by the construction of mapping tables. The mapping tables include initiating events and transfer events of fire, and internal PSA basic events affected by a fire. This year, KAERI is making mapping tables for the one top model for Ulchin Unit 3 and 4 fire PSA with previously conducted Fire PSA results for Ulchin Unit 3 and 4. A Fire PSA requires a PSA analyst to determine component failure modes affected by a fire. The component failure modes caused by a fire depend on several factors. These several factors are whether components are located at fire initiation and propagation areas or not, fire effects on control and power cables for components, designed failure modes of components, success criteria in a PSA model, etc. Thus, it is not easy to manually determine component failure modes caused by a fire. In this paper, we propose the use of decision trees for the determination of component failure modes affected by a fire and the selection of internal PSA basic events. Section 2 presents the procedure for previously performed the Ulchin Unit 3 and 4 fire PSA and mapping technique. Section 3 presents the process for identification of basic events and decision trees. Section 4 presents the concluding remarks

Using failure mode and effect analysis in identification of components sensitive to ageing

International Nuclear Information System (INIS)

Nitoi, Mirela; Turcu, Ilie; Apostol, Minodora; Farcasiu, Mita; Popa, Adrian; Florescu, Gheorghe; Pavelescu, Margarit

2008-01-01

Ageing represents a phenomenon of concern since any degradation that may occur in time could lower a component performance and so reduce its reliability. If the phenomenon is left unchecked and unmitigated, the ageing could increase the risk associated with the facility operation. To understand the ageing degradation of a component, it is first necessary to identify and understand the ageing processes. Since these processes involve constituent materials, parts and the service conditions of components, it is necessary to know the design, materials, service conditions, performance requirements, operating experience (operation, surveillance and maintenance histories) and relevant research results for the component of interest. The purpose of the Ageing Failure Mode and Effect Analysis (AFMEA) is to study the results or effects of item failure caused by ageing, on system operation and to classify each potential failure according to its severity The paper will present the advantages of using AFMEA in identification of most sensitive to ageing components, as the results obtained for a particular case. For each component analyzed, the stressors will be established, the corresponding ageing mechanisms will be identified, as the failure modes induced by the ageing mechanisms. (authors)

Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

Energy Technology Data Exchange (ETDEWEB)

Duffy, Stephen [Cleveland State Univ., Cleveland, OH (United States)

2013-09-09

This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.

Modeling Stress Strain Relationships and Predicting Failure Probabilities For Graphite Core Components

International Nuclear Information System (INIS)

Duffy, Stephen

2013-01-01

This project will implement inelastic constitutive models that will yield the requisite stress-strain information necessary for graphite component design. Accurate knowledge of stress states (both elastic and inelastic) is required to assess how close a nuclear core component is to failure. Strain states are needed to assess deformations in order to ascertain serviceability issues relating to failure, e.g., whether too much shrinkage has taken place for the core to function properly. Failure probabilities, as opposed to safety factors, are required in order to capture the bariability in failure strength in tensile regimes. The current stress state is used to predict the probability of failure. Stochastic failure models will be developed that can accommodate possible material anisotropy. This work will also model material damage (i.e., degradation of mechanical properties) due to radiation exposure. The team will design tools for components fabricated from nuclear graphite. These tools must readily interact with finite element software--in particular, COMSOL, the software algorithm currently being utilized by the Idaho National Laboratory. For the eleastic response of graphite, the team will adopt anisotropic stress-strain relationships available in COMSO. Data from the literature will be utilized to characterize the appropriate elastic material constants.

Structural integrity of stainless steel components exposed to neutron irradiation. Change in failure strength of cracked components due to cold working

International Nuclear Information System (INIS)

Kamaya, Masayuki; Hojo, Tomohiro; Mochizuki, Masahito

2015-01-01

Load carrying capacity of austenitic stainless steel component is increased due to hardening caused by neutron irradiation if no crack is included in the component. On the other hand, if a crack is initiated in the reactor components, the hardening may decrease the load carrying capacity due to reduction in fracture toughness. In this paper, in order to develop a failure assessment procedure of irradiated cracked components, characteristics of change in failure strength of stainless steels due to cold working were investigated. It was experimentally shown that the proof and tensile strengths were increased by the cold working, whereas the fracture toughness was decreased. The fracture strengths of a cylinder with a circumferential surface crack were analyzed using the obtained material properties. Although the cold working altered the failure mode from plastic collapse to the unsteady ductile crack growth, it did not reduce failure strengths even if 50% cold working was applied. The increase in failure strength was caused not only by increase in flow stress but also by reduction in J-integral value, which was brought by the change in stress-strain curve. It was shown that the failure strength of the hardened stainless steel components could be derived by the two-parameter method, in which the change in material properties could be reasonably considered. (author)

Analysis of failure dependent test, repair and shutdown strategies for redundant trains

International Nuclear Information System (INIS)

Uryasev, S.; Samanta, P.

1994-09-01

Failure-dependent testing implies a test of a redundant components (or trains) when failure of one component has been detected. The purpose of such testing is to detect any common cause failures (CCFs) of multiple components so that a corrective action such as repair or plant shutdown can be taken to reduce the residence time of multiple failures, given a failure has been detected. This type of testing focuses on reducing the conditional risk of CCFs. Formulas for calculating the conditional failure probability of a two train system with different test, repair and shutdown strategies are developed. A methodology is presented with an example calculation showing the risk-effectiveness of failure-dependent strategies for emergency diesel generators (EDGs) in nuclear power plants (NPPs)

Probability of failure of the waste hoist brake system at the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

Greenfield, M.A.; Sargent, T.J.; Stanford Univ., CA

1998-01-01

In its most recent report on the annual probability of failure of the waste hoist brake system at the Waste Isolation Pilot Plant (WIPP), the annual failure rate is calculated to be 1.3E(-7)(1/yr), rounded off from 1.32E(-7). A calculation by the Environmental Evaluation Group (EEG) produces a result that is about 4% higher, namely 1.37E(-7)(1/yr). The difference is due to a minor error in the US Department of Energy (DOE) calculations in the Westinghouse 1996 report. WIPP's hoist safety relies on a braking system consisting of a number of components including two crucial valves. The failure rate of the system needs to be recalculated periodically to accommodate new information on component failure, changes in maintenance and inspection schedules, occasional incidents such as a hoist traveling out-of-control, either up or down, and changes in the design of the brake system. This report examines DOE's last two reports on the redesigned waste hoist system. In its calculations, the DOE has accepted one EEG recommendation and is using more current information about the component failures rates, the Nonelectronic Parts Reliability Data (NPRD). However, the DOE calculations fail to include the data uncertainties which are described in detail in the NPRD reports. The US Nuclear Regulatory Commission recommended that a system evaluation include mean estimates of component failure rates and take into account the potential uncertainties that exist so that an estimate can be made on the confidence level to be ascribed to the quantitative results. EEG has made this suggestion previously and the DOE has indicated why it does not accept the NRC recommendation. Hence, this EEG report illustrates the importance of including data uncertainty using a simple statistical example

Assessment of electronic component failure rates on the basis of experimental data

International Nuclear Information System (INIS)

Nitsch, R.

1991-01-01

Assessment and prediction of failure rates of electronic systems are made using experimental data derived from laboratory-scale tests or from the practice, as for instance from component failure rate statistics or component repair statistics. Some problems and uncertainties encountered in an evaluation of such field data are discussed in the paper. In order to establish a sound basis for comparative assessment of data from various sources, the items of comparison and the procedure in case of doubt have to be defined. The paper explains two standard methods proposed for practical failure rate definition. (orig.) [de

Reliability prediction system based on the failure rate model for electronic components

International Nuclear Information System (INIS)

Lee, Seung Woo; Lee, Hwa Ki

2008-01-01

Although many methodologies for predicting the reliability of electronic components have been developed, their reliability might be subjective according to a particular set of circumstances, and therefore it is not easy to quantify their reliability. Among the reliability prediction methods are the statistical analysis based method, the similarity analysis method based on an external failure rate database, and the method based on the physics-of-failure model. In this study, we developed a system by which the reliability of electronic components can be predicted by creating a system for the statistical analysis method of predicting reliability most easily. The failure rate models that were applied are MILHDBK- 217F N2, PRISM, and Telcordia (Bellcore), and these were compared with the general purpose system in order to validate the effectiveness of the developed system. Being able to predict the reliability of electronic components from the stage of design, the system that we have developed is expected to contribute to enhancing the reliability of electronic components

Integrity evaluation of power plant components by fracture mechanics and related techniques

International Nuclear Information System (INIS)

Mukherjee, B.; Vanderglas, M.L.; Davies, P.H.

1982-01-01

Power plant components can be subject to unexpected failures with serious consequences, unless careful attention is paid to minute crack defects and their possible growth. The Linear Elastic Fracture Mechanics approach to structural integrity evaluation, as it appears in the ASME Code, is discussed. Projects related to material data generation and the development of structural analysis methods to make the above method usable are described. Several integrity-related questions outside the scope of the Code guidelines are documented, concluding with comments on possible future developments

Component Fragility Research Program: Phase 1 component prioritization

International Nuclear Information System (INIS)

Holman, G.S.; Chou, C.K.

1987-06-01

Current probabilistic risk assessment (PRA) methods for nuclear power plants utilize seismic ''fragilities'' - probabilities of failure conditioned on the severity of seismic input motion - that are based largely on limited test data and on engineering judgment. Under the NRC Component Fragility Research Program (CFRP), the Lawrence Livermore National Laboratory (LLNL) has developed and demonstrated procedures for using test data to derive probabilistic fragility descriptions for mechanical and electrical components. As part of its CFRP activities, LLNL systematically identified and categorized components influencing plant safety in order to identify ''candidate'' components for future NRC testing. Plant systems relevant to safety were first identified; within each system components were then ranked according to their importance to overall system function and their anticipated seismic capacity. Highest priority for future testing was assigned to those ''very important'' components having ''low'' seismic capacity. This report describes the LLNL prioritization effort, which also included application of ''high-level'' qualification data as an alternate means of developing probabilistic fragility descriptions for PRA applications

Packaging-induced failure of semiconductor lasers and optical telecommunications components

Energy Technology Data Exchange (ETDEWEB)

Sharps, J.A. [Corning Inc., NY (United States)

1996-12-31

Telecommunications equipment for field deployment generally have specified lifetimes of > 100,000 hr. To achieve this high reliability, it is common practice to package sensitive components in hermetic, inert gas environments. The intent is to protect components from particulate and organic contamination, oxidation, and moisture. However, for high power density 980 nm diode lasers used in optical amplifiers, the authors found that hermetic, inert gas packaging induced a failure mode not observed in similar, unpackaged lasers. They refer to this failure mode as packaging-induced failure, or PIF. PIF is caused by nanomole amounts of organic contamination which interact with high intensity 980 nm light to form solid deposits over the emitting regions of the lasers. These deposits absorb 980 nm light, causing heating of the laser, narrowing of the band gap, and eventual thermal runaway. The authors have found PIF is averted by packaging with free O{sub 2} and/or a getter material that sequesters organics.

Data collection on component malfunctions and failures of JET ICRH system

International Nuclear Information System (INIS)

Pinna, T.; Cambi, G.

2007-01-01

The objective of the activity was to collect and analyse data coming out from operating experiences gained in the Joint European Torus (JET) for the Ion Cyclotron Resonance Heating (ICRH) system in order to enrich the data collection on failures of components used in fusion facilities. Alarms/Failures and malfunctions occurred in the years of operations from March 1996 to November 2005, including information on failure modes and, where possible, causes of the failures, have been identified. Beyond information on failures and alarms events, also data related to crowbar events have been collected. About 3400 events classified as alarms or failures related to specific components or sub-systems were identified by analysing the 25 hand-written logbooks made available by the ICRH operation staff. Information about the JET pulses in which the ICRH system was operated has been extracted from the tick sheets covering the whole considered time interval. 20 hand written tick sheets cover the period from March 1996 to middle May 2003, while tick sheets recorded as excel files cover the period from May 2003 to November 2005. By analysing the tick sheets it results that the ICRH was operated during about 12000 plasma pulses. Main statistical values, such as rates of alarms/failures and corresponding standard errors and confidence intervals, have been estimated. Failure rates of systems and components have been evaluated both with regard to the ICRH operation pulses and operating days (days in which at least one ICRH module was requested to operate). Failure probabilities on demand have been evaluated with regard to number of pulses operated. Some of the results are the following: - The highest number of alarms/failures (1243) appears to be related to Erratic /No-output of the Instrumentation and Control (I and C) apparatus, followed by faults (829) of the Tetrode circuits, by faults (466) of the High Voltage Power Supply system and by faults (428) of the Tuning elements. - The

Design and implementation of component reliability database management system for NPP

International Nuclear Information System (INIS)

Kim, S. H.; Jung, J. K.; Choi, S. Y.; Lee, Y. H.; Han, S. H.

1999-01-01

KAERI is constructing the component reliability database for Korean nuclear power plant. This paper describes the development of data management tool, which runs for component reliability database. This is running under intranet environment and is used to analyze the failure mode and failure severity to compute the component failure rate. Now we are developing the additional modules to manage operation history, test history and algorithms for calculation of component failure history and reliability

A real-time expert system for nuclear power plant failure diagnosis and operational guide

International Nuclear Information System (INIS)

Naito, N.; Sakuma, A.; Shigeno, K.; Mori, N.

1987-01-01

A real-time expert system (DIAREX) has been developed to diagnose plant failure and to offer a corrective operational guide for boiling water reactor (BWR) power plants. The failure diagnosis model used in DIAREX was systematically developed, based mainly on deep knowledge, to cover heuristics. Complex paradigms for knowledge representation were adopted, i.e., the process representation language and the failure propagation tree. The system is composed of a knowledge base, knowledge base editor, preprocessor, diagnosis processor, and display processor. The DIAREX simulation test has been carried out for many transient scenarios, including multiple failures, using a real-time full-scope simulator modeled after the 1100-MW(electric) BWR power plant. Test results showed that DIAREX was capable of diagnosing a plant failure quickly and of providing a corrective operational guide with a response time fast enough to offer valuable information to plant operators

Long-term effects as the cause of failure in electronic components

International Nuclear Information System (INIS)

Renz, H.; Kreichgauer, H.

1989-01-01

After a brief presentation of the utilisation properties of electronic components, their failure rates are discussed with particular reference to the socalled bath-tub curve. The main emphasis is on the construction and manufacture of integrated circuits and the possible types and causes of failure arising from the individual manufacturing stages (layout faults, internal corrosion, masking and etching errors, leakage currents, inadequate heat removal, etc.). A technical insurance assessment is then provided of the long-term failures associated with technological matters. (orig.) [de

Sizes of secondary plant components for modularized IRIS balance of plant design

International Nuclear Information System (INIS)

Williamson, Martin; Townsend, Lawrence

2003-01-01

Herein we report on a conceptual design for a balance of plant (BOP) layout to coordinate with IRIS-like plants. The report consists of results of calculations that sizes of various BOP components. These calculations include the thermodynamic analyses and general sizing of the components in order to determine plant capability and plant layout for studies on modularity and transportability. Mathematical modeling of the BOP system involves a modified ORCENT2 code as well as standard heat transfer methods. Using typical values for PWR type plants, a general BOP design, and IRIS steam generator values, an ORCENT2 heat balance is carried out for the secondary side of the plant. Using the ORCENT2 output, standard heat transfer methods are then used to calculate system performance and component sizes. (author)

Failure analysis of storage tank component in LNG regasification unit using fault tree analysis method (FTA)

Science.gov (United States)

Mulyana, Cukup; Muhammad, Fajar; Saad, Aswad H.; Mariah, Riveli, Nowo

2017-03-01

Storage tank component is the most critical component in LNG regasification terminal. It has the risk of failure and accident which impacts to human health and environment. Risk assessment is conducted to detect and reduce the risk of failure in storage tank. The aim of this research is determining and calculating the probability of failure in regasification unit of LNG. In this case, the failure is caused by Boiling Liquid Expanding Vapor Explosion (BLEVE) and jet fire in LNG storage tank component. The failure probability can be determined by using Fault Tree Analysis (FTA). Besides that, the impact of heat radiation which is generated is calculated. Fault tree for BLEVE and jet fire on storage tank component has been determined and obtained with the value of failure probability for BLEVE of 5.63 × 10-19 and for jet fire of 9.57 × 10-3. The value of failure probability for jet fire is high enough and need to be reduced by customizing PID scheme of regasification LNG unit in pipeline number 1312 and unit 1. The value of failure probability after customization has been obtained of 4.22 × 10-6.

Design issues and implications for the structural integrity and lifetime of fusion power plant components

International Nuclear Information System (INIS)

Karditas, P.J.

1996-05-01

This review discusses, with example calculations, the criteria, and imposed constraints and limitations, for the design of fusion components and assesses the implications for successful design and power plant operation. The various loading conditions encountered during the operation of a tokamak lead to structural damage and possible failure by such mechanisms as yielding, thermal creep rupture and fatigue due to thermal cycling, plastic strain cycling (ratcheting), crack growth-propagation and radiation induced swelling and creep. Of all the possible damage mechanisms, fatigue, creep and their combination are the most important in the structural design and lifetime of fusion power plant components operating under steady or load varying conditions. Also, the effect of neutron damage inflicted onto the structural materials and the degradation of key properties is of major concern in the design and lifetime prediction of components. Structures are classified by, and will be restricted by existing or future design codes relevant to medium and high temperature power plant environments. The ways in which existing design codes might be used in present and near future design activities, and the implications, are discussed; the desirability of an early start towards the development of fusion-specific design codes is emphasised. (UK)

Development of web-based integrity evaluation system for primary components in a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J. [SAFE Research Center, Sungkyunkwan Univ., Suwon (Korea); Choi, S.N.; Jang, K.S.; Hong, S.Y. [Korea Electronic Power Research Inst., Daejeon (Korea)

2004-07-01

A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

Development of web-based integrity evaluation system for primary components in a nuclear power plant

International Nuclear Information System (INIS)

Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J.; Choi, S.N.; Jang, K.S.; Hong, S.Y.

2004-01-01

A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

Feasibility study of component risk ranking for plant maintenance

International Nuclear Information System (INIS)

Ushijima, Koji; Yonebayashi, Kenji; Narumiya, Yoshiyuki; Sakata, Kaoru; Kumano, Tetsuji

1999-01-01

Nuclear power is the base load electricity source in Japan, and reduction of operation and maintenance cost maintaining or improving plant safety is one of the major issues. Recently, Risk Informed Management (RIM) is focused as a solution. In this paper, the outline regarding feasibility study of component risk ranking for plant maintenance for a typical Japanese PWR plant is described. A feasibility study of component risk raking for plant maintenance optimization is performed on check valves and motor-operated valves. Risk ranking is performed in two steps using probabilistic analysis (quantitative method) for risk ranking of components, and deterministic examination (qualitative method) for component review. In this study, plant components are ranked from the viewpoint of plant safety / reliability, and the applicability for maintenance is assessed. As a result, distribution of maintenance resources using risk ranking is considered effective. (author)

A quasi-independence model to estimate failure rates

International Nuclear Information System (INIS)

Colombo, A.G.

1988-01-01

The use of a quasi-independence model to estimate failure rates is investigated. Gate valves of nuclear plants are considered, and two qualitative covariates are taken into account: plant location and reactor system. Independence between the two covariates and an exponential failure model are assumed. The failure rate of the components of a given system and plant is assumed to be a constant, but it may vary from one system to another and from one plant to another. This leads to the analysis of a contingency table. A particular feature of the model is the different operating time of the components in the various cells which can also be equal to zero. The concept of independence of the covariates is then replaced by that of quasi-independence. The latter definition, however, is used in a broader sense than usual. Suitable statistical tests are discussed and a numerical example illustrates the use of the method. (author)

Reactor materials program process water component failure probability

International Nuclear Information System (INIS)

Daugherty, W. L.

1988-01-01

The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

A Methodology for Modeling Nuclear Power Plant Passive Component Aging in Probabilistic Risk Assessment under the Impact of Operating Conditions, Surveillance and Maintenance Activities

Science.gov (United States)

Guler Yigitoglu, Askin

In the context of long operation of nuclear power plants (NPPs) (i.e., 60-80 years, and beyond), investigation of the aging of passive systems, structures and components (SSCs) is important to assess safety margins and to decide on reactor life extension as indicated within the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. In the traditional probabilistic risk assessment (PRA) methodology, evaluating the potential significance of aging of passive SSCs on plant risk is challenging. Although passive SSC failure rates can be added as initiating event frequencies or basic event failure rates in the traditional event-tree/fault-tree methodology, these failure rates are generally based on generic plant failure data which means that the true state of a specific plant is not reflected in a realistic manner on aging effects. Dynamic PRA methodologies have gained attention recently due to their capability to account for the plant state and thus address the difficulties in the traditional PRA modeling of aging effects of passive components using physics-based models (and also in the modeling of digital instrumentation and control systems). Physics-based models can capture the impact of complex aging processes (e.g., fatigue, stress corrosion cracking, flow-accelerated corrosion, etc.) on SSCs and can be utilized to estimate passive SSC failure rates using realistic NPP data from reactor simulation, as well as considering effects of surveillance and maintenance activities. The objectives of this dissertation are twofold: The development of a methodology for the incorporation of aging modeling of passive SSC into a reactor simulation environment to provide a framework for evaluation of their risk contribution in both the dynamic and traditional PRA; and the demonstration of the methodology through its application to pressurizer surge line pipe weld and steam generator tubes in commercial nuclear power plants. In the proposed methodology, a

The qualification of electrical components and instrumentations relevant to safety

CERN Document Server

Zambardi, F

1989-01-01

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and ...

Reliability prediction of engineering systems with competing failure modes due to component degradation

International Nuclear Information System (INIS)

Son, Young Kap

2011-01-01

Reliability of an engineering system depends on two reliability metrics: the mechanical reliability, considering component failures, that a functional system topology is maintained and the performance reliability of adequate system performance in each functional configuration. Component degradation explains not only the component aging processes leading to failure in function, but also system performance change over time. Multiple competing failure modes for systems with degrading components in terms of system functionality and system performance are considered in this paper with the assumption that system functionality is not independent of system performance. To reduce errors in system reliability prediction, this paper tries to extend system performance reliability prediction methods in open literature through combining system mechanical reliability from component reliabilities and system performance reliability. The extended reliability prediction method provides a useful way to compare designs as well as to determine effective maintenance policy for efficient reliability growth. Application of the method to an electro-mechanical system, as an illustrative example, is explained in detail, and the prediction results are discussed. Both mechanical reliability and performance reliability are compared to total system reliability in terms of reliability prediction errors

A technique of including the effect of aging of passive components in probabilistic risk assessments

International Nuclear Information System (INIS)

Phillips, J.H.; Weidenhamer, G.H.

1992-01-01

The probabilistic risk assessments (PRAS) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing methods for selecting risk-significant passive components and including them in PRAS. These methods provide effective ways to prioritize passive components for inspection, and where inspection reveals aging damage, mitigation or repair can be employed to reduce the likelihood of component failure. We demonstrated a method by selecting a weld in the auxiliary feedwater (AFW) system, basing our selection on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then modified and used the Piping Reliability Analysis Including Seismic Events (PRAISE) computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The PRAISE code was modified to include the effects of changing design material properties with age and changing stress cycles. The calculation included the effects of mechanical loads and thermal transients typical of the service loads for this piping design and the effects of thermal cycling caused by a leaking check valve. However, this particular calculation showed little change in low component failure probability and plant risk for 48 years of service. However, sensitivity studies showed that if the probability of component failure is high, the effect on plant risk is significant. The success of this demonstration shows that this method could be applied to nuclear power plants. The demonstration showed the method is too involved (PRAISE takes a long time to perform the calculation and the input information is extensive) for handling a large number of passive components and therefore simpler methods are needed

Study of aging effects in PWR power plants components - 15043

International Nuclear Information System (INIS)

Silva Borges, D. da; Lava, D.D.; Guimaraes, A.C.F.; Moreira, M. de L.

2015-01-01

In this paper we present a simulation about the aging process of the containment spray injection system (CSIS) of a pressurized water reactor (PWR) using the fault tree method (FT). The FT has the capacity to present the logic of events that leads to system unavailability, to capture frequency estimation of events, to model and calculate hazardous events frequency (before they happen) and help developing protective layers. The Monte Carlo method and Fussell-Vesely importance measure are used in this paper to determine the system unavailability probability and the most sensitive events to the aging process. The injection system fault tree consists of a main tree and 10 sub-trees. The main tree is composed of 35 basic events, 5 gates and 1 top event. The paper details the methodology. It can be seen that the increase of the failure rate of components due to the aging process, generates the increase in the general unavailability of the system that contains these components. The extension of the operating life of nuclear power plant must be accompanied by a special attention to the aging process of its components

Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

International Nuclear Information System (INIS)

Azarm, M.A.; Boccio, J.L.; Mitra, S.

1985-12-01

This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs

Analysis of failure events for expansion joints in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Sato, Masahiro [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Although a large number of expansion joints are used in nuclear power plants with light water reactors, their failure events have not been paid as much attention as those of vessels and pipes. However, as the operation period of nuclear power plants becomes longer, it is necessary to pay attention to their failure events as well as those of vessels and pipes, because aging problems and latent troubles originated in design or fabrication of expansion joints may appear during their long period operation. In this work, we investigated failure event reports of expansion joints in nuclear power plants both in Japan and in U.S.A. and analyzed (1) the influence to output power level, (2) the position and (3) the cause of each failure. It is revealed that the failure events of expansion joints have continuously occurred, some of which have exerted influence upon power level and have caused fatal or injury accidents of personnel, and hence the importance of corrective actions to prevent the recurrence of such events is pointed out. The importance of countermeasures to the following individual events is also pointed out: (1) corrosion of expansion joints in service water systems, (2) degradation of rubber expansion joints in main condensers, (3) vibration and fatigue of expansion joints in extraction steam lines and (4) transgranular stress corrosion cracking of penetration bellows of containments. (author)

Exploitation of a component event data bank for common cause failure analysis

International Nuclear Information System (INIS)

Games, A.M.; Amendola, A.; Martin, P.

1985-01-01

Investigations into using the European Reliability Data System Component Event Data Bank for common cause failure analysis have been carried out. Starting from early exercises where data were analyzed without computer aid, different types of linked multiple failures have been identified. A classification system is proposed based on this experience. It defines a multiple failure event space wherein each category defines causal, modal, temporal and structural links between failures. It is shown that a search algorithm which incorporates the specific interrogative procedures of the data bank can be developed in conjunction with this classification system. It is concluded that the classification scheme and the search algorithm are useful organizational tools in the field of common cause failures studies. However, it is also suggested that the use of the term common cause failure should be avoided since it embodies to many different types of linked multiple failures

Cases of corrosion in power plant components at NTPC

International Nuclear Information System (INIS)

Sanyal, S.K.; Bhakta, U.C.; Sinha, Ashwini

2000-01-01

Power plants are one of the major industries suffering from severe corrosion problems resulting in substantial losses. The problem is becoming more prominent as the plants are getting older. NTPC as the leading power utility with very good performance track record, had been conscious of the menace of corrosion prevailing in the industry and had established a Research and Development Centre to cater to applied O and M needs of the plants. A specialized group has been involved in studying the corrosion related problems and recommending suitable cost effective solutions to such problems. The present paper aims at discussing various corrosion related analysis carried out at the Research and Development Centre of NTPC and the remedial measures suggested. The paper also describes some of the case studies of corrosion related failures with recommendations given for preventing such failures in future. (author)

Failure mode, effect and criticality analysis (FMECA) on mechanical subsystems of diesel generator at NPP

International Nuclear Information System (INIS)

Kim, Tae Woon; Singh, Brijendra; Sung, Tae Yong; Park, Jin Hee; Lee, Yoon Hwan

1996-06-01

Largely, the RCM approach can be divided in three phases; (1) Functional failure analysis (FFA) on the selected system or subsystem, (2) Failure mode, effect and criticality analysis (FMECA) to identify the impact of failure to plant safety or economics, (3) Logical tree analysis (LTA) to select appropriate preventive maintenance and surveillance tasks. This report presents FMECA results for six mechanical subsystems of the diesel generators of nuclear power plants. The six mechanical subsystems are Starting air, Lub oil, Governor, Jacket water cooling, Fuel, and Engine subsystems. Generic and plant-specific failure and maintenance records are reviewed to identify critical components/failure modes. FMECA was performed for these critical component/failure modes. After reviewing current preventive maintenance activities of Wolsung unit 1, draft RCM recommendations are developed. 6 tabs., 16 refs. (Author)

Failure mode, effect and criticality analysis (FMECA) on mechanical subsystems of diesel generator at NPP

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Woon; Singh, Brijendra; Sung, Tae Yong; Park, Jin Hee; Lee, Yoon Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

Largely, the RCM approach can be divided in three phases; (1) Functional failure analysis (FFA) on the selected system or subsystem, (2) Failure mode, effect and criticality analysis (FMECA) to identify the impact of failure to plant safety or economics, (3) Logical tree analysis (LTA) to select appropriate preventive maintenance and surveillance tasks. This report presents FMECA results for six mechanical subsystems of the diesel generators of nuclear power plants. The six mechanical subsystems are Starting air, Lub oil, Governor, Jacket water cooling, Fuel, and Engine subsystems. Generic and plant-specific failure and maintenance records are reviewed to identify critical components/failure modes. FMECA was performed for these critical component/failure modes. After reviewing current preventive maintenance activities of Wolsung unit 1, draft RCM recommendations are developed. 6 tabs., 16 refs. (Author).

Risk-based assessment of the allowable outage times for the unit 1 leningrad nuclear power plant ECCS components

International Nuclear Information System (INIS)

Koukhar, Sergey; Vinnikov, Bronislav

2009-01-01

Present paper describes a method for risk - informed assessment of the Allowable Outage Times (AOTs). The AOT is the time, when components of a safety system allowed to be out of service during power operation or during shutdown operation off a plant. If the components are not restored during the time, the plant in operation must be shut down or the plant in a given shutdown mode has to go to safer shutdown mode. Application of the method is also provided for the equipment of the Unit 1 Leningrad NPP ECCS components. For solution of the problem it is necessary to carry out two series of computations using a Living PSA model, level 1. In the first series of the computations the core damage frequency (CDFb) for the base configuration of the plant is determined (there is no equipment out of service). Here the symbol 'b' means the base configuration of a plant. In the second series of the computations the core damage frequency (CDFi) for the configuration of the plant with the component (which is out of service) is calculated. That is here CDFi is determined for the failure probability of the component equal to 1.0 (component 'i' is unavailable). Then it is necessary to determine so called Risk Increase Factor (RIF) using the following ratio: RIFi = CDFi / CDFb. At last the AOT is calculated with the help of the ratio: AOTi = Tppr / RIFi, where Tppr is a period of time between two Planned Preventive Repairs (PPRs). 1. Using the risk based approach the AOTs were calculated for a set of the components of the Unit 1 Leningrad NPP ECCS components. 2. The main conclusion from the analysis is that the current deterministic AOTs for the ECCS components are conservative and should be extended. 3. The risk based extension of the AOTs for the ECCS components can prevent the Unit 1 Leningrad NPP to enter into the operating modes with increased risk. (author)

The qualification of electrical components and instrumentations relevant to safety

International Nuclear Information System (INIS)

Zambardi, F.

1989-03-01

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and peculiar aspects are discussed and the qualification by testing is taken into special account. Technical and organizational aspects related to a plant qualification program are also focused. The report ends with a look to the most significant research and development activities. (author)

Risk analysis of geothermal power plants using Failure Modes and Effects Analysis (FMEA) technique

International Nuclear Information System (INIS)

Feili, Hamid Reza; Akar, Navid; Lotfizadeh, Hossein; Bairampour, Mohammad; Nasiri, Sina

2013-01-01

Highlights: • Using Failure Modes and Effects Analysis (FMEA) to find potential failures in geothermal power plants. • We considered 5 major parts of geothermal power plants for risk analysis. • Risk Priority Number (RPN) is calculated for all failure modes. • Corrective actions are recommended to eliminate or decrease the risk of failure modes. - Abstract: Renewable energy plays a key role in the transition toward a low carbon economy and the provision of a secure supply of energy. Geothermal energy is a versatile source as a form of renewable energy that meets popular demand. Since some Geothermal Power Plants (GPPs) face various failures, the requirement of a technique for team engineering to eliminate or decrease potential failures is considerable. Because no specific published record of considering an FMEA applied to GPPs with common failure modes have been found already, in this paper, the utilization of Failure Modes and Effects Analysis (FMEA) as a convenient technique for determining, classifying and analyzing common failures in typical GPPs is considered. As a result, an appropriate risk scoring of occurrence, detection and severity of failure modes and computing the Risk Priority Number (RPN) for detecting high potential failures is achieved. In order to expedite accuracy and ability to analyze the process, XFMEA software is utilized. Moreover, 5 major parts of a GPP is studied to propose a suitable approach for developing GPPs and increasing reliability by recommending corrective actions for each failure mode

Holistic and component plant phenotyping using temporal image sequence.

Science.gov (United States)

Das Choudhury, Sruti; Bashyam, Srinidhi; Qiu, Yumou; Samal, Ashok; Awada, Tala

2018-01-01

Image-based plant phenotyping facilitates the extraction of traits noninvasively by analyzing large number of plants in a relatively short period of time. It has the potential to compute advanced phenotypes by considering the whole plant as a single object (holistic phenotypes) or as individual components, i.e., leaves and the stem (component phenotypes), to investigate the biophysical characteristics of the plants. The emergence timing, total number of leaves present at any point of time and the growth of individual leaves during vegetative stage life cycle of the maize plants are significant phenotypic expressions that best contribute to assess the plant vigor. However, image-based automated solution to this novel problem is yet to be explored. A set of new holistic and component phenotypes are introduced in this paper. To compute the component phenotypes, it is essential to detect the individual leaves and the stem. Thus, the paper introduces a novel method to reliably detect the leaves and the stem of the maize plants by analyzing 2-dimensional visible light image sequences captured from the side using a graph based approach. The total number of leaves are counted and the length of each leaf is measured for all images in the sequence to monitor leaf growth. To evaluate the performance of the proposed algorithm, we introduce University of Nebraska-Lincoln Component Plant Phenotyping Dataset (UNL-CPPD) and provide ground truth to facilitate new algorithm development and uniform comparison. The temporal variation of the component phenotypes regulated by genotypes and environment (i.e., greenhouse) are experimentally demonstrated for the maize plants on UNL-CPPD. Statistical models are applied to analyze the greenhouse environment impact and demonstrate the genetic regulation of the temporal variation of the holistic phenotypes on the public dataset called Panicoid Phenomap-1. The central contribution of the paper is a novel computer vision based algorithm for

Prevention of bolting degradation or failure in pressure boundary and support applications

International Nuclear Information System (INIS)

Merrick, E.A.; Rivers, A.; Bickford, J.; Marston, T.U.

1986-01-01

A discussion is presented of bolting degradation or failure experience in pressure boundary and component support applications in US commercial nuclear plants and the industry program to prevent failures in the future. The focus turns to steps which plant owners can take today to guard against pressure boundary bolt failure or degradation for existing plants or units being constructed. 'Tools' or products which the plant owner can expect from current industry programs which will be available in the near future to aid in understanding and improving bolting practices are described. (author)

Validation gets underway on Sizewell ''Incredibility of Failure'' components

International Nuclear Information System (INIS)

Anon.

1992-01-01

The Inspection Validation Centre (IVC) of AEA Reactor Services in the UK has begun an eighteen month programme to validate the procedures and personnel of OIS plc, the inspection agents chosen by Nuclear Electric to carry out the pre-service ultrasonic inspection of the Sizewell B Pressurized Water Reactor components assigned to the ''Incredibility of Failure'' (IoF) category. The work involves several Sizewell B primary circuit components - the steam generators, pressurizer, and primary pumps - and will consider the inspections to be applied to the circumferential and nozzle-to-shell welds, nozzle inner radii and the pump fly-wheel forging. The validation will provide independent confirmation that OIS personnel are capable of using manual and automated methods to find and size any flaws of structural concern in these components. (author)

Ground failure in direct current systems of the Itaipu Hydroelectric Power Plant, Parana, Brazil. Impact in the operation; Falla a tierra en sistemas de corriente continua en la Central Hidroelectrica Itaipu, PR, Brasil. Impacto en la operacion

Energy Technology Data Exchange (ETDEWEB)

Soto Santacruz, Heriberto [Itaipu Binacional, Foz do Iguacu, PR (Brazil)]. E-mail: soto@itaipu.gov.py

1998-07-01

The objective of this work is to share with other companies the operation experience obtained by researching the direct current systems ground failure, in the Itaipu Hydroelectric Power Plant. During the research process electrical and/or electronic components can be damaged, and also human failures can occurred during the circuit connection and disconnection manoeuvres, necessary for the identification of the components causing the failures.

A study on the optimal replacement periods of digital control computer's components of Wolsung nuclear power plant unit 1

International Nuclear Information System (INIS)

Mok, Jin Il; Seong, Poong Hyun

1993-01-01

Due to the failure of the instrument and control devices of nuclear power plants caused by aging, nuclear power plants occasionally trip. Even a trip of a single nuclear power plant (NPP) causes an extravagant economical loss and deteriorates public acceptance of nuclear power plants. Therefore, the replacement of the instrument and control devices with proper consideration of the aging effect is necessary in order to prevent the inadvertent trip. In this paper we investigated the optimal replacement periods of the control computer's components of Wolsung nuclear power plant Unit 1. We first derived mathematical models of optimal replacement periods to the digital control computer's components of Wolsung NPP Unit 1 and calculated the optimal replacement periods analytically. We compared the periods with the replacement periods currently used at Wolsung NPP Unit 1. The periods used at Wolsung is not based on mathematical analysis, but on empirical knowledge. As a consequence, the optimal replacement periods analytically obtained and those used in the field show a little difference. (Author)

Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

International Nuclear Information System (INIS)

Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1992-10-01

In a study sponsored by the US Nuclear regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. North Anna was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the North Anna plant

Component failures that lead to reactor scrams

International Nuclear Information System (INIS)

Burns, E.T.; Wilson, R.J.; Lim, E.Y.

1980-04-01

This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation

Stochastic models and reliability parameter estimation applicable to nuclear power plant safety

International Nuclear Information System (INIS)

Mitra, S.P.

1979-01-01

A set of stochastic models and related estimation schemes for reliability parameters are developed. The models are applicable for evaluating reliability of nuclear power plant systems. Reliability information is extracted from model parameters which are estimated from the type and nature of failure data that is generally available or could be compiled in nuclear power plants. Principally, two aspects of nuclear power plant reliability have been investigated: (1) The statistical treatment of inplant component and system failure data; (2) The analysis and evaluation of common mode failures. The model inputs are failure data which have been classified as either the time type of failure data or the demand type of failure data. Failures of components and systems in nuclear power plant are, in general, rare events.This gives rise to sparse failure data. Estimation schemes for treating sparse data, whenever necessary, have been considered. The following five problems have been studied: 1) Distribution of sparse failure rate component data. 2) Failure rate inference and reliability prediction from time type of failure data. 3) Analyses of demand type of failure data. 4) Common mode failure model applicable to time type of failure data. 5) Estimation of common mode failures from 'near-miss' demand type of failure data

Machinery failure analysis and troubleshooting practical machinery management for process plants

CERN Document Server

Bloch, Heinz P

2012-01-01

Solve the machinery failure problems costing you time and money with this classic, comprehensive guide to analysis and troubleshooting  Provides detailed, complete and accurate information on anticipating risk of component failure and avoiding equipment downtime Includes numerous photographs of failed parts to ensure you are familiar with the visual evidence you need to recognize Covers proven approaches to failure definition and offers failure identification and analysis methods that can be applied to virtually all problem situations Demonstr

The Hanford Site generic component failure-rate database compared with other generic failure-rate databases

International Nuclear Information System (INIS)

Reardon, M.F.; Zentner, M.D.

1992-11-01

The Risk Assessment Technology Group, Westinghouse Hanford Company (WHC), has compiled a component failure rate database to be used during risk and reliability analysis of nonreactor facilities. Because site-specific data for the Hanford Site are generally not kept or not compiled in a usable form, the database was assembled using information from a variety of other established sources. Generally, the most conservative failure rates were chosen from the databases reviewed. The Hanford Site database has since been used extensively in fault tree modeling of many Hanford Site facilities and systems. The purpose of this study was to evaluate the reasonableness of the data chosen for the Hanford Site database by comparing the values chosen with the values from the other databases

Investigation of component failure rates for pulsed versus steady state tokamak operation

International Nuclear Information System (INIS)

Cadwallader, L.C.

1992-07-01

This report presents component failure rate data sources applicable to magnetic fusion systems, and defines multiplicative factors to adjust these data for specific use on magnetic fusion experiment designs. The multipliers address both long pulse and steady state tokamak operation. Thermal fatigue and radiation damage are among the leading reasons for large multiplier values in pulsed operation applications. Field failure rate values for graphite protective tiles are presented, and beryllium tile failure rates in laboratory testing are also given. All of these data can be used for reliability studies, safety analyses, design tradeoff studies, and risk assessments

PV System Component Fault and Failure Compilation and Analysis.

Energy Technology Data Exchange (ETDEWEB)

Klise, Geoffrey Taylor; Lavrova, Olga; Gooding, Renee Lynne

2018-02-01

This report describes data collection and analysis of solar photovoltaic (PV) equipment events, which consist of faults and fa ilures that occur during the normal operation of a distributed PV system or PV power plant. We present summary statistics from locations w here maintenance data is being collected at various intervals, as well as reliability statistics gathered from that da ta, consisting of fault/failure distributions and repair distributions for a wide range of PV equipment types.

Steady-State Plant Model to Predict Hydroden Levels in Power Plant Components

Energy Technology Data Exchange (ETDEWEB)

Glatzmaier, Greg C.; Cable, Robert; Newmarker, Marc

2017-06-27

The National Renewable Energy Laboratory (NREL) and Acciona Energy North America developed a full-plant steady-state computational model that estimates levels of hydrogen in parabolic trough power plant components. The model estimated dissolved hydrogen concentrations in the circulating heat transfer fluid (HTF), and corresponding partial pressures within each component. Additionally for collector field receivers, the model estimated hydrogen pressure in the receiver annuli. The model was developed to estimate long-term equilibrium hydrogen levels in power plant components, and to predict the benefit of hydrogen mitigation strategies for commercial power plants. Specifically, the model predicted reductions in hydrogen levels within the circulating HTF that result from purging hydrogen from the power plant expansion tanks at a specified target rate. Our model predicted hydrogen partial pressures from 8.3 mbar to 9.6 mbar in the power plant components when no mitigation treatment was employed at the expansion tanks. Hydrogen pressures in the receiver annuli were 8.3 to 8.4 mbar. When hydrogen partial pressure was reduced to 0.001 mbar in the expansion tanks, hydrogen pressures in the receiver annuli fell to a range of 0.001 mbar to 0.02 mbar. When hydrogen partial pressure was reduced to 0.3 mbar in the expansion tanks, hydrogen pressures in the receiver annuli fell to a range of 0.25 mbar to 0.28 mbar. Our results show that controlling hydrogen partial pressure in the expansion tanks allows us to reduce and maintain hydrogen pressures in the receiver annuli to any practical level.

Component failures at pressurized water reactors. Final report

International Nuclear Information System (INIS)

Reisinger, M.F.

1980-10-01

Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis

Drosophila melanogaster "a potential model organism" for identification of pharmacological properties of plants/plant-derived components.

Science.gov (United States)

Panchal, Komal; Tiwari, Anand K

2017-05-01

Plants/plant-derived components have been used from ancient times to treat/cure several human diseases. Plants and their parts possess several chemical components that play the vital role in the improvement of human health and their life expectancy. Allopathic medicines have been playing a key role in the treatment of several diseases. Though allopathic medicines provide fast relief, long time consumption cause serious health concerns such as hyperallergic reactions, liver damage, etc. So, the study of medicinal plants which rarely cause any side effect is very important to mankind. Plants contain many health benefit properties like antioxidant, anti-aging, neuroprotective, anti-genotoxic, anti-mutagenic and bioinsecticidal activity. Thus, identification of pharmacological properties of plants/plant-derived components are of utmost importance to be explored. Several model organisms have been used to identify the pharmacological properties of the different plants or active components therein and Drosophila is one of them. Drosophila melanogaster "fruit fly" is a well understood, high-throughput model organism being used more than 110 years to study the different biological aspects related to the development and diseases. Most of the developmental and cell signaling pathways and ∼75% human disease-related genes are conserved between human and Drosophila. Using Drosophila, one can easily analyze the pharmacological properties of plants/plant-derived components by performing several assays available with flies such as survivorship, locomotor, antioxidant, cell death, etc. The current review focuses on the potential of Drosophila melanogaster for the identification of medicinal/pharmacological properties associated with plants/plant-derived components. Copyright © 2017 Elsevier Masson SAS. All rights reserved.

Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

International Nuclear Information System (INIS)

Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E.

1991-09-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab

T-book. Reliability data of components in Nordic nuclear power plants. 6. ed

International Nuclear Information System (INIS)

2005-01-01

The main objective of the T-Book is to provide reliability data for the unavailability computations that are made for each component that is considered in the compulsory, probabilistic safety assessments (PSA) of nuclear power plants. As the use of PSA is large in the normal safety work at the NPPs, there is a need for easily accessible and reliable failure data. The failure characteristics presented in the T-Book are primarily based on the failure reports stored in the central database TUD and the Licensee Event Reports delivered to the Swedish Nuclear Power Inspectorate (SKI). Fortunately, the TUD database was started already in the middle of the seventies by the Swedish power companies. In 1981, the Finnish power company TVO, operating two reactor units of Swedish design, joined the data collection system. Before the TUD data are statistically treated they are carefully examined with respect to the consistency and correctness. This T-Book comprises only critical failures, i.e. failures that stops the function of components or leads to repair. The first edition of the T-Book was issued in 1982 encompassing operational statistics from 21 reactor years. The second edition was published 1985, based on operating data covering about 40 reactor years. The T-Book 3 was published in 1992 and included data up to the operating year 1987 (108 reactor years). Edition 4 was published 1994 containing information up to and including 1992 (178 reactor years). Edition 5 was published year 2000 containing information up to and including 1996 (234 reactor years). This edition 6 contains information including year 2002 (315 reactor years). At the same time as the amount of data has increased with the successive editions of the T-Book there has been a continuous work to improve the methods for the statistical inference and related program tools, required to derive the reliability parameters from the operational data in the database. Already in the initial edition there was a Bayesian

Performance of materials in the component cooling water systems of pressurized water reactors

International Nuclear Information System (INIS)

Lee, B.S.

1993-01-01

The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

Analysis of soft rock mineral components and roadway failure mechanism

Institute of Scientific and Technical Information of China (English)

陈杰

2001-01-01

The mineral components and microstructure of soft rock sampled from roadway floor inXiagou pit are determined by X-ray diffraction and scanning electron microscope. Ccmbined withthe test of expansion and water softening property of the soft rock, the roadway failure mechanism is analyzed, and the reasonable repair supporting principle of roadway is put forward.

Failure investigation of super heater tubes of coal fired power plant

Directory of Open Access Journals (Sweden)

A.K. Pramanick

2017-10-01

Full Text Available Cause of failure of two adjacent super heater tubes made of Cr-Mo steel of a coal based 60 MW thermal power plant has been portrayed in present investigation. Oxide deposits were found on internal surface of tubes. Deposits created significant resistance to heat transfer and resulted in undesirable rise in component temperature. This situation, in turn, aggravated the condition of gas side that was exposed to high temperature. Localized heating coarsened carbides as well as propelled precipitation of new brittle phases along grain boundary resulting in embrittlement of tube material. Continuous exposure to high temperature softened the tube material and tube wall was thinned down with bulging toward outside. Creep void formation along grain boundary was observed and steered intergranular cracking. All these effects contributed synergistically and tubes were failed ultimately due to overload under high Hoop stress.

Feasibility of risk-informed technology for japanese nuclear power plants

International Nuclear Information System (INIS)

Yoshida, Tomoo; Fujioka, Terutaka; Kirimoto, Yorihiro; Ueda, Nobuyuki; Kinoshita, Izumi; Kashima, Koichi

2000-01-01

Risk-informed technology utilizes Probabilistic Safety Assessment for streamlining the maintenance of nuclear power plants. With this technology, plant components are categorized as either high or low-safety-significant components. The Maintenance requirements focuses on high safety-significant components and are relieved for low safety significant ones. This is expected to reduce plant cost while maintaining safety. We investigated especially risk-informed inservice inspection of piping in U.S. nuclear power plants in the interest of determining its feasibility for Japanese plants. Quantitative and qualitative RI-ISI methods were developed by the ASME/Westinghouse Owners Group and EPRI, respectively. These methods have been incorporated in the ASME Section 11 Code Cases and endorsed by the U.S. Nuclear Regulatory Commission. The quantitative method evaluates component segment risks in terms of pipe failure probability calculated with a probabilistic fracture mechanics(PFM) model and pipe failure impact categorization on core damage frequency(CDF) calculated with PSA. The qualitative method uses pipe failure potential categorization derived from the plant service experiences and pipe failure impact on CDF derived from the PSA insight. The PFM model is applicable only to failures from initial welding defects and stress corrosion cracking, therefore it does not cover such significant failure mechanisms found in nuclear power plants as corrosion or high-cycle fatigue, etc.. Thus, a qualitative failure potential categorization method was developed on the basis of the service experiences of the U.S. nuclear power plants, so that appropriate categorization rules must be developed on the service experiences in Japanese plants. Accordingly, we have devised a software framework with a computer-aided system for the selection of risk significant elements. This system consists of a piping failure database module, a piping failure analysis module, and a piping failure potential

Component Repair Times Obtained from MSPI Data

International Nuclear Information System (INIS)

Eide, Steven A.; Cadwallader, Lee

2015-01-01

Information concerning times to repair or restore equipment to service given a failure is valuable to probabilistic risk assessments (PRAs). Examples of such uses in modern PRAs include estimation of the probability of failing to restore a failed component within a specified time period (typically tied to recovering a mitigating system before core damage occurs at nuclear power plants) and the determination of mission times for support system initiating event (SSIE) fault tree models. Information on equipment repair or restoration times applicable to PRA modeling is limited and dated for U.S. commercial nuclear power plants. However, the Mitigating Systems Performance Index (MSPI) program covering all U.S. commercial nuclear power plants provides up-to-date information on restoration times for a limited set of component types. This paper describes the MSPI program data available and analyzes the data to obtain median and mean component restoration times as well as non-restoration cumulative probability curves. The MSPI program provides guidance for monitoring both planned and unplanned outages of trains of selected mitigating systems deemed important to safety. For systems included within the MSPI program, plants monitor both train UA and component unreliability (UR) against baseline values. If the combined system UA and UR increases sufficiently above established baseline results (converted to an estimated change in core damage frequency or CDF), a ''white'' (or worse) indicator is generated for that system. That in turn results in increased oversight by the US Nuclear Regulatory Commission (NRC) and can impact a plant's insurance rating. Therefore, there is pressure to return MSPI program components to service as soon as possible after a failure occurs. Three sets of unplanned outages might be used to determine the component repair durations desired in this article: all unplanned outages for the train type that includes the component of

Condition Based Prognostics of Passive Components - A New Era for Nuclear Power Plant Life Management

International Nuclear Information System (INIS)

Bakhtiari, S.; Mohanty, S.; Prokofiev, I.; Tregoning, R.

2012-01-01

As part of a research project sponsored by the U.S. NRC, Argonne National Laboratory (ANL) conducted scoping studies to identify viable and promising sensors and techniques for in-situ inspection and real-time monitoring of degradation in nuclear power plant (NPP) systems, structures, and components (SSC). Significant advances have been made over the past two decades toward development of online monitoring (OLM) techniques for detection, diagnostics, and prognostics of degradation in active nuclear power plant (NPP) components (e.g., pumps, valves). However, early detection of damage and degradation in safety-critical passive components, (e.g. piping, tubing pressure vessel), is challenging, and will likely remain so for the foreseeable future. Ensuring the structural integrity of the reactor pressure vessel (RPV) and piping systems in particular is a prerequisite to long term safe operation of NPPs. The current practice is to implement inservice inspection (ISI) and preventive maintenance programs. While these programs have generally been successful, they are limited in that information is only obtained during plant outages. Additionally, these inspections, often the critical path in the outage schedule, are costly, time consuming, and involve potentially high dose to nondestructive examination/evaluation (NDE) personnel. A viable plant-wide on-line structural health monitoring program for continuous and automatic monitoring of critical SSCs could be a more effective approach for guarding against unexpected failures. Specifically, OLM information about the current condition of the SSCs could be input to an online prognostics (OLP) system to forecast their remaining useful life in real time. This paper provides an overview of scoping studies performed at ANL on assessing the viability of OLM and OLP systems for real time and automated monitoring and remaining of condition and the remaining useful life of passive components in NPPs. (author)

Component Repair Times Obtained from MSPI Data

Energy Technology Data Exchange (ETDEWEB)

Eide, Steven A. [Curtiss-Wright/Scietech, Ketchum, ID (United States); Cadwallader, Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

Information concerning times to repair or restore equipment to service given a failure is valuable to probabilistic risk assessments (PRAs). Examples of such uses in modern PRAs include estimation of the probability of failing to restore a failed component within a specified time period (typically tied to recovering a mitigating system before core damage occurs at nuclear power plants) and the determination of mission times for support system initiating event (SSIE) fault tree models. Information on equipment repair or restoration times applicable to PRA modeling is limited and dated for U.S. commercial nuclear power plants. However, the Mitigating Systems Performance Index (MSPI) program covering all U.S. commercial nuclear power plants provides up-to-date information on restoration times for a limited set of component types. This paper describes the MSPI program data available and analyzes the data to obtain median and mean component restoration times as well as non-restoration cumulative probability curves. The MSPI program provides guidance for monitoring both planned and unplanned outages of trains of selected mitigating systems deemed important to safety. For systems included within the MSPI program, plants monitor both train UA and component unreliability (UR) against baseline values. If the combined system UA and UR increases sufficiently above established baseline results (converted to an estimated change in core damage frequency or CDF), a “white” (or worse) indicator is generated for that system. That in turn results in increased oversight by the US Nuclear Regulatory Commission (NRC) and can impact a plant’s insurance rating. Therefore, there is pressure to return MSPI program components to service as soon as possible after a failure occurs. Three sets of unplanned outages might be used to determine the component repair durations desired in this article: all unplanned outages for the train type that includes the component of interest, only

Selected component failure rate values from fusion safety assessment tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee Charles

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected component failure rate values from fusion safety assessment tasks

International Nuclear Information System (INIS)

Cadwallader, L.C.

1998-01-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers

Auxiliary feedwater system risk-based inspection guide for the Palo Verde Nuclear Power Plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Sloan, J.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Palo Verde was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Palo Verde plants

Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1994-05-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant

Auxiliary feedwater system risk-based inspection guide for the Maine Yankee Nuclear Power Plant

International Nuclear Information System (INIS)

Gore, B.F.; Vo, T.V.; Moffitt, N.E.; Bumgardner, J.D.

1992-10-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. The information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Maine Yankee was selected as one of a series of plants for study. ne product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Maine Yankee plant

Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

International Nuclear Information System (INIS)

Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Vehec, T.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Point Beach was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRS. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Point Beach plant

Failure Modes and Effects Analysis (FMEA) of the Residual Heat Removal System

International Nuclear Information System (INIS)

Eggleston, F.T.

1976-01-01

The Residual Heat Removal System (RHRS) transfer heat from the Reactor Coolant System (RCS) to the reactor plant Component Cooling System (CCS) to reduce the temperature of the RCS at a controlled rate during the second part of normal plant cooldown and maintains the desired temperature until the plant is restarted. By the use of an analytic tool, the Failure Modes and Effects Analysis, it is shown that the RHRS, because of its redundant two train design, is able to accommodate any credible component single failure with the only effect being an extension in the required cooldown time, thus demonstrating the reliability of the RHRS to perform its intended function

Safety classification of nuclear power plant systems, structures and components

International Nuclear Information System (INIS)

1992-01-01

The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

A dynamic failure evaluation of a simplified digital control system of a nuclear power plant pressurizer

International Nuclear Information System (INIS)

Pinto, J.M.O.; Melo, P.F. Frutuoso e; Saldanha, P.L.C.

2010-01-01

Given the increasing use of digital systems in nuclear power plants, a specific approach to reliability and risk analysis has been required. The digital system reflects many interactions between hardware, software, process variables, and human actions. At the same time, the software, does not have a reliability approach as well-defined as the one existing for the other physical components of the system. Then, its reliability analysis is still under development due to difficulties arising from the complexity, flexibility and interactions present in such systems.The traditional approach of using fault trees is static and does not approach the dynamic interactions in such systems, such as delays in capture and processing information, memory, logic loops, system states, etc. It is necessary to find a reliability methodology that takes into account these issues without violating the existing requirements concerning safety analysis, such as: ability to distinguish between common-cause failures, availability of relevant information to users, like minimal cut sets, and failure probabilities as long as the possibility of incorporating the results into existing probabilistic safety assessments (PSA).One approach is to trace all the possible errors of the digital system through dynamic methodologies. The DFM (Dynamic Flow-graph Methodology) is one of the methodologies that better meets the requirements for modeling dynamic systems. It discretizes the most relevant variables of the analyzed system in states that reflect their behavior, sets the logic that connects them through decision tables and finally performs a system analysis, aiming, for example, the root causes (prime implicants) of a given top event of failure. Three aspects have been addressed, the modeling of the system itself, the incorporation of results to probabilistic safety analyses and identification of software failures.To illustrate the DFM, a simplified digital control system of a typical PWR pressurizer

A multi-level maintenance policy for a multi-component and multifailure mode system with two independent failure modes

International Nuclear Information System (INIS)

Zhu, Wenjin; Fouladirad, Mitra; Bérenguer, Christophe

2016-01-01

This paper studies the maintenance modelling of a multi-component system with two independent failure modes with imperfect prediction signal in the context of a system of systems. Each individual system consists of multiple series components and the failure modes of all the components are divided into two classes due to their consequences: hard failure and soft failure, where the former causes system failure while the later results in inferior performance (production reduction) of system. Besides, the system is monitored and can be alerted by imperfect prediction signal before hard failure. Based on an illustration example of offshore wind farm, in this paper three maintenance strategies are considered: periodic routine, reactive and opportunistic maintenance. The periodic routine maintenance is scheduled at fixed period for each individual system in the perspective of system of systems. Between two successive routine maintenances, the reactive maintenance is instructed by the imperfect prediction signal according to two criterion proposed in this study for the system components. Due to the high setup cost and practical restraints of implementing maintenance activities, both routine and reactive maintenance can create the opportunities of maintenance for the other components of an individual system. The life cycle of the system and the cost of the proposed maintenance policies are analytically derived. Restrained by the complexity from both the system failure modelling and maintenance strategies, the performances and application scope of the proposed maintenance model are evaluated by numerical simulations. - Highlights: • We study the life behavior of a complex system with two failure modes. • We consider the imperfect prediction signal of potential failure by monitoring. • We propose an integrated maintenance policy with three levels based on wind turbine. • We derive the mathematical cost formulations for the proposed maintenance policy.

Application of the failure modes and effects analysis technique to the emergency cooling system of an experimental nuclear power plant

International Nuclear Information System (INIS)

Conceicao Junior, Osmar; Silva, Antonio Teixeira e

2009-01-01

This study consists on the application of the failure modes and effects analysis (FMEA), a hazard identification and a risk assessment technique, to the emergency cooling system (ECS), of an experimental nuclear power plant. The choice of this technique was due to its detailed analysis of each component of the system, enabling the identification of all possible ways of failure and its related consequences (in order of importance), allowing the designer to improve the system, maximizing its security and reliability. Through the application of this methodology, it could be observed that the ECS is an intrinsically safe system, in spite of the modifications proposed. (author)

SASSYS-1 balance-of-plant component models for an integrated plant response

International Nuclear Information System (INIS)

Ku, J.-Y.

1989-01-01

Models of power plant heat transfer components and rotating machinery have been added to the balance-of-plant model in the SASSYS-1 liquid metal reactor systems analysis code. This work is part of a continuing effort in plant network simulation based on the general mathematical models developed. The models described in this paper extend the scope of the balance-of-plant model to handle non-adiabatic conditions along flow paths. While the mass and momentum equations remain the same, the energy equation now contains a heat source term due to energy transfer across the flow boundary or to work done through a shaft. The heat source term is treated fully explicitly. In addition, the equation of state is rewritten in terms of the quality and separate parameters for each phase. The models are simple enough to run quickly, yet include sufficient detail of dominant plant component characteristics to provide accurate results. 5 refs., 16 figs., 2 tabs

Inductive analysis of failure patterns and of their impact on thermohydraulic circuits of nuclear power plants

International Nuclear Information System (INIS)

Limnios, N.

1983-01-01

The APACHE code (Automatic Analysis of Failures of Hydraulic and Thermohydraulic Circuits more particularly of Water) situates in an important program of computer codes development in the field of studies on reliability and safety of systems in nuclear power plants. APACHE is an automatic generation code of failure pattern and of their effects. After a presentation of the theoretical basis, the methodological principles of the theory of networks are developed. Then, the model of the code is developed: model of individual behavior of each classical model component of normal behavior and model of failure pattern with specifications. The global model of hydraulic systems and the resolution systems are then developed. More particularly, some aspects of the theory of graphs, and the algorithms developed for the automatic construction of the equation systems and especially the algorithm of the research of meshes are presented. The computer aspect of the code and the programming of the code with its limits and some specifications are described. The practical aspect of utilization is finally presented [fr

IAEA's experience in compiling a generic component reliability data base

International Nuclear Information System (INIS)

Tomic, B.; Lederman, L.

1988-01-01

Reliability data are an essential part of probabilistic safety assessment. The quality of data can determine the quality of the study as a whole. It is obvious that component failure data originated from the plant being analyzed would be most appropriate. However, in few cases complete reliance on plant experience is possible, mainly because of the rather limited operating experience. Nuclear plants, although of different design, often use fairly similar components, so some of the experience could be combined and transferred from one plant to another. In addition information about component failures is available also from experts with knowledge on component design, manufacturing and operation. That bring us to the importance of assessing generic data. (Generic is meant to be everything that is not plant specific regarding the plant being analyzed). The generic data available in the open literature, can be divided in three broad categories. The first one includes data base used in previous analysis. These can be plant specific or updated from generic with plant specific information (latter case deserve special attention). The second one is based on compilation of plants' operating experience usually based on some kind of event reporting system. The third category includes data sources based on expert opinions (single or aggregate) or combination of expert opinions and other nuclear and non-nuclear experience. This paper reflects insights gained in compiling data from generic data sources and highlights advantages and pitfalls of using generic component reliability data in PSAs

Trend evaluation of incident and failure data from japanese nuclear power plants

International Nuclear Information System (INIS)

Kondo, S.; Hada, M.; Mikami, Y.

1990-01-01

Major incident and failure at nuclear power plants in Japan have to be reported to the regulatory agency i.e. Ministry of International Trade and Industry (MITI). Nuclear Power Safety Information Research Center (NUSIRC) has established a system for the collection, classification and analysis of this report under the contract to MITI. In this paper, the authors give several results of trend analyses of the incidents related to electric and instrumentation and control (I and C) systems reported, especially, the trend of the contribution of troubles in I and C system to the operation states, analysis of dominant contributors to the failure of I and C systems. Also, the relations of failure frequency of these systems with the plant age and effect of periodic inspections of it are discussed in some detail

Bayesian methodology for generic seismic fragility evaluation of components in nuclear power plants

International Nuclear Information System (INIS)

Yamaguchi, Akira; Campbell, R.D.; Ravindra, M.K.

1991-01-01

Bayesian methodology for updating the seismic fragility of components in nuclear power plants is presented. The generic fragility data which have been evaluated based on the past SPSAs are combined with the seismic experience data. Although the seismic experience is limited to the acceleration range below the median capacity of the components, it has been found that the evidence is effective to update the fragility tail. In other words, the uncertainty of the fragility is reduced although the median capacity itself is not modified to a great extent. The annual frequency of failure is also reduced as a result of the updating of the fragility tail. The PDF of the seismic capacity is handled in discrete form, which enables the use of arbitrary type of prior distribution. Accordingly, the Log-N prior can be used which is consistent with the widely used fragility model. For evaluating posterior fragility parameters (A m and B U ), two methods have been proposed. Furthermore, it has been found that the importance of evidence used in the Bayesian methodology can be quantified by the entropy of the evidence. Only the events with high entropy need to be considered in the Bayesian updating of the fragility. The currently available seismic experience database for typical components can be utilized to develop the fragility tail which is contributive to the seismically-induced failure frequency. The combined use of generic fragility and seismic experience data, with the aid of Bayesian methodology, provides refined generic fragility curves which are useful for SPSA studies. (author)

Failure Investigation of Radiant Platen Superheater Tube of Thermal Power Plant Boiler

Science.gov (United States)

Ghosh, D.; Ray, S.; Mandal, A.; Roy, H.

2015-04-01

This paper highlights a case study of typical premature failure of a radiant platen superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement and chemical analysis, are conducted as part of the investigations. Apart from these, metallographic analysis and fractography are also conducted to ascertain the probable cause of failure. Finally it has been concluded that the premature failure of the super heater tube can be attributed to localized creep at high temperature. The corrective actions has also been suggested to avoid this type of failure in near future.

The Component And System Reliability Analysis Of Multipurpose Reactor G.A. Subway's Based On The Failure Rate Curve

International Nuclear Information System (INIS)

Sriyono; Ismu Wahyono, Puradwi; Mulyanto, Dwijo; Kusmono, Siamet

2001-01-01

The main component of Multipurpose G.A.Siwabessy had been analyzed by its failure rate curve. The main component ha'..e been analyzed namely, the pump of ''Fuel Storage Pool Purification System'' (AK-AP), ''Primary Cooling System'' (JE01-AP), ''Primary Pool Purification System'' (KBE01-AP), ''Warm Layer System'' (KBE02-AP), ''Cooling Tower'' (PA/D-AH), ''Secondary Cooling System'', and Diesel (BRV). The Failure Rate Curve is made by component database that was taken from 'log book' operation of RSG GAS. The total operation of that curve is 2500 hours. From that curve it concluded that the failure rate of components form of bathtub curve. The maintenance processing causes the curve anomaly

Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1993-12-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant

Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

International Nuclear Information System (INIS)

Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V.; Garner, L.W.

1993-08-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant

Component protection based automatic control

International Nuclear Information System (INIS)

Otaduy, P.J.

1992-01-01

Control and safety systems as well as operation procedures are designed on the basis of critical process parameters limits. The expectation is that short and long term mechanical damage and process failures will be avoided by operating the plant within the specified constraints envelopes. In this paper, one of the Advanced Liquid Metal Reactor (ALMR) design duty cycles events is discussed to corroborate that the time has come to explicitly make component protection part of the control system. Component stress assessment and aging data should be an integral part of the control system. Then transient trajectory planning and operating limits could be aimed at minimizing component specific and overall plant component damage cost functions. The impact of transients on critical components could then be managed according to plant lifetime design goals. The need for developing methodologies for online transient trajectory planning and assessment of operating limits in order to facilitate the explicit incorporation of damage assessment capabilities to the plant control and protection systems is discussed. 12 refs

Reliability Data Handbook for Piping Components in Nordic Nuclear Power Plants - R Book, Phase 2

International Nuclear Information System (INIS)

Hedtjaern Swaling, Vidar; Olsson, Anders

2011-02-01

This report presents results of a long research and development project financed by the regulatory body Straalsaekerhetsmyndigheten (SSM) (former SKI), the Swedish nuclear power plant licensees. The report presents a harmonized method for estimating Reliability Data for Piping Components in ASME code class 1 and 2 piping components (R-Book). Data in the R-Book is measured based on 'data driven' strategy. This first version of the R-Book comprises rupture frequencies and failure rates for all systems where ASME Code Class 1 or 2 events could be found in the OECD OPDE database. Nordic and Non-Nordic data are presented separately. Worldwide experience data is used to set up the relevant calculation cases, i.e. intersections of attributes for which there are at least one event present

Interim report on the state-of-the-art of solid-state motor controllers. Part 4. Failure-rate and failure-mode data

International Nuclear Information System (INIS)

Jaross, R.A.

1983-09-01

An assessment of the reliability of solid-state motor controllers for nuclear power plants is made. Available data on failure-rate and failure-mode data for solid-state motor controllers based on industrial operating experience is meager; the data are augmented by data on other solid-state power electronic devices that are shown to have components similar to those found in solid-state motor controllers. In addition to large nonnuclear solid-state adjustable-speed motor drives, the reliability of nuclear plant inverter systems and high-voltage solid-state dc transmission-line converters is assessed. Licensee Event Report analyses from several sources, the open literature, and personal communications are used to determine the realiability of solid-state devices typical of those expected to be used in nuclear power plants in terms of failures per hour

Role of scanning electron microscope )SEM) in metal failure analysis

International Nuclear Information System (INIS)

Shaiful Rizam Shamsudin; Hafizal Yazid; Mohd Harun; Siti Selina Abd Hamid; Nadira Kamarudin; Zaiton Selamat; Mohd Shariff Sattar; Muhamad Jalil

2005-01-01

Scanning electron microscope (SEM) is a scientific instrument that uses a beam of highly energetic electrons to examine the surface and phase distribution of specimens on a micro scale through the live imaging of secondary electrons (SE) and back-scattered electrons (BSE) images. One of the main activities of SEM Laboratory at MINT is for failure analysis on metal part and components. The capability of SEM is excellent for determining the root cause of metal failures such as ductility or brittleness, stress corrosion, fatigue and other types of failures. Most of our customers that request for failure analysis are from local petrochemical plants, manufacturers of automotive components, pipeline maintenance personnel and engineers who involved in the development of metal parts and component. This paper intends to discuss some of the technical concepts in failure analysis associated with SEM. (Author)

Application of nonhomogeneous Poisson process to reliability analysis of repairable systems of a nuclear power plant with rates of occurrence of failures time-dependent

International Nuclear Information System (INIS)

Saldanha, Pedro L.C.; Simone, Elaine A. de; Melo, Paulo Fernando F.F. e

1996-01-01

Aging is used to mean the continuous process which physical characteristics of a system, a structure or an equipment changes with time or use. Their effects are increases in failure probabilities of a system, a structure or an equipment, and their are calculated using time-dependent failure rate models. The purpose of this paper is to present an application of the nonhomogeneous Poisson process as a model to study rates of occurrence of failures when they are time-dependent. To this application, an analysis of reliability of service water pumps of a typical nuclear power plant is made, as long as the pumps are effectively repaired components. (author)

Redundancy allocation problem of a system with increasing failure rates of components based on Weibull distribution: A simulation-based optimization approach

International Nuclear Information System (INIS)

Guilani, Pedram Pourkarim; Azimi, Parham; Niaki, S.T.A.; Niaki, Seyed Armin Akhavan

2016-01-01

The redundancy allocation problem (RAP) is a useful method to enhance system reliability. In most works involving RAP, failure rates of the system components are assumed to follow either exponential or k-Erlang distributions. In real world problems however, many systems have components with increasing failure rates. This indicates that as time passes by, the failure rates of the system components increase in comparison to their initial failure rates. In this paper, the redundancy allocation problem of a series–parallel system with components having an increasing failure rate based on Weibull distribution is investigated. An optimization method via simulation is proposed for modeling and a genetic algorithm is developed to solve the problem. - Highlights: • The redundancy allocation problem of a series–parallel system is aimed. • Components possess an increasing failure rate based on Weibull distribution. • An optimization method via simulation is proposed for modeling. • A genetic algorithm is developed to solve the problem.

Component design considerations for gas turbine HTGR waste-heat power plant

International Nuclear Information System (INIS)

McDonald, C.F.; Vrable, D.L.

1976-01-01

Component design considerations are described for the ammonia waste-heat power conversion system of a large helium gas-turbine nuclear power plant under development by General Atomic Company. Initial component design work was done for a reference plant with a 3000-MW(t) High-Temperature Gas-Cooled Reactor (HTGR), and this is discussed. Advanced designs now being evaluated include higher core outlet temperature, higher peak system pressures, improved loop configurations, and twin 4000-MW(t) reactor units. Presented are the design considerations of the major components (turbine, condenser, heat input exchanger, and pump) for a supercritical ammonia Rankine waste heat power plant. The combined cycle (nuclear gas turbine and waste-heated plant) has a projected net plant efficiency of over 50 percent. While specifically directed towards a nuclear closed-cycle helium gas-turbine power plant (GT-HTGR), it is postulated that the bottoming waste-heat cycle component design considerations presented could apply to other low-grade-temperature power conversion systems such as geothermal plants

Detection of sensor failures in nuclear plants using analytic redundancy

International Nuclear Information System (INIS)

Kitamura, M.

1980-01-01

A method for on-line, nonperturbative detection and identification of sensor failures in nuclear power plants was studied to determine its feasibility. This method is called analytic redundancy, or functional redundancy. Sensor failure has traditionally been detected by comparing multiple signals from redundant sensors, such as in two-out-of-three logic. In analytic redundancy, with the help of an assumed model of the physical system, the signals from a set of sensors are processed to reproduce the signals from all system sensors

Best of failure analysis of turbomachinery components. Highlights from two decades' of laboratory practice; Best of Schadensanalyse an Turbomaschinen. Die Highlights aus 20 Jahren Laborpraxis

Energy Technology Data Exchange (ETDEWEB)

Neidel, Andreas; Cagliyan, Erhan; Gaedicke, Tobias; Giller, Madleine; Hartanto, Vincentius; Kramm, Christine; Riesenbeck, Susanne; Ullrich, Thomas; Wallich, Sebastian; Woehl, Eric [Siemens AG, Power and Gas, Berlin (Germany). Werkstoffprueflabor

2017-01-15

In this contribution, the most interesting and educational failure cases are presented that the author came across during his over twenty years of laboratory practice as manager of the Materials Testing Laboratory of the Berlin Gas Turbine Plant of Siemens' Power and Gas Division. The case studies are presented and categorised in accordance with VDI Guideline 3822, the German failure analyst's guide to the subject of how to organise and run a root cause failure analysis. An effort was made to have each of the main four categories of failure causes represented, namely failures due to mechanical loading, corrosive failures, failures due to thermal loading, and tribological failures. Case studies include turbomachinery components that failed due to tensile overload, stress corrosion cracking, intergranular corrosion, hydrogen embrittlement, hot cracking, fretting, erosion, and galling. Affected components include valves, retaining rings, tubing and piping, burners, rotor disks, lifting lugs, and casings. Some of the presented cases were published in the new section ''Failure Analysis'' of Practical Metallography between October 2011 and the present time. Others were oral presentations at the Metallography conferences and at the annual failure analysis conferences ''VDI Jahrestagung Schadensanalyse'', held during that time. The focus of discussion of the failure cases in this paper is the metallurgical evaluation of failure causes. This is the approach taken in many small and industrial laboratories. A holistic approach of a failure case, which includes calculation and simulation methods such as finite element analysis, and which also implies a knowledge of the service stresses intended by design as well as the actual loading situation of the failed part, is not the aim of this contribution.

Common cause failure rate estimates for diesel generators in nuclear power plants

International Nuclear Information System (INIS)

Steverson, J.A.; Atwood, C.L.

1982-01-01

Common cause fault rates for diesel generators in nuclear power plants are estimated, using Licensee Event Reports for the years 1976 through 1978. The binomial failure rate method, used for obtaining the estimates, is briefly explained. Issues discussed include correct classification of common cause events, grouping of the events into homogeneous data subsets, and dealing with plant-to-plant variation

A study on the water chemistry in nuclear power plants

International Nuclear Information System (INIS)

Chae, Sung Ki; Yang, Kyung Rin; Koo Je Hyoo; Lee, Eun Hee; Kim, Joung Soo; Jang, Soon Shik; Park, Su Hoon; Song, Myung Ho; Jeon, Kyung Soo

1987-12-01

Significant corrosion-failures occurring in the important components or facilities in the secondary-side system cause various problems in safety due to the leakage of radioactive substances and consequently induce the reduction of the operational efficiency of the plants. In addition, the replacement of the failed components or facilities results in the tremendous expenses and a long term shutdown. The objective of the research was to ensure the safety and integrity of the plants, to improve the efficiency of the plant operation, and to prevent the shortening of plant life by improving the controlling technique of the water chemistry and minimizing the corrosion-failures in the important components and/or facilities of the plants

Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

International Nuclear Information System (INIS)

Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1991-07-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab

Quantitative relationships between aging failure data and risk

International Nuclear Information System (INIS)

Vesely, W.E.; Vora, J.P.

1988-01-01

As part of the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research program, a project is being carried out to quantify the risk effects of aging. The project is called the Risk Evaluation of Aging Phenomena (REAP) Project. With the REAP Project, a procedure has been developed to quantify nuclear power plant risks from aging failure data. The procedure utilizes the linear aging model and its extensions in order to relate component aging failure rates to aging mechanism parameters which are estimable from failure and maintenance data. The aging failure rates can then be used to quantify the age dependent plant risks, such as system unavailabilities, core melt frequency and public health risks. The REAP procedure is different from standard time dependent approaches in that the failure rates are phenomenologically based, allowing engineering information to be utilized. Furthermore, gross data and incomplete data can be utilized. A software package has been developed which systematically analyzes data for aging effects and interfaces with a time dependent risk analysis module to determine the risk implications of the aging effects. (author). 10 refs, 10 figs

Ventilation systems and components of nuclear power plants

International Nuclear Information System (INIS)

1997-01-01

The most important radiation and nuclear safety requirements for the design and manufacture of nuclear power plant ventilation systems and components are presented in the guide. Also the regulatory activities of the Finnish Centre for Radiation and Nuclear Safety (STUK) as regards the ventilation systems and components are explained. Documents and data which shall be submitted to STUK during the various phases of the regulatory procedure relating to the design, construction, commissioning and operation of the nuclear power plants are presented. (13 refs.)

Application of environmentally-corrected fatigue curves to nuclear power plant components

International Nuclear Information System (INIS)

Ware, A.G.; Morton, D.K.; Nitzel, M.E.

1996-01-01

Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four US nuclear steam supply system vendors. For each facility, six locations were studied including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This paper discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

Failure analysis and success analysis: roles in plant aging assessments

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1985-06-01

Component aging investigations are an important element in NRC's Nuclear Plant Aging Research (NPAR) strategy. Potential sources of components include plants in decommissioning and commercial plant, both for in situ tests and for examination of equipment removed from service. Nuclear utilities currently have voluntary programs addressing aspects of equipment reliability, such as root cause analysis for safety-related equipment that malfunctions, and trending analysis to follow the course of both successful and abnormal equipment performance. Properly coordinated, the NPAR and utility programs offer an important approach to establish the data base necessary for life extension of nuclear electrical generating plants

A CANDU designed for more tolerance to failures in large components

International Nuclear Information System (INIS)

Spinks, N.J.; Barclay, F.W.; Allen, P.J.; Yee, F.

1988-06-01

Current designs of CANDU reactors have several groups of fuel channels each served by an upstream coolant supply-train consisting of an outlet header, a steam generator, one or more pumps in parallel and an inlet header. Postulated failures in these large components put the heaviest demands on the safety systems. For example, the rupture of a header sets the requirements for the speed of shutdown and for the speed and capacity of emergency coolant injection, and it has a large impact on containment design. A CANDU design is being investigated to reduce the impact of failures in large components. Each group of fuel channels is supplied by more than one train so that if one train fails the rest continue to work. Reverse flow limiters reduce the loss-of-coolant from the unbroken trains to a broken supply train. The paper describes several design options for making the piping connections from multi supply-trains to fuel channels. It discusses progress in design and testing of flow limiters. A preliminary analysis is given of affected accidents

Plant systems/components modularization study. Final report

International Nuclear Information System (INIS)

1977-07-01

The final results are summarized of a Plant Systems/Components Modularization Study based on Stone and Webster's Pressurized Water Reactor Reference Design. The program has been modified to include evaluation of the most promising areas for modular consideration based on the level of the Sundesert Project engineering design completion and the feasibility of their incorporation into the plant construction effort

Analysis on Single Point Vulnerabilities of Plant Control System

International Nuclear Information System (INIS)

Chi, Moon Goo; Lee, Eun Chan; Bae, Yeon Kyoung

2011-01-01

The Plant Control System (PCS) is a system that controls pumps, valves, dampers, etc. in nuclear power plants with an OPR-1000 design. When there is a failure or spurious actuation of the critical components in the PCS, it can result in unexpected plant trips or transients. From this viewpoint, single point vulnerabilities are evaluated in detail using failure mode effect analyses (FMEA) and fault tree analyses (FTA). This evaluation demonstrates that the PCS has many vulnerable components and the analysis results are provided for OPR-1000 plants for reliability improvements that can reduce their vulnerabilities

Analysis on Single Point Vulnerabilities of Plant Control System

Energy Technology Data Exchange (ETDEWEB)

Chi, Moon Goo; Lee, Eun Chan; Bae, Yeon Kyoung [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2011-08-15

The Plant Control System (PCS) is a system that controls pumps, valves, dampers, etc. in nuclear power plants with an OPR-1000 design. When there is a failure or spurious actuation of the critical components in the PCS, it can result in unexpected plant trips or transients. From this viewpoint, single point vulnerabilities are evaluated in detail using failure mode effect analyses (FMEA) and fault tree analyses (FTA). This evaluation demonstrates that the PCS has many vulnerable components and the analysis results are provided for OPR-1000 plants for reliability improvements that can reduce their vulnerabilities.

Expert environment for the development of nuclear power plants failure diagnosis systems

International Nuclear Information System (INIS)

Guido, P.N.; Oggianu, S.; Etchepareborda, A.; Fernandez, O.

1996-01-01

The present work explores some of the developing stages of an Expert Environment for plant failures Diagnosis Systems starting from Knowledge Based Systems. We present a prototype that carries out an inspection of anomalous symptoms and a diagnosis process based on a Plant Abnormality Model of a PHWR secondary system

Extending the life of thermal power plants components by using the triad: checking - diagnosis - restoring; Extinderea duratei de viata a componentelor termoenergetice utilizand Triada: expertiza - diagnoza - remediere

Energy Technology Data Exchange (ETDEWEB)

Lupescue, L.; Nicolescu, N.; Delamarian, C.

2004-12-01

The current state of thermal power plants components requires a clear cut definition of how to apply the concept of 'Assessing the state and lifetime' to them. The application of the concept of prolonging the life of thermal power plants components represents a practical alternative to the activity of preventive maintenance, by using 'the component oriented maintenance'. A feasibility study gives results that are a covering prediction for the level of technical risk characterizing the operation of technological equipment. There are many methods of analysing the dangers and assessing the risk, as two basic types can be established: a deductive one, in which the final event is presupposed and the events that might cause this final event are searched, and an inductive one, in which the failure of a component is presupposed, and the analysis is to identify the events that led to failure. In compliance with the worldwide trends the authors of the present paper make efforts to apply the methods specific of the probability analyses, efficiently and harmoniously supplementing the concerns and results of the activities based on deterministic methods and models. 13 refs., 1 fig.

Phenomenological uncertainty analysis of early containment failure at severe accident of nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Su Won

2011-02-15

The severe accident has inherently significant uncertainty due to wide range of conditions and performing experiments, validation and practical application are extremely difficult because of its high temperature and pressure. Although internal and external researches were put into practice, the reference used in Korean nuclear plants were foreign data of 1980s and safety analysis as the probabilistic safety assessment has not applied the newest methodology. Also, it is applied to containment pressure formed into point value as results of thermal hydraulic analysis to identify the probability of containment failure in level 2 PSA. In this paper, the uncertainty analysis methods for phenomena of severe accident influencing early containment failure were developed, the uncertainty analysis that apply Korean nuclear plants using the MELCOR code was performed and it is a point of view to present the distribution of containment pressure as a result of uncertainty analysis. Because early containment failure is important factor of Large Early Release Frequency(LERF) that is used as representative criteria of decision-making in nuclear power plants, it was selected in this paper among various modes of containment failure. Important phenomena of early containment failure at severe accident based on previous researches were comprehended and methodology of 7th steps to evaluate uncertainty was developed. The MELCOR input for analysis of the severe accident reflected natural circulation flow was developed and the accident scenario for station black out that was representative initial event of early containment failure was determined. By reviewing the internal model and correlation for MELCOR model relevant important phenomena of early containment failure, the uncertainty factors which could affect on the uncertainty were founded and the major factors were finally identified through the sensitivity analysis. In order to determine total number of MELCOR calculations which can

Basic factors to forecast maintenance cost and failure processes for nuclear power plants

International Nuclear Information System (INIS)

Popova, Elmira; Yu, Wei; Kee, Ernie; Sun, Alice; Richards, Drew; Grantom, Rick

2006-01-01

Two types of maintenance interventions are usually administered at nuclear power plants: planned and corrective. The cost incurred includes the labor (manpower) cost, cost for new parts, or emergency order of expensive items. At the plant management level there is a budgeted amount of money to be spent every year for such operations. It is very important to have a good forecast for this cost since unexpected events can trigger it to a very high level. In this research we present a statistical factor model to forecast the maintenance cost for the incoming month. One of the factors is the expected number of unplanned (due to failure) maintenance interventions. We introduce a Bayesian model for the failure rate of the equipment, which is input to the cost forecasting model. The importance of equipment reliability and prediction in the commercial nuclear power plant is presented along with applicable governmental and industry organization requirements. A detailed statistical analysis is performed on a set of maintenance cost and failure data gathered at the South Texas Project Nuclear Operating Company (STPNOC) in Bay City, Texas, USA

Thermal Power Plant Performance Analysis

CERN Document Server

2012-01-01

The analysis of the reliability and availability of power plants is frequently based on simple indexes that do not take into account the criticality of some failures used for availability analysis. This criticality should be evaluated based on concepts of reliability which consider the effect of a component failure on the performance of the entire plant. System reliability analysis tools provide a root-cause analysis leading to the improvement of the plant maintenance plan.   Taking in view that the power plant performance can be evaluated not only based on  thermodynamic related indexes, such as heat-rate, Thermal Power Plant Performance Analysis focuses on the presentation of reliability-based tools used to define performance of complex systems and introduces the basic concepts of reliability, maintainability and risk analysis aiming at their application as tools for power plant performance improvement, including: ·         selection of critical equipment and components, ·         defini...

Evolutionary conservation of plant gibberellin signalling pathway components

Directory of Open Access Journals (Sweden)

Reski Ralf

2007-11-01

Full Text Available Abstract Background: Gibberellins (GA are plant hormones that can regulate germination, elongation growth, and sex determination. They ubiquitously occur in seed plants. The discovery of gibberellin receptors, together with advances in understanding the function of key components of GA signalling in Arabidopsis and rice, reveal a fairly short GA signal transduction route. The pathway essentially consists of GID1 gibberellin receptors that interact with F-box proteins, which in turn regulate degradation of downstream DELLA proteins, suppressors of GA-controlled responses. Results: Arabidopsis sequences of the gibberellin signalling compounds were used to screen databases from a variety of plants, including protists, for homologues, providing indications for the degree of conservation of the pathway. The pathway as such appears completely absent in protists, the moss Physcomitrella patens shares only a limited homology with the Arabidopsis proteins, thus lacking essential characteristics of the classical GA signalling pathway, while the lycophyte Selaginella moellendorffii contains a possible ortholog for each component. The occurrence of classical GA responses can as yet not be linked with the presence of homologues of the signalling pathway. Alignments and display in neighbour joining trees of the GA signalling components confirm the close relationship of gymnosperms, monocotyledonous and dicotyledonous plants, as suggested from previous studies. Conclusion: Homologues of the GA-signalling pathway were mainly found in vascular plants. The GA signalling system may have its evolutionary molecular onset in Physcomitrella patens, where GAs at higher concentrations affect gravitropism and elongation growth.

Parameter Estimation of a Reliability Model of Demand-Caused and Standby-Related Failures of Safety Components Exposed to Degradation by Demand Stress and Ageing That Undergo Imperfect Maintenance

Directory of Open Access Journals (Sweden)

S. Martorell

2017-01-01

Full Text Available One can find many reliability, availability, and maintainability (RAM models proposed in the literature. However, such models become more complex day after day, as there is an attempt to capture equipment performance in a more realistic way, such as, explicitly addressing the effect of component ageing and degradation, surveillance activities, and corrective and preventive maintenance policies. Then, there is a need to fit the best model to real data by estimating the model parameters using an appropriate tool. This problem is not easy to solve in some cases since the number of parameters is large and the available data is scarce. This paper considers two main failure models commonly adopted to represent the probability of failure on demand (PFD of safety equipment: (1 by demand-caused and (2 standby-related failures. It proposes a maximum likelihood estimation (MLE approach for parameter estimation of a reliability model of demand-caused and standby-related failures of safety components exposed to degradation by demand stress and ageing that undergo imperfect maintenance. The case study considers real failure, test, and maintenance data for a typical motor-operated valve in a nuclear power plant. The results of the parameters estimation and the adoption of the best model are discussed.

Prestudy - Development of trend analysis of component failure

International Nuclear Information System (INIS)

Poern, K.

1995-04-01

The Bayesian trend analysis model that has been used for the computation of initiating event intensities (I-book) is based on the number of events that have occurred during consecutive time intervals. The model itself is a Poisson process with time-dependent intensity. For the analysis of aging it is often more relevant to use times between failures for a given component as input, where by 'time' is meant a quantity that best characterizes the age of the component (calendar time, operating time, number of activations etc). Therefore, it has been considered necessary to extend the model and the computer code to allow trend analysis of times between events, and also of several sequences of times between events. This report describes this model extension as well as an application on an introductory ageing analysis of centrifugal pumps defined in Table 5 of the T-book. The application in turn directs the attention to the need for further development of both the trend model and the data base. Figs

Tube failures due to cooling process problem and foreign materials in power plants

Energy Technology Data Exchange (ETDEWEB)

Ahmad, J. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia); Purbolaksono, J., E-mail: judha@uniten.edu.m [Department of Mechanical Engineering, Universiti Tenaga Nasional, Km 7 Jalan Kajang-Puchong, Kajang 43009, Selangor (Malaysia); Beng, L.C. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia)

2010-07-15

Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

Tube failures due to cooling process problem and foreign materials in power plants

International Nuclear Information System (INIS)

Ahmad, J.; Purbolaksono, J.; Beng, L.C.

2010-01-01

Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

Failure analysis and failure prevention in electric power systems

International Nuclear Information System (INIS)

Rau, C.A. Jr.; Becker, D.G.; Besuner, P.M.; Cipolla, R.C.; Egan, G.R.; Gupta, P.; Johnson, D.P.; Omry, U.; Tetelman, A.S.; Rettig, T.W.; Peters, D.C.

1977-01-01

New methods have been developed and applied to better quantify and increase the reliability, safety, and availability of electric power plants. Present and potential problem areas have been identified both by development of an improved computerized data base of malfunctions in nuclear power plants and by detailed metallurgical and mechanical failure analyses of selected problems. Significant advances in the accuracy and speed of structural analyses have been made through development and application of the boundary integral equation and influence function methods of stress and fracture mechanics analyses. The currently specified flaw evaluation procedures of the ASME Boiler and Pressure Vessel Code have been computerized. Results obtained from these procedures for evaluation of specific in-service inspection indications have been compared with results obtained utilizing the improved analytical methods. Mathematical methods have also been developed to describe and analyze the statistical variations in materials properties and in component loading, and uncertainties in the flaw size that might be passed by quality assurance systems. These new methods have been combined to develop accurate failure rate predictions based upon probabilistic fracture mechanics. Improved failure prevention strategies have been formulated by combining probabilistic fracture mechanics and cost optimization techniques. The approach has been demonstrated by optimizing the nondestructive inspection level with regard to both reliability and cost. (Auth.)

Preliminary Analysis of the Common Cause Failure Events for Domestic Nuclear Power Plants

International Nuclear Information System (INIS)

Kang, Daeil; Han, Sanghoon

2007-01-01

It is known that the common cause failure (CCF) events have a great effect on the safety and probabilistic safety assessment (PSA) results of nuclear power plants (NPPs). However, the domestic studies have been mainly focused on the analysis method and modeling of CCF events. Thus, the analysis of the CCF events for domestic NPPs were performed to establish a domestic database for the CCF events and to deliver them to the operation office of the international common cause failure data exchange (ICDE) project. This paper presents the analysis results of the CCF events for domestic nuclear power plants

Interpretation of risk significance of passive component aging using probabilistic structural analysis

International Nuclear Information System (INIS)

Phillips, J.H.; Atwood, C.L.

1993-01-01

The probabilistic risk assessments (PRAs) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. Except as initiating events, the possible failure of passive components is given little consideration. The NRC is sponsoring a project at INEL to investigate the risk significance of passive components as they age. For this project, we developed a technique to calculate the failure probability of passive components over time, and demonstrated the technique by applying it to a weld in the auxiliary feedwater (AFW) system. A decreasing yearly rupture rate for this weld was calculated instead of the increasing rupture rate trend one might expect. We attribute this result to infant mortality; that is, most of those initial flaws that will eventually lead to rupture will do so early in life. This means that although each weld in a population may be wearing out, the population as a whole can exhibit a decreasing rupture rate. This observation has implications for passive components in commercial nuclear plants and other facilities where aging is a concern. For the population of passive components that exhibit a decreasing failure rate, risk increase is not a concern. The next step of the work is to identify the attributes that contribute to this decreasing rate, and to determine any attributes that would contribute to an increasing failure rate and thus to an increased risk

Combinatorial analysis of systems with competing failures subject to failure isolation and propagation effects

International Nuclear Information System (INIS)

Xing Liudong; Levitin, Gregory

2010-01-01

This paper considers the reliability analysis of binary-state systems, subject to propagated failures with global effect, and failure isolation phenomena. Propagated failures with global effect are common-cause failures originated from a component of a system/subsystem causing the failure of the entire system/subsystem. Failure isolation occurs when the failure of one component (referred to as a trigger component) causes other components (referred to as dependent components) within the same system to become isolated from the system. On the one hand, failure isolation makes the isolated dependent components unusable; on the other hand, it prevents the propagation of failures originated from those dependent components. However, the failure isolation effect does not exist if failures originated in the dependent components already propagate globally before the trigger component fails. In other words, there exists a competition in the time domain between the failure of the trigger component that causes failure isolation and propagated failures originated from the dependent components. This paper presents a combinatorial method for the reliability analysis of systems subject to such competing propagated failures and failure isolation effect. Based on the total probability theorem, the proposed method is analytical, exact, and has no limitation on the type of time-to-failure distributions for the system components. An illustrative example is given to demonstrate the basics and advantages of the proposed method.

A discussion about simplified methodologies for failure assessment of nuclear reactor components

International Nuclear Information System (INIS)

Cruz, J.R.B.; Andrade, A.H.P. de; Landes, J.D.

1996-01-01

Failure of nuclear reactor components like pressure vessels and piping must be avoided for all phases of reactor operation. Especially severe loading conditions come from postulated accident scenarios during which the integrity of the component is required. The use of Fracture Mechanics concepts to investigate the mechanical behavior of flawed structures in the non-linear regime is a complex subject due to the fact that the crack driving force (expressed in terms of J or CTOD) is not /only a function of the cracked geometry, but depends also on the plastic flow properties of the material. Since the numerical solutions by the finite element method are expensive and time consuming, the existence of simplified engineering procedures is of great relevance. These allow a ready identification of the main parameters affecting the crack driving force, and permit a fast and simple evaluation of the structural integrity of the cracked component. This paper presents an overview of the major simplified ductile fracture methodologies that have been proposed in the literature trying to point out their similarities, strong points and negative aspects. Once the best characteristics of each method are identified, they could then be combined to develop a single methodology, one that would be both easy to use and capable of making accurate failure predictions

Failure of Titanium Condenser Tubes after 24 Years Power Plant Service

DEFF Research Database (Denmark)

Montgomery, Melanie; Enemark, Allan; Hangaard, Anders

2014-01-01

The titanium condenser has been in operation for 24 years at Amager unit 3 power plant. In February 2012, the plant was contaminated by seawater due to a failed condenser tube and some tubes were plugged. A month later, the plant tripped again. Small leaks were found again and finally approx. 200...... a plant trip. In addition, small amounts of titanium hydride were revealed to be present in the tubes within the tubesheet indicating that the carbon steel tubesheet was corroding due to ingress of salt water. Although this was not the reason for the failure, it indicated the need for repair of the epoxy...

Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

International Nuclear Information System (INIS)

Dongiovanni, Danilo Nicola; Iesmantas, Tomas

2016-01-01

Highlights: • RAMI (Reliability, Availability, Maintainability and Inspectability) assessment of secondary heat transfer loop for a DEMO nuclear fusion plant. • Definition of a fault tree for a nuclear steam turbine operated in pulsed mode. • Turbine failure rate models update by mean of a Bayesian network reflecting the fault tree analysis in the considered scenario. • Sensitivity analysis on system availability performance. - Abstract: Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability

Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

Energy Technology Data Exchange (ETDEWEB)

Dongiovanni, Danilo Nicola, E-mail: danilo.dongiovanni@enea.it [ENEA, Nuclear Fusion and Safety Technologies Department, via Enrico Fermi 45, Frascati 00040 (Italy); Iesmantas, Tomas [LEI, Breslaujos str. 3 Kaunas (Lithuania)

2016-11-01

Highlights: • RAMI (Reliability, Availability, Maintainability and Inspectability) assessment of secondary heat transfer loop for a DEMO nuclear fusion plant. • Definition of a fault tree for a nuclear steam turbine operated in pulsed mode. • Turbine failure rate models update by mean of a Bayesian network reflecting the fault tree analysis in the considered scenario. • Sensitivity analysis on system availability performance. - Abstract: Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability

Noise diagnosis - a method for early detection of failures in a nuclear plant

International Nuclear Information System (INIS)

Brinckmann, H.F.

1981-01-01

Noise diagnosis constitutes one method for early detection of plant failures. The method is based on the fact that nearly all undesired processes in a nuclear power plant make a measurable contribution to the noise portion of signals. Well-known examples of undesired processes in pressurized water reactors include core-barrel movement, the vibration of control elements, the appearance of loose parts in the coolant flow, and the process of coolant boiling. Each of these processes has been implicated in past nuclear plant failures. In the German Democratic Republic (GDR) P. Liewers and his colleagues have introduced noise analysis systems into the primary circuit of WWER-440 pressurized water reactors (PWR). The most progressive version (RAS-II) has become a prototype for research and routine investigations. This system is described. (author)

On the structural integrity evaluation about aged components

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

About one third of the nuclear power plants in Japan have been operated more than 30 years and flaws due to age-related degradation mechanisms have been detected in some components such as piping systems or core shrouds these years. Moreover, several severe earthquakes such as the Tohoku District - off the Pacific Ocean Earthquake or the Niigata-ken Chuetsu-oki Earthquake have struck some nuclear power plants in Japan recent years. Therefore, the structural integrity evaluation about nuclear installations and components considering seismic loads and aging mechanisms has become more and more important. In this study, several evaluation methods were proposed to assess the crack growth rate under the seismic loading conditions, to assess the failure conditions or the realistic failure capacities of the aged piping systems considering seismic or general loading conditions. Furthermore, analysis codes were developed considering aging mechanisms to carry out the integrity evaluation, or the failure probability evaluation which is useful in the seismic PSA evaluation. All of these assessment methods and analysis codes are being used and will be used more and more in the cross-check analyses or the safety reviews about nuclear installations and components. (author)

Some failure analyses of South African Air Force aircraft engine and airframe components

CSIR Research Space (South Africa)

Benson, JM

1998-06-01

Full Text Available Failure analyses of various engine and airframe components from South African Air Force aircraft have been performed by the Division of Materials Science and Technology over several years and these have ranged from crash investigations to minor...

Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, Units 1 and 2 nuclear power plants

International Nuclear Information System (INIS)

Lloyd, R.C.; Vehec, T.A.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Rossbach, L.W.; Sena, P.P. III

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Beaver Valley Units 1 and 2 were selected as two of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at Beaver Valley Units 1 and 2

The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

International Nuclear Information System (INIS)

Martz, H.F.

1995-05-01

It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example

Computational models for residual creep life prediction of power plant components

International Nuclear Information System (INIS)

Grewal, G.S.; Singh, A.K.; Ramamoortry, M.

2006-01-01

All high temperature - high pressure power plant components are prone to irreversible visco-plastic deformation by the phenomenon of creep. The steady state creep response as well as the total creep life of a material is related to the operational component temperature through, respectively, the exponential and inverse exponential relationships. Minor increases in the component temperature can thus have serious consequences as far as the creep life and dimensional stability of a plant component are concerned. In high temperature steam tubing in power plants, one mechanism by which a significant temperature rise can occur is by the growth of a thermally insulating oxide film on its steam side surface. In the present paper, an elegantly simple and computationally efficient technique is presented for predicting the residual creep life of steel components subjected to continual steam side oxide film growth. Similarly, fabrication of high temperature power plant components involves extensive use of welding as the fabrication process of choice. Naturally, issues related to the creep life of weldments have to be seriously addressed for safe and continual operation of the welded plant component. Unfortunately, a typical weldment in an engineering structure is a zone of complex microstructural gradation comprising of a number of distinct sub-zones with distinct meso-scale and micro-scale morphology of the phases and (even) chemistry and its creep life prediction presents considerable challenges. The present paper presents a stochastic algorithm, which can be' used for developing experimental creep-cavitation intensity versus residual life correlations for welded structures. Apart from estimates of the residual life in a mean field sense, the model can be used for predicting the reliability of the plant component in a rigorous probabilistic setting. (author)

Development of failure diagnosis method based on transient information of nuclear power plant

International Nuclear Information System (INIS)

Washio, Takashi; Kitamura, Masaharu; Sugiyama, Kazusuke

1987-01-01

This paper proposes a new method of failure diagnosis of nuclear power plant (NPP). Transient behavior of the NPP includes ample failure information even for a limited period of time from the failure onset. We tried to develop a diagnosis system with high capability of identifying the failure cause and of estimating failure severeness. The Walsh function transformation of transient time series data and the reduction of the Walsh coefficients into ternary valued amplitude indicators were utilized to extract the essential characteristics of failure. The correspondences of the transient characteristics and causes were summarized in a failure symptom database. A method of ternary tree search using an information measure as a heuristic strategy was adopted to conduct the efficient retrieval of failure causes in the database. Through numerical experiments using a simulation model of a NPP, the diagnostic capability of the system was proved to be satisfactory. (author)

Issues and approaches in risk-based aging analyses of passive components

International Nuclear Information System (INIS)

Uryasev, S.P.; Samanta, P.K.; Vesely, W.E.

1994-01-01

In previous NRC-sponsored work a general methodology was developed to quantify the risk contributions from aging components at nuclear plants. The methodology allowed Probabilistic Risk Analyses (PRAs) to be modified to incorporate the age-dependent component failure rates and also aging maintenance models to evaluate and prioritize the aging contributions from active components using the linear aging failure rate model and empirical components aging rates. In the present paper, this methodology is extended to passive components (for example, the pipes, heat exchangers, and the vessel). The analyses of passive components bring in issues different from active components. Here, we specifically focus on three aspects that need to be addressed in risk-based aging prioritization of passive components

Main factors for fatigue failure probability of pipes subjected to fluid thermal fluctuation

International Nuclear Information System (INIS)

Machida, Hideo; Suzuki, Masaaki; Kasahara, Naoto

2015-01-01

It is very important to grasp failure probability and failure mode appropriately to carry out risk reduction measures of nuclear power plants. To clarify the important factors for failure probability and failure mode of pipes subjected to fluid thermal fluctuation, failure probability analyses were performed by changing the values of a stress range, stress ratio, stress components and threshold of stress intensity factor range. The important factors for the failure probability are range, stress ratio (mean stress condition) and threshold of stress intensity factor range. The important factor for the failure mode is a circumferential angle range of fluid thermal fluctuation. When a large fluid thermal fluctuation acts on the entire circumferential surface of the pipe, the probability of pipe breakage increases, calling for measures to prevent such a failure and reduce the risk to the plant. When the circumferential angle subjected to fluid thermal fluctuation is small, the failure mode of piping is leakage and the corrective maintenance might be applicable from the viewpoint of risk to the plant. (author)

Probabilistic physics-of-failure models for component reliabilities using Monte Carlo simulation and Weibull analysis: a parametric study

International Nuclear Information System (INIS)

Hall, P.L.; Strutt, J.E.

2003-01-01

In reliability engineering, component failures are generally classified in one of three ways: (1) early life failures; (2) failures having random onset times; and (3) late life or 'wear out' failures. When the time-distribution of failures of a population of components is analysed in terms of a Weibull distribution, these failure types may be associated with shape parameters β having values 1 respectively. Early life failures are frequently attributed to poor design (e.g. poor materials selection) or problems associated with manufacturing or assembly processes. We describe a methodology for the implementation of physics-of-failure models of component lifetimes in the presence of parameter and model uncertainties. This treats uncertain parameters as random variables described by some appropriate statistical distribution, which may be sampled using Monte Carlo methods. The number of simulations required depends upon the desired accuracy of the predicted lifetime. Provided that the number of sampled variables is relatively small, an accuracy of 1-2% can be obtained using typically 1000 simulations. The resulting collection of times-to-failure are then sorted into ascending order and fitted to a Weibull distribution to obtain a shape factor β and a characteristic life-time η. Examples are given of the results obtained using three different models: (1) the Eyring-Peck (EP) model for corrosion of printed circuit boards; (2) a power-law corrosion growth (PCG) model which represents the progressive deterioration of oil and gas pipelines; and (3) a random shock-loading model of mechanical failure. It is shown that for any specific model the values of the Weibull shape parameters obtained may be strongly dependent on the degree of uncertainty of the underlying input parameters. Both the EP and PCG models can yield a wide range of values of β, from β>1, characteristic of wear-out behaviour, to β<1, characteristic of early-life failure, depending on the degree of

Real-time sensor failure detection by dynamic modelling of a PWR plant

International Nuclear Information System (INIS)

Turkcan, E.; Ciftcioglu, O.

1992-06-01

Signal validation and sensor failure detection is an important problem in real-time nuclear power plant (NPP) surveillance. Although conventional sensor redundancy, in a way, is a solution, identification of faulty sensor is necessary for further preventive actions to be taken. A comprehensive solution for the system so that any sensory reading is verified by its model based estimated counterpart, in real-time. Such a realization is accomplished by means of dynamic system's states estimation methodology using Kalman filter modelling technique. The method is investigated by means of real-time data of the steam generator of Borssele nuclear power plant and the method has proved to be satisfactory for real-time sensor failure detection as well as model validation verification. (author). 5 refs.; 6 figs.; 1 tab

Importance of biotic and abiotic components in feedback between plants and soil

OpenAIRE

Hanzelková, Věra

2017-01-01

The plant-soil feedback affects the forming of a plant community. Plants affect their own species as well as other species. The plant-soil feedback can be both positive and negative. Plants affect soil, change its properties, and the soil affects the plants reciprocally. Soil components can be divided into biotic and abiotic ones. The abiotic component is represented by physical and chemical properties of the soil. The main properties are the soil structure, the soil moisture, the soil temper...

Seismic fragility of nuclear power plant components (Phase II)

International Nuclear Information System (INIS)

Bandyopadhyay, K.K.; Hofmayer, C.H.; Kassir, M.K.; Pepper, S.E.

1990-02-01

As part of the Component Fragility Program which was initiated in FY 1985, three additional equipment classes have been evaluated. This report contains the fragility results and discussions on these equipment classes which are switchgear, I and C panels and relays. Both low and medium voltage switchgear assemblies have been considered and a separate fragility estimate for each type is provided. Test data on cabinets from the nuclear instrumentation/neutron monitoring system, plant/process protection system, solid state protective system and engineered safeguards test system comprise the BNL data base for I and C panels (NSSS). Fragility levels have been determined for various failure modes of switchgear and I ampersand C panels, and the deterministic results are presented in terms of test response spectra. In addition, the test data have been evaluated for estimating the respective probabilistic fragility levels which are expressed in terms of a median value, an uncertainty coefficient, a randomness coefficient and an HCLPF value. Due to a wide variation of relay design and the fragility level, a generic fragility level cannot be established for relays. 7 refs., 13 figs., 12 tabs

Uncertainties and quantification of common cause failure rates and probabilities for system analyses

International Nuclear Information System (INIS)

Vaurio, Jussi K.

2005-01-01

Simultaneous failures of multiple components due to common causes at random times are modelled by constant multiple-failure rates. A procedure is described for quantification of common cause failure (CCF) basic event probabilities for system models using plant-specific and multiple-plant failure-event data. Methodology is presented for estimating CCF-rates from event data contaminated with assessment uncertainties. Generalised impact vectors determine the moments for the rates of individual systems or plants. These moments determine the effective numbers of events and observation times to be input to a Bayesian formalism to obtain plant-specific posterior CCF-rates. The rates are used to determine plant-specific common cause event probabilities for the basic events of explicit fault tree models depending on test intervals, test schedules and repair policies. Three methods are presented to determine these probabilities such that the correct time-average system unavailability can be obtained with single fault tree quantification. Recommended numerical values are given and examples illustrate different aspects of the methodology

Statistical investigations of the failure behaviour of components in the AVR experimental nuclear power plant. Vol. 2

International Nuclear Information System (INIS)

Meyna, A.; Mock, R.; Tietze, A.; Hennings, W.

1989-08-01

From operational records of the years 1977 to 1986, service life distributions of helium valves in gas circuits of the AVR were determined. Results are constant failure rates in the range from 3 to 6x10 -6 /h and, for some populations, indications of time dependent failure rates. Nonparametric methods showed only limited efficiency. For a Bayesian approach the necessary prior information was missing. Furthermore, the main failure causes could be determined. (orig./HP) [de

Automotive component failures

CSIR Research Space (South Africa)

Heyes, AM

1998-06-01

Full Text Available in service for approximately 19\\999 km[ 1[1[ Visual examination Upon stripping the engine it was found that one of the combustion chambers showed heavy carbonaceous deposits indicative of the burning of oil "Fig[ 2# Circumferential black marks were found... whether failures in other vehicles could be expected[ 2[1[ Visual and stereo microscope examination The section of torsion bar submitted for examination was coated with a black paint coating which had ~aked o} at localised spots\\ where light rusting had...

Failure Behavior of Elbows with Local Wall Thinning

Science.gov (United States)

Lee, Sung-Ho; Lee, Jeong-Keun; Park, Jai-Hak

Wall thinning defect due to corrosion is one of major aging phenomena in carbon steel pipes in most plant industries, and it results in reducing load carrying capacity of the piping components. A failure test system was set up for real scale elbows containing various simulated wall thinning defects, and monotonic in-plane bending tests were performed under internal pressure to find out the failure behavior of them. The failure behavior of wall-thinned elbows was characterized by the circumferential angle of thinned region and the loading conditions to the piping system.

Auxiliary feedwater system risk-based inspection guide for the J.M. Farley Nuclear Power Plant

International Nuclear Information System (INIS)

Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G.

1990-10-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment(PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab

Probability problems in seismic risk analysis and load combinations for nuclear power plants

International Nuclear Information System (INIS)

George, L.L.

1983-01-01

This workshop describes some probability problems in power plant reliability and maintenance analysis. The problems are seismic risk analysis, loss of load probability, load combinations, and load sharing. The seismic risk problem is to compute power plant reliability given an earthquake and the resulting risk. Component survival occurs if its peak random response to the earthquake does not exceed its strength. Power plant survival is a complicated Boolean function of component failures and survivals. The responses and strengths of components are dependent random processes, and the peak responses are maxima of random processes. The resulting risk is the expected cost of power plant failure

Analyses of component degradation to evaluate maintenance effectiveness and aging effects

International Nuclear Information System (INIS)

Samanta, P.K.; Hsu, F.; Subudhi, M.; Vesely, W.E.

1991-01-01

This paper describes degradation modeling, an approach for analyzing degradation and failure of components to understand the aging process of components. As used in our study, degradation modeling is the analysis of information on degradation of components for developing models of the degradation process and its implications. This modeling focuses on the analysis of the times of degradations of components, to model how the rate of degradation changes with the age of the component. With this methodology we also determine the effectiveness of maintenance as applicable to aging evaluations. The specific applications which are performed show quantitative models of degradation rates of components and failure rates of components from plant-specific data. The statistical techniques allow aging trends to be identified in the degradation data and in the failure data. Initial estimates of the effectiveness of maintenance in limiting degradations from becoming failures are developed. These results are important first steps in degradation modeling, and show that degradation can be modeled to identify aging trends. 2 refs., 8 figs., 1 tab

Prediction of the time-dependent failure rate for normally operating components taking into account the operational history

International Nuclear Information System (INIS)

Vrbanic, I.; Simic, Z.; Sljivac, D.

2008-01-01

The prediction of the time-dependent failure rate has been studied, taking into account the operational history of a component used in applications such as system modeling in a probabilistic safety analysis in order to evaluate the impact of equipment aging and maintenance strategies on the risk measures considered. We have selected a time-dependent model for the failure rate which is based on the Weibull distribution and the principles of proportional age reduction by equipment overhauls. Estimation of the parameters that determine the failure rate is considered, including the definition of the operational history model and likelihood function for the Bayesian analysis of parameters for normally operating repairable components. The operational history is provided as a time axis with defined times of overhauls and failures. An example for demonstration is described with prediction of the future behavior for seven different operational histories. (orig.)

Detection and mitigation of aging and service wear effects of nuclear power plant components in Canada

International Nuclear Information System (INIS)

Pachner, J.

1987-07-01

In Canada, the operational safety management of nuclear power plants employs methods which are intended to prevent, detect, correct and mitigate system and component failures from any cause, including the effects of aging and service wear degradation. The paper gives an overview of the application of these methods in the detection and mitigation of aging effects before they impact on plant safety and production reliability. Regulatory audits of these methods, to ensure that an acceptable level of plant safety is maintained by the nuclear power plant licensees, are also described. The methods are: a preventive maintenance program, Significant Event Reporting system, and a reliability based assessment of performance of safety related systems. The above methods are discussed and illustrated by examples. The soundness of the approach has been proven by the results achieved in 163 reactor-years of operation. Present and future developments include reviews of current monitoring, testing and inspection methods to ensure that appropriate time variant parameters (capable of revealing aging degradation before loss of functional capability) are monitored, and reviews of the effectiveness of existing maintenance programs and methods in mitigating aging and service wear effects

IAEA's experience in compiling a generic component reliability data base

International Nuclear Information System (INIS)

Tomic, B.; Lederman, L.

1991-01-01

Reliability data are essential in probabilistic safety assessment, with component reliability parameters being particularly important. Component failure data which is plant specific would be most appropriate but this is rather limited. However, similar components are used in different designs. Generic data, that is all data that is not plant specific to the plant being analyzed but which relates to components more generally, is important. The International Atomic Energy Agency has compiled the Generic Component Reliability Data Base from data available in the open literature. It is part of the IAEA computer code package for fault/event tree analysis. The Data Base contains 1010 different records including most of the components used in probabilistic safety analyses of nuclear power plants. The data base input was quality controlled and data sources noted. The data compilation procedure and problems associated with using generic data are explained. (UK)

Age-Related Degradation of Nuclear Power Plant Structures and Components

International Nuclear Information System (INIS)

Braverman, J.; Chang, T.-Y.; Chokshi, N.; Hofmayer, C.; Morante, R.; Shteyngart, S.

1999-01-01

This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what was the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk

Ageing effects modelling in probabilistic safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Nitoi, M.; Turcu, I.; Florescu, G.; Apostol, M.; Farcasiu, M.; Pavelescu, M.

2005-01-01

Ageing management has become a major concern for many responsible organizations during the last years, because as the operating power plants have got older, they may have the tendency to become less safe. The effects of age-related degradation of plant components, systems and structures are necessary to be assessed in order to assure a continuous safe operation of nuclear power plants. The Probabilistic Safety Analysis (PSA) is an efficient system analysis method which is used to assess the risk of operation of nuclear power plants. In the assessment of risk level for a plant, most of the PSA studies generally didn't take into account the ageing effects, and uses a time averaged unavailability. By incorporation of ageing effects, the results enable an identification of the components that have the greatest effect on risk if their failure rates increase due to ageing effects modelling. In this paper, it was assessed the impact on Class IV Electrical Power System unavailability of the assumed increase in components failure probability caused by components ageing. The electrical system was chosen for the study because there are a lot of cables and for these types of equipment there is no planned preventive or corrective maintenance, and they are originally designed to reach the end of plant life with an adequate safety margin. To quantify the effects of age-related degradation on components, the linear ageing model was used. In this model, the failure rate of a component λ (t) is expressed as a sum of two independent failure rates, one associated with random failure, λ 0 , and the other associated with failures due to aging α, so: λ(t) = λ 0 + αt. The basic events were coded using a computer code similar to CAFTA, developed in INR Pitesti. For the reliability data allocation for basic events a intern data base was used. This data base contains data from the following generic data bases: IAEA Component Reliability Data for use in PSA, Point Lepreau Component

An analysis of the annual probability of failure of the waste hoist brake system at the Waste Isolation Pilot Plant (WIPP)

Energy Technology Data Exchange (ETDEWEB)

Greenfield, M.A. [Univ. of California, Los Angeles, CA (United States); Sargent, T.J.

1995-11-01

The Environmental Evaluation Group (EEG) previously analyzed the probability of a catastrophic accident in the waste hoist of the Waste Isolation Pilot Plant (WIPP) and published the results in Greenfield (1990; EEG-44) and Greenfield and Sargent (1993; EEG-53). The most significant safety element in the waste hoist is the hydraulic brake system, whose possible failure was identified in these studies as the most important contributor in accident scenarios. Westinghouse Electric Corporation, Waste Isolation Division has calculated the probability of an accident involving the brake system based on studies utilizing extensive fault tree analyses. This analysis conducted for the U.S. Department of Energy (DOE) used point estimates to describe the probability of failure and includes failure rates for the various components comprising the brake system. An additional controlling factor in the DOE calculations is the mode of operation of the brake system. This factor enters for the following reason. The basic failure rate per annum of any individual element is called the Event Probability (EP), and is expressed as the probability of failure per annum. The EP in turn is the product of two factors. One is the {open_quotes}reported{close_quotes} failure rate, usually expressed as the probability of failure per hour and the other is the expected number of hours that the element is in use, called the {open_quotes}mission time{close_quotes}. In many instances the {open_quotes}mission time{close_quotes} will be the number of operating hours of the brake system per annum. However since the operation of the waste hoist system includes regular {open_quotes}reoperational check{close_quotes} tests, the {open_quotes}mission time{close_quotes} for standby components is reduced in accordance with the specifics of the operational time table.

An analysis of the annual probability of failure of the waste hoist brake system at the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

Greenfield, M.A.; Sargent, T.J.

1995-11-01

The Environmental Evaluation Group (EEG) previously analyzed the probability of a catastrophic accident in the waste hoist of the Waste Isolation Pilot Plant (WIPP) and published the results in Greenfield (1990; EEG-44) and Greenfield and Sargent (1993; EEG-53). The most significant safety element in the waste hoist is the hydraulic brake system, whose possible failure was identified in these studies as the most important contributor in accident scenarios. Westinghouse Electric Corporation, Waste Isolation Division has calculated the probability of an accident involving the brake system based on studies utilizing extensive fault tree analyses. This analysis conducted for the U.S. Department of Energy (DOE) used point estimates to describe the probability of failure and includes failure rates for the various components comprising the brake system. An additional controlling factor in the DOE calculations is the mode of operation of the brake system. This factor enters for the following reason. The basic failure rate per annum of any individual element is called the Event Probability (EP), and is expressed as the probability of failure per annum. The EP in turn is the product of two factors. One is the open-quotes reportedclose quotes failure rate, usually expressed as the probability of failure per hour and the other is the expected number of hours that the element is in use, called the open-quotes mission timeclose quotes. In many instances the open-quotes mission timeclose quotes will be the number of operating hours of the brake system per annum. However since the operation of the waste hoist system includes regular open-quotes reoperational checkclose quotes tests, the open-quotes mission timeclose quotes for standby components is reduced in accordance with the specifics of the operational time table

Application of risk-based methods for inspection of nuclear power plant components

International Nuclear Information System (INIS)

Balkey, K.R.

1992-01-01

In-service inspections (ISIs) can play a significant role in minimizing equipment and structural failures. All aspects of inspections, i.e., objectives, method, timing, and the acceptance criteria for detected flaws can affect the probability of component failure. Where ISI programs exist, they are primarily based on prior experience and engineering judgment. At best, some include an implicit consideration of risk (probability of failure multiplied by consequence). Since late 1988, a multidisciplined American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection Guidelines has been addressing the general question of how to formally incorporate risk considerations into plans and requirements for the ISI of components and structural systems. The task force and steering committee that guided the project have concluded that appropriate analytical methods exist for evaluating and quantifying risks associated with pressure boundary and structural failures. With the support of about a dozen industry and government organizations, the research group has recommended a general methodology for establishing a risk-based inspection program that could be applied to any nuclear system or structural system

Development of a SPV management program for nuclear power plants

International Nuclear Information System (INIS)

Lee, Eun-Chan; Na, Jang-Hwan; Lee, Doo-Young; Oh, Seong-Jong; Jerng, Dong-Wook

2009-01-01

The Single Point Vulnerability (SPV) is a characteristic of a component whose failure results in plant transients. KHNP (Korea Hydro and Nuclear Power Co.) has evaluated major systems with critical components and developed a SPV management program to reduce the trip frequency and to raise plant's availability. This study includes a comprehensive methodology for SPV evaluation and its results. This methodology consists of the qualitative evaluation focused on the SPV component list preparation and the quantitative evaluation through FMEA (Failure Mode Effect Analyses) and FTA (Fault Tree Analyses) of critical systems. The qualitative evaluation reduced differences of the SPV lists between the identically designed plants and established strategies for the improvement of the SPV component reliability. The quantitative evaluation identified additional SPV components and developed the fault tree model for a Trip Monitor which showed logic relationships of channel components in the trip-related systems. (author)

Plant systems/components modularization study. Final report. [PWR

Energy Technology Data Exchange (ETDEWEB)

1977-07-01

The final results are summarized of a Plant Systems/Components Modularization Study based on Stone and Webster's Pressurized Water Reactor Reference Design. The program has been modified to include evaluation of the most promising areas for modular consideration based on the level of the Sundesert Project engineering design completion and the feasibility of their incorporation into the plant construction effort.

Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

Energy Technology Data Exchange (ETDEWEB)

Nyman, R; Erixon, S; Tomic, B; Lydell, B

1995-12-01

SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs.

Reliability of piping system components. Volume 1: Piping reliability - A resource document for PSA applications

International Nuclear Information System (INIS)

Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

1995-12-01

SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs

Heavy steel casting components for power plants 'mega-components' made of high Cr-steels

Energy Technology Data Exchange (ETDEWEB)

Hanus, Reinhold [voestalpine Giesserei Linz GmbH, Linz (Austria)

2010-07-01

Steel castings of creep resistant steels play a key role in fossil fuel fired power plants for highly loaded components in the high and intermediate pressure section of the turbines. Inner and outer casings, valve casings, inlet connections and elbows are examples of such critical components. The most important characteristic in a power plant is the efficiency, which mainly drives the CO2-emission. As a consequence of steadily improving power plant efficiencies and ever stricter emission standards, steam parameters become more critical and the creep resistance of the cast materials must also be constantly improved. The foundries voestalpine Giesserei Linz and voestalpine Giesserei Traisen participated in the development of the new 9-10% Cr-steels for application up to 625 C/650 C and in the THERMIE project where Ni-base alloys for 700 C-power plants were developed. Beside the material development in the European research projects the commercial production had to be established for industrial processes and the newly developed materials have to be transferred from research into the commercial production of heavy cast components. After selecting the most promising alloy from the laboratory melts, welding tests were performed - mostly with matching electrodes also produced within COST/THERMIE. Base material and welds were investigated in respect of microstructure, creep resistance, mechanical properties and weldability. Heat treatment investigations were also necessary for optimization of the mechanical properties. Based on the results of these studies, pilot components and plates for testing welding processes were cast in order to verify the castability and weldability of larger parts and to make any necessary adjustments to chemical composition, heat treatment or welding parameters. Parallel to the ongoing creep tests within COST/THERMIE-program, the newly developed steel grades were introduced into the commercial production of large components. This involved finding

Operating Experience of Digital, Software-based Components Used in I and C and Electrical Systems in German NPPs

International Nuclear Information System (INIS)

Blum, Stefanie; Lochthofen, Andre; Quester, Claudia; Arians, Robert

2015-01-01

In recent years, many components in instrumentation and control (I and C) and electrical systems of nuclear power plants (NPPs) were replaced by digital, software-based components. Due to the more complex structure, software-based I and C and electrical components show the potential for new failure mechanisms and an increasing number of failure possibilities, including the potential for common cause failures. An evaluation of the operating experience of digital, software-based components may help to determine new failure modes of these components. In this paper, we give an overview over the results of the evaluation of the operating experience of digital, software-based components used in I and C and electrical systems in NPPs in Germany. (authors)

How insects overcome two-component plant chemical defence

DEFF Research Database (Denmark)

Pentzold, Stefan; Zagrobelny, Mika; Rook, Frederik

2014-01-01

Insect herbivory is often restricted by glucosylated plant chemical defence compounds that are activated by plant β-glucosidases to release toxic aglucones upon plant tissue damage. Such two-component plant defences are widespread in the plant kingdom and examples of these classes of compounds...... are alkaloid, benzoxazinoid, cyanogenic and iridoid glucosides as well as glucosinolates and salicinoids. Conversely, many insects have evolved a diversity of counteradaptations to overcome this type of constitutive chemical defence. Here we discuss that such counter-adaptations occur at different time points......, before and during feeding as well as during digestion, and at several levels such as the insects’ feeding behaviour, physiology and metabolism. Insect adaptations frequently circumvent or counteract the activity of the plant β-glucosidases, bioactivating enzymes that are a key element in the plant’s two...

Coupling failure between stem and femoral component in a constrained revision total knee arthroplasty.

LENUS (Irish Health Repository)

Butt, Ahsan Javed

2013-02-01

Knee revision using constrained implants is associated with greater stresses on the implant and interface surfaces. The present report describes a case of failure of the screw coupling between the stem and the femoral component. The cause of the failure is surmised with outline of the treatment in this case with extensive femoral bone loss. Revision implant stability was augmented with the use of a cemented femoral stem, screw fixation and the metaphyseal sleeve of an S-ROM modular hip system (DePuy international Ltd).

The effects of age on nuclear power plant containment cooling systems

Energy Technology Data Exchange (ETDEWEB)

Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A. [Brookhaven National Lab., Upton, NY (United States); Davis, J. [Science Applications International Corp., New York, NY (United States)

1994-04-01

A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated.

The effects of age on nuclear power plant containment cooling systems

International Nuclear Information System (INIS)

Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A.; Davis, J.

1994-04-01

A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated

A review of typical thermal fatigue failure models for solder joints of electronic components

Science.gov (United States)

Li, Xiaoyan; Sun, Ruifeng; Wang, Yongdong

2017-09-01

For electronic components, cyclic plastic strain makes it easier to accumulate fatigue damage than elastic strain. When the solder joints undertake thermal expansion or cold contraction, different thermal strain of the electronic component and its corresponding substrate is caused by the different coefficient of thermal expansion of the electronic component and its corresponding substrate, leading to the phenomenon of stress concentration. So repeatedly, cracks began to sprout and gradually extend [1]. In this paper, the typical thermal fatigue failure models of solder joints of electronic components are classified and the methods of obtaining the parameters in the model are summarized based on domestic and foreign literature research.

Proportional and scale change models to project failures of mechanical components with applications to space station

Science.gov (United States)

Taneja, Vidya S.

1996-01-01

In this paper we develop the mathematical theory of proportional and scale change models to perform reliability analysis. The results obtained will be applied for the Reaction Control System (RCS) thruster valves on an orbiter. With the advent of extended EVA's associated with PROX OPS (ISSA & MIR), and docking, the loss of a thruster valve now takes on an expanded safety significance. Previous studies assume a homogeneous population of components with each component having the same failure rate. However, as various components experience different stresses and are exposed to different environments, their failure rates change with time. In this paper we model the reliability of a thruster valves by treating these valves as a censored repairable system. The model for each valve will take the form of a nonhomogeneous process with the intensity function that is either treated as a proportional hazard model, or a scale change random effects hazard model. Each component has an associated z, an independent realization of the random variable Z from a distribution G(z). This unobserved quantity z can be used to describe heterogeneity systematically. For various models methods for estimating the model parameters using censored data will be developed. Available field data (from previously flown flights) is from non-renewable systems. The estimated failure rate using such data will need to be modified for renewable systems such as thruster valve.

Concept of a new method for fatigue monitoring of nuclear power plant components

International Nuclear Information System (INIS)

Zafosnik, M.; Cizelj, L.

2007-01-01

Fatigue is one of the well-understood aging mechanisms affecting mechanical components in many industrial facilities including nuclear power plants. Operational experience of nuclear power plants worldwide to date confirmed adequate design of safety related components against fatigue. In some cases however, for example when the plant life extension is envisioned, it may be very useful to monitor the remaining fatigue life of safety related components. Nuclear power plants components are classified into safety classes regarding their importance in mitigating the consequences of hypothetic accidents. Service life of components subjected to fatigue loading can be estimated with Usage Factor uk. A concept of the new method aiming both at monitoring the current state of the component and predicting its remaining lifetime in the life-extension conditions is presented. The method is based on determination of partial Usage Factor of components in which operating transients will be considered and compared to design transients. (author)

Enhanced Component Performance Study: Emergency Diesel Generators 1998-2014

International Nuclear Information System (INIS)

Schroeder, John Alton

2015-01-01

This report presents an enhanced performance evaluation of emergency diesel generators (EDGs) at U.S. commercial nuclear power plants. This report evaluates component performance over time using (1) Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES) data from 1998 through 2014 and (2) maintenance unavailability (UA) performance data from Mitigating Systems Performance Index (MSPI) Basis Document data from 2002 through 2014. The objective is to show estimates of current failure probabilities and rates related to EDGs, trend these data on an annual basis, determine if the current data are consistent with the probability distributions currently recommended for use in NRC probabilistic risk assessments, show how the reliability data differ for different EDG manufacturers and for EDGs with different ratings; and summarize the subcomponents, causes, detection methods, and recovery associated with each EDG failure mode. Engineering analyses were performed with respect to time period and failure mode without regard to the actual number of EDGs at each plant. The factors analyzed are: sub-component, failure cause, detection method, recovery, manufacturer, and EDG rating. Six trends with varying degrees of statistical significance were identified in the data.

Comparison of Failure Analysis and Operating Experiences of Digital Control Systems

Energy Technology Data Exchange (ETDEWEB)

Lee, Eun Chan; Shin, Tae Young [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2014-08-15

This study focuses on digital control systems that have the same functions but different designs. Some differences and common points between these two digital control systems are analyzed in terms of vulnerabilities in plant operation. In addition, this study confirms why unexpected outcomes can occur through a comparison of the system failure experiences with the analytic results of FMEA and FTA. This evaluation demonstrates that the digital system may have vulnerable components whose single failures can cause plant transients even if the system has a redundant structure according to its system design.

How simulation of failure risk can improve structural reliability - application to pressurized components and pipes

OpenAIRE

Cioclov, Dimitru Dragos

2013-01-01

Probabilistic methods for failure risk assessment are introduced, with reference to load carrying structures, such as pressure vessels (PV) and components of pipes systems. The definition of the failure risk associated with structural integrity is made in the context of the general approach to structural reliability. Sources of risk are summarily outlined with emphasis on variability and uncertainties (V&U) which might be encountered in the analysis. To highlight the problem, in its practical...

Nuclear power plant component protection

International Nuclear Information System (INIS)

Michel, E.; Ruf, R.; Dorner, H.

1976-01-01

Described is a nuclear power plant installation which includes a concrete biological shield forming a pit in which a reactor pressure vessel is positioned. A steam generator on the outside of the shield is connected with the pressure vessel via coolant pipe lines which extend through the shield, the coolant circulation being provided by a coolant pump which is also on the outside of the shield. To protect these components on the outside of the shield and which are of mainly or substantially cylindrical shape, semicylindrical concrete segments are interfitted around them to form complete outer cylinders which are retained against outward separation radially from the components, by rings of high tensile steel which may be interspaced so closely that they provide, in effect, an outer steel cylinder. The invention is particularly applicable to pressurized-water coolant reactor installations

Multiple-failure signal validation in nuclear power plants using artificial neural networks

International Nuclear Information System (INIS)

Fantoni, P.F.; Mazzola, A.

1996-01-01

The possibility of using a neural network to validate process signals during normal and abnormal plant conditions is explored. In boiling water reactor plants, signal validation is used to generate reliable thermal limits calculation and to supply reliable inputs to other computerized systems that support the operator during accident scenarios. The way that autoassociative neural networks can promptly detect faulty process signal measurements and produce a best estimate of the actual process values even in multifailure situations is shown. A method was developed to train the network for multiple sensor-failure detection, based on a random failure simulation algorithm. Noise was artificially added to the input to evaluate the network's ability to respond in a very low signal-to-noise ratio environment. Training and test data sets were simulated by the real-time transient simulator code APROS

Research on fault characteristics about switching component failures for distribution electronic power transformers

Science.gov (United States)

Sang, Z. X.; Huang, J. Q.; Yan, J.; Du, Z.; Xu, Q. S.; Lei, H.; Zhou, S. X.; Wang, S. C.

2017-11-01

The protection is an essential part for power device, especially for those in power grid, as the failure may cost great losses to the society. A study on the voltage and current abnormality in the power electronic devices in Distribution Electronic Power Transformer (D-EPT) during the failures on switching components is presented, as well as the operational principles for 10 kV rectifier, 10 kV/400 V DC-DC converter and 400 V inverter in D-EPT. Derived from the discussion on the effects of voltage and current distortion, the fault characteristics as well as a fault diagnosis method for D-EPT are introduced.

A quantitative impact analysis of sensor failures on human operator's decision making in nuclear power plants

International Nuclear Information System (INIS)

Seong, Poong Hyun

2004-01-01

In emergency or accident situations in nuclear power plants, human operators take important roles in generating appropriate control signals to mitigate accident situation. In human reliability analysis (HRA) in the framework of probabilistic safety assessment (PSA), the failure probabilities of such appropriate actions are estimated and used for the safety analysis of nuclear power plants. Even though understanding the status of the plant is basically the process of information seeking and processing by human operators, it seems that conventional HRA methods such as THERP, HCR, and ASEP does not pay a lot of attention to the possibilities of providing wrong information to human operators. In this paper, a quantitative impact analysis of providing wrong information to human operators due to instrument faults or sensor failures is performed. The quantitative impact analysis is performed based on a quantitative situation assessment model. By comparing the situation in which there are sensor failures and the situation in which there are not sensor failures, the impact of sensor failures can be evaluated quantitatively. It is concluded that the impact of sensor failures are quite significant at the initial stages, but the impact is gradually reduced as human operators make more and more observations. Even though the impact analysis is highly dependent on the situation assessment model, it is expected that the conclusions made based on other situation assessment models with be consistent with the conclusion made in this paper. (author)

Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

International Nuclear Information System (INIS)

Lydell, Bengt; Olsson, Anders

2008-01-01

This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning meetings that were

Nuclear plant reliability data system. 1979 annual reports of cumulative system and component reliability

International Nuclear Information System (INIS)

1979-01-01

The primary purposes of the information in these reports are the following: to provide operating statistics of safety-related systems within a unit which may be used to compare and evaluate reliability performance and to provide failure mode and failure rate statistics on components which may be used in failure mode effects analysis, fault hazard analysis, probabilistic reliability analysis, and so forth

A 'cost-effective' probabilistic model to select the dominant factors affecting the variation of the component failure rate

International Nuclear Information System (INIS)

Kirchsteiger, C.

1992-11-01

Within the framework of a Probabilistic Safety Assessment (PSA), the component failure rate λ is a key parameter in the sense that the study of its behavior gives the essential information for estimating the current values as well as the trends in the failure probabilities of interest. Since there is an infinite variety of possible underlying factors which might cause changes in λ (e.g. operating time, maintenance practices, component environment, etc.), an 'importance ranking' process of these factors is considered most desirable to prioritize research efforts. To be 'cost-effective', the modeling effort must be small, i.e. essentially involving no estimation of additional parameters other than λ. In this paper, using a multivariate data analysis technique and various statistical measures, such a 'cost-effective' screening process has been developed. Dominant factors affecting the failure rate of any components of interest can easily be identified and the appropriateness of current research plans (e.g. on the necessity of performing aging studies) can be validated. (author)

Methods for dependency estimation and system unavailability evaluation based on failure data statistics

International Nuclear Information System (INIS)

Azarm, M.A.; Hsu, F.; Martinez-Guridi, G.; Vesely, W.E.

1993-07-01

This report introduces a new perspective on the basic concept of dependent failures where the definition of dependency is based on clustering in failure times of similar components. This perspective has two significant implications: first, it relaxes the conventional assumption that dependent failures must be simultaneous and result from a severe shock; second, it allows the analyst to use all the failures in a time continuum to estimate the potential for multiple failures in a window of time (e.g., a test interval), therefore arriving at a more accurate value for system unavailability. In addition, the models developed here provide a method for plant-specific analysis of dependency, reflecting the plant-specific maintenance practices that reduce or increase the contribution of dependent failures to system unavailability. The proposed methodology can be used for screening analysis of failure data to estimate the fraction of dependent failures among the failures. In addition, the proposed method can evaluate the impact of the observed dependency on system unavailability and plant risk. The formulations derived in this report have undergone various levels of validations through computer simulation studies and pilot applications. The pilot applications of these methodologies showed that the contribution of dependent failures of diesel generators in one plant was negligible, while in another plant was quite significant. It also showed that in the plant with significant contribution of dependency to Emergency Power System (EPS) unavailability, the contribution changed with time. Similar findings were reported for the Containment Fan Cooler breakers. Drawing such conclusions about system performance would not have been possible with any other reported dependency methodologies

Development of a generic data base for failure rate

International Nuclear Information System (INIS)

Mosleh, A.; Apostolakis, G.

1985-01-01

The data analysis task in a probabilistic risk assessment (PRA) involves the assessment of data needs, the collection of information, and, finally, the analysis of the data to generate estimates for various parameters. This paper describes a framework for developing a data base for component failure rates and presents mathematical methods for the analysis of various types of information. The discussion is focused on the development of generic data bases used in PRAs. For plants without an operating history, the generic distributions are used directly to calculate component unavailability. In the case of plants that have operated for some time, the generic distributions can be used as priors in Bayesian analysis and, thus, specialized by plant-specific experience

Formation of higher plant component microbial community in closed ecological system

Science.gov (United States)

Tirranen, L. S.

2001-07-01

Closed ecological systems (CES) place at the disposal of a researcher unique possibilities to study the role of microbial communities in individual components and of the entire system. The microbial community of the higher plant component has been found to form depending on specific conditions of the closed ecosystem: length of time the solution is reused, introduction of intrasystem waste water into the nutrient medium, effect of other component of the system, and system closure in terms of gas exchange. The higher plant component formed its own microbial complex different from that formed prior to closure. The microbial complex of vegetable polyculture is more diverse and stable than the monoculture of wheat. The composition of the components' microflora changed, species diversity decreased, individual species of bacteria and fungi whose numbers were not so great before the closure prevailed. Special attention should be paid to phytopathogenic and conditionally pathogenic species of microorganisms potentially hazardous to man or plants and the least controlled in CES. This situation can endanger creation of CES and make conjectural existence of preplanned components, man, specifically, and consequently, of CES as it is.

Steam explosions-induced containment failure studies for Swiss nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Zuchuat, O.; Schmocker, U. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland); Esmaili, H.; Khatib-Rahbar, M.

1998-01-01

The assessment of the consequences of both in-vessel and ex-vessel energetic fuel-coolant interaction for Beznau (a Westinghouse pressurized water reactor with a large, dry containment), Goesgen (a Siemens/KWU pressurized water reactor with a large, dry containment) and Leibstadt (a General Electric boiling water reactor-6 with a free standing steel, MARK-III containment) nuclear power plants is presented in this paper. The Conditional Containment Failure Probability of the steel containment of these Swiss nuclear power plants is determined based on different probabilistic approaches. (author)

Unavailability of repairable components with failures detectable upon demand: Remarks on a work of Caldarola

International Nuclear Information System (INIS)

Souza Borges, W. de; Silva Pagetti, P. da

1987-01-01

In this paper an exact expression has been obtained for the asymptotic mean unavailability in time domain of components with failures detected upon demand. The model is more general than those proposed in the literature since it allows the use of general distributions for component life times, repair times and inter-demand times. Expressions for the special case of exponential life times have also been derived. (orig.)

System diagnostics using qualitative analysis and component functional classification

International Nuclear Information System (INIS)

Reifman, J.; Wei, T.Y.C.

1993-01-01

A method for detecting and identifying faulty component candidates during off-normal operations of nuclear power plants involves the qualitative analysis of macroscopic imbalances in the conservation equations of mass, energy and momentum in thermal-hydraulic control volumes associated with one or more plant components and the functional classification of components. The qualitative analysis of mass and energy is performed through the associated equations of state, while imbalances in momentum are obtained by tracking mass flow rates which are incorporated into a first knowledge base. The plant components are functionally classified, according to their type, as sources or sinks of mass, energy and momentum, depending upon which of the three balance equations is most strongly affected by a faulty component which is incorporated into a second knowledge base. Information describing the connections among the components of the system forms a third knowledge base. The method is particularly adapted for use in a diagnostic expert system to detect and identify faulty component candidates in the presence of component failures and is not limited to use in a nuclear power plant, but may be used with virtually any type of thermal-hydraulic operating system. 5 figures

Prioritization of reactor control components susceptible to fire damage as a consequence of aging

International Nuclear Information System (INIS)

Lowry, W.; Vigil, R.; Nowlen, S.

1994-01-01

The Fire Vulnerability of Aged Electrical Components Test Program is to identify and assess issues of plant aging that could lead to an increase in nuclear power plant risk because of fires. Historical component data and prior analyses are used to prioritize a list of components with respect to aging and fire vulnerability and the consequences of their failure on plant safety systems. The component list emphasizes safety system control components, but excludes cables, large equipment, and devices encompassed in the Equipment Qualification (EQ) program. The test program selected components identified in a utility survey and developed test and fire conditions necessary to maximize the effectiveness of the test program. Fire damage considerations were limited to purely thermal effects

Pilot program to identify valve failures which impact the safety and operation of light water nuclear power plants

International Nuclear Information System (INIS)

Tsacoyeanes, J.C.; Raju, P.P.

1980-04-01

The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents

Statistical evaluation of failures and repairs of the V-1 measuring and control system

International Nuclear Information System (INIS)

Laurinec, R.; Korec, J.; Mitosinka, J.; Zarnovican, V.

1984-01-01

A failure record card system was introduced for evaluating the reliability of the measurement and control equipment of the V-1 nuclear power plant. The SPU-800 microcomputer system is used for recording data on magnetic tape and their transmission to the central data processing department. The data are used for evaluating the reliability of components and circuits and a selection is made of the most failure-prone components, and the causes of failures are evaluated as are failure identification, repair and causes of outages. The system provides monthly, annual and total assessment data since the system was commissioned. The results of the statistical evaluation of failures are used for planning preventive maintenance and for determining optimal repair intervals. (E.S.)

Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

International Nuclear Information System (INIS)

Baron, J.; Caballero, C.; Zarate, S.M.

1997-01-01

The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

Plant components and authenticity of landscape architecture monuments

Directory of Open Access Journals (Sweden)

Miloš Pejchal

2011-01-01

Full Text Available Plants specifications emphasize the fundamental meaning of the “fourth space dimension” – time by their usage: (a the space cannot be composed as a static image; (b some used plants are not the planned part of the target state; (c delayed onset of full functionality; (d substantial importance of care for achieving and maintaining of the full functionality; (e cultivation measures must be implemented in a certain time period, i.e. the “time window”; (f replacement of already obsolete generation of full-grown and long-aged trees with a new generation is often carried out in the amended site conditions and different social situation. Historical authenticity of the plant components has the following specifics: (a its basic assumption may not be the original specimens of plants, it is the preservation of the principle contained in this original substance; (b the period during which the plant is able to represent the principle of the original substance is often shorter than the length of its existence; (c gradual recovery of surviving individuals is often difficult to impossible in plants groups and stands; (d it is often impossible to meet the recommendations of Venice Charter to not to apply the hypothesis and differentiation of added parts from the original ones. There was not paid enough attention to following aspects of the authenticity of plant components: (a the importance of particular developmental stages of the element; (b the role of age structure (the same age – different age for different types of elements; (c the effect of different length of the existence of space-formative elements (different periods of their recovery to the overall composition effect; (d role of historical technologies.

In-plant reliability data base for nuclear power plant components: data collection and methodology report

International Nuclear Information System (INIS)

Drago, J.P.; Borkowski, R.J.; Pike, D.H.; Goldberg, F.F.

1982-07-01

The development of a component reliability data for use in nuclear power plant probabilistic risk assessments and reliabiilty studies is presented in this report. The sources of the data are the in-plant maintenance work request records from a sample of nuclear power plants. This data base is called the In-Plant Reliability Data (IPRD) system. Features of the IPRD system are compared with other data sources such as the Licensee Event Report system, the Nuclear Plant Reliability Data system, and IEEE Standard 500. Generic descriptions of nuclear power plant systems formulated for IPRD are given

Component failure-rate data with potential applicability to the hot experimental facility. Technical information

International Nuclear Information System (INIS)

Dexter, A.H.

1980-12-01

A literature search, that was aided by computer searches of a number of data bases, resulted in the compilation of approximately 1223 pieces of component failure-rate data under 136 subject categories. The data bank can be provided upon request as a punched-card deck or on magnetic tape

Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

International Nuclear Information System (INIS)

2000-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

Overheating failure of superheater suspension tubes of a captive thermal power plant boiler

International Nuclear Information System (INIS)

Bhattacharya, Sova; Amir, Q.M.; Kannan, C.; Mahapatra, S.B.

2000-01-01

Failure of boiler tubes is the foremost cause of forced boiler outages. One of the predominant failure mechanism of boiler tubes is the stress rupture failure in the form of either short term overheating or long term overheating which are normally encountered in superheater and reheater sections working in the creep range. The strength of the boiler tube depends on the stress level as well on the temperature of exposure in the creep range. An increase in either can reduce the time to rupture. Time at the exposure temperature is an important factor based on which the failures are categorised as either short term or long term. Though there is no established time duration criteria demarcating the short or long term stress rupture failures, depending on the various manifestations on the failed samples, one can categorise the failure. This paper addresses one such stress rupture failure in the superheater section of a captive thermal power plant of a refinery. Multiple failures on the suspension coil of a superheater section was investigated for the cause of failure. Laboratory investigation of the failed sample involved visual inspection, dimensional measurements, chemical analysis of internal deposits and microstructural study. On the basis of these, the failure was attributed to deposition of trisodium phosphate carried over by the feed water into the superheater section resulting in chokage and increase in local operating hoop stresses of the tube. The ultimate failure was thus categorised as long term overheating failure. (author)

Enhanced Component Performance Study: Motor-Driven Pumps 1998-2014

International Nuclear Information System (INIS)

Schroeder, John Alton

2016-01-01

This report presents an enhanced performance evaluation of motor-driven pumps at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The motor-driven pump failure modes considered for standby systems are failure to start, failure to run less than or equal to one hour, and failure to run more than one hour; for normally running systems, the failure modes considered are failure to start and failure to run. An eight hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified in pump run hours per reactor year. Statistically significant decreasing trends were identified for standby systems industry-wide frequency of start demands, and run hours per reactor year for runs of less than or equal to one hour.

Modulating the level of components within plants

Science.gov (United States)

Bobzin, Steven Craig; Apuya, Nestor; Chiang, Karen; Doukhanina, Elena; Feldmann, Kenneth; Jankowski, Boris; Margolles-Clark, Emilio; Mumenthaler, Daniel; Okamuro, Jack; Park, Joon-Hyun; Van Fleet, Jennifer E.; Zhang, Ke

2017-09-12

Materials and Methods for identifying lignin regulatory region-regulatory protein associations are disclosed. Materials and methods for modulating lignin accumulation are also disclosed. In addition, methods and materials for modulating (e.g., increasing or decreasing) the level of a component (e.g., protein, oil, lignin, carbon, a carotenoid, or a triterpenoid) in plants are disclosed.

Automated ultrasonic inspection of nuclear plant components

International Nuclear Information System (INIS)

Baron, J.A.; Dolbey, M.P.

1982-01-01

For reasons of safety and efficiency, automated systems are used in performing ultrasonic inspection of nuclear components. An automated system designed specifically for the inspection of headers in a nuclear plant is described. In-service inspection results obtained with this system are shown to correlate with pre-service inspection results obtained by manual methods

Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

International Nuclear Information System (INIS)

Ware, A.G.; Morton, D.K.; Nitzel, M.E.

1995-03-01

Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four U.S. nuclear steam supply system vendors. For each facility, six locations were studied, including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This report discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

Solving Component Structural Dynamic Failures Due to Extremely High Frequency Structural Response on the Space Shuttle Program

Science.gov (United States)

Frady, Greg; Nesman, Thomas; Zoladz, Thomas; Szabo, Roland

2010-01-01

For many years, the capabilities to determine the root-cause failure of component failures have been limited to the analytical tools and the state of the art data acquisition systems. With this limited capability, many anomalies have been resolved by adding material to the design to increase robustness without the ability to determine if the design solution was satisfactory until after a series of expensive test programs were complete. The risk of failure and multiple design, test, and redesign cycles were high. During the Space Shuttle Program, many crack investigations in high energy density turbomachines, like the SSME turbopumps and high energy flows in the main propulsion system, have led to the discovery of numerous root-cause failures and anomalies due to the coexistences of acoustic forcing functions, structural natural modes, and a high energy excitation, such as an edge tone or shedding flow, leading the technical community to understand many of the primary contributors to extremely high frequency high cycle fatique fluid-structure interaction anomalies. These contributors have been identified using advanced analysis tools and verified using component and system tests during component ground tests, systems tests, and flight. The structural dynamics and fluid dynamics communities have developed a special sensitivity to the fluid-structure interaction problems and have been able to adjust and solve these problems in a time effective manner to meet budget and schedule deadlines of operational vehicle programs, such as the Space Shuttle Program over the years.

Replacement of major nuclear power plant components for service life extension

International Nuclear Information System (INIS)

Novak, S.

1987-01-01

Problems are discussed associated with replacement of nuclear power plant components with the aim to extend their original scheduled life. The existing foreign experience shows that it is technically feasible to replace practically all basic components for which the necessity of replacement is established. Data is summed up on the replacement of steam generators in US and West German nuclear power plants showing the duration of the job, the total consumption of manhours, the collective dose equivalent and the cost. Attention is also focused on implemented and projected replacements of circulation pipes in nuclear power plants abroad. Based on these figures, the cost is estimated of the replacement of the reactor vessel and the steam generators for WWER-440 nuclear power plants. The conclusion is arrived at that even based on a conservative estimate, the extension by 20 years of the service life of a nuclear power plant is economically more effective than the construction of a new plant. (Z.M.) 2 tabs., 15 refs., 3 figs

Residual life assessment of major LWR components: NPAR approach and results

International Nuclear Information System (INIS)

Shah, V.N.; Weidenhamer, G.H.; Vora, J.P.

1991-01-01

The nuclear plant aging research (NPAR) program is systematically addressing the technical issues associated with understanding and managing aging of major LWR components. Twenty-one major components have been identified and prioritized according to their relevance to plant safety. Qualitative aging assessment has identified pertinent design features, materials, stressors, environments, aging mechanisms. and failure modes for each of the components. Emerging inspection, surveillance, and monitoring methods to characterize aging damage and mitigation methods to reduce the damage are currently being assessed. The results of all these assessments are used to develop life-assessment procedures for the components and are included in appropriate documents supporting the regulatory requirements for license renewal. (author)

[Content and distribution of active components in cultivated and wild Taxus chinensis var. mairei plants].

Science.gov (United States)

Yu, Shao-Shuai; Sun, Qi-Wu; Zhang, Xiao-Ping; Tian, Sheng-Ni; Bo, Pei-Lei

2012-10-01

Taxus chinensis var. mairei is an endemic and endangered plant species in China. The resources of T. chinensis var. mairei have been excessively exploited due to its anti-cancer potential, accordingly, the extant T. chinensis var. mairei population is decreasing. In this paper, ultrasonic extraction and HPLC were adopted to determine the contents of active components paclitaxel, 7-xylosyltaxol and cephalomannine in cultivated and wild T. chinensis var. mairei plants, with the content distribution of these components in different parts of the plants having grown for different years and at different slope aspects investigated. There existed obvious differences in the contents of these active components between cultivated and wild T. chinensis var. mairei plants. The paclitaxel content in the wild plants was about 0.78 times more than that in the cultivated plants, whereas the 7-xylosyltaxol and cephalomannine contents were slishtly higher in the cultivated plants. The differences in the three active components contents between different parts and tree canopies of the plants were notable, being higher in barks and upper tree canopies. Four-year old plants had comparatively higher contents of paclitaxel, 7-xylosyltaxol and cephalomannine (0.08, 0.91 and 0.32 mg x g(-1), respectively), and the plants growing at sunny slope had higher contents of the three active components, with significant differences in the paclitaxel and 7-xylosyltaxol contents and unapparent difference in the cephalomannine content of the plants at shady slope. It was suggested that the accumulation of the three active components in T. chinensis var. mairei plants were closely related to the sunshine conditions. To appropriately increase the sunshine during the artificial cultivation of T. chinensis var. mairei would be beneficial to the accumulation of the three active components in T. chinensis var. mairei plants.

Process pump operating problems and equipment failures, F-Canyon Reprocessing Facility, Savannah River Plant

International Nuclear Information System (INIS)

Durant, W.S.; Starks, J.B.; Galloway, W.D.

1987-02-01

A compilation of operating problems and equipment failures associated with the process pumps in the Savannah River Plant F-Canyon Fuel Reprocessing Facility is presented. These data have been collected over the 30-year operation of the facility. An analysis of the failure rates of the pumps is also presented. A brief description of the pumps and the data bank from which the information was sorted is also included

Analysis of Millstone Unit 1 system failure and maintenance data

International Nuclear Information System (INIS)

Bickel, J.H.; Beveridge, R.L.; Jain, N.K.; Owens, D.B.; Radder, J.A.

1985-01-01

As a result of a task force plan developed four years ago at Northeast Utilities, plant-specific probabilistic safety analysis models are being developed for all Northeast Utilities operating nuclear plants. An essential feature of these models is their reliance on plant-specific reliability information to the maximum extent possible. This assures that future design efforts and decisions on backfitting or procedure changes are made with full knowledge of existing plant reliability. The use of plant-specific reliability data assures that the impacts of problem components are given appropriate attention and that proper credit is given for those components, which because of plant-specific maintenance practices, have exhibited better than industry average performance. A case study of a portion of the Millstone-1 cooling system demonstrates differing results obtained by fault tree analysis and a reliability analysis using plant-specific failure data. When risk assessment techniques are being applied in resource allocation, usage of plant data clearly becomes essential for sound decision making

ASSESSMENT OF DYNAMIC PRA TECHNIQUES WITH INDUSTRY AVERAGE COMPONENT PERFORMANCE DATA

Energy Technology Data Exchange (ETDEWEB)

Yadav, Vaibhav; Agarwal, Vivek; Gribok, Andrei V.; Smith, Curtis L.

2017-06-01

In the nuclear industry, risk monitors are intended to provide a point-in-time estimate of the system risk given the current plant configuration. Current risk monitors are limited in that they do not properly take into account the deteriorating states of plant equipment, which are unit-specific. Current approaches to computing risk monitors use probabilistic risk assessment (PRA) techniques, but the assessment is typically a snapshot in time. Living PRA models attempt to address limitations of traditional PRA models in a limited sense by including temporary changes in plant and system configurations. However, information on plant component health are not considered. This often leaves risk monitors using living PRA models incapable of conducting evaluations with dynamic degradation scenarios evolving over time. There is a need to develop enabling approaches to solidify risk monitors to provide time and condition-dependent risk by integrating traditional PRA models with condition monitoring and prognostic techniques. This paper presents estimation of system risk evolution over time by integrating plant risk monitoring data with dynamic PRA methods incorporating aging and degradation. Several online, non-destructive approaches have been developed for diagnosing plant component conditions in nuclear industry, i.e., condition indication index, using vibration analysis, current signatures, and operational history [1]. In this work the component performance measures at U.S. commercial nuclear power plants (NPP) [2] are incorporated within the various dynamic PRA methodologies [3] to provide better estimates of probability of failures. Aging and degradation is modeled within the Level-1 PRA framework and is applied to several failure modes of pumps and can be extended to a range of components, viz. valves, generators, batteries, and pipes.

Aging assessment of large electric motors in nuclear power plants

International Nuclear Information System (INIS)

Villaran, M.; Subudhi, M.

1996-03-01

Large electric motors serve as the prime movers to drive high capacity pumps, fans, compressors, and generators in a variety of nuclear plant systems. This study examined the stressors that cause degradation and aging in large electric motors operating in various plant locations and environments. The operating history of these machines in nuclear plant service was studied by review and analysis of failure reports in the NPRDS and LER databases. This was supplemented by a review of motor designs, and their nuclear and balance of plant applications, in order to characterize the failure mechanisms that cause degradation, aging, and failure in large electric motors. A generic failure modes and effects analysis for large squirrel cage induction motors was performed to identify the degradation and aging mechanisms affecting various components of these large motors, the failure modes that result, and their effects upon the function of the motor. The effects of large motor failures upon the systems in which they are operating, and on the plant as a whole, were analyzed from failure reports in the databases. The effectiveness of the industry's large motor maintenance programs was assessed based upon the failure reports in the databases and reviews of plant maintenance procedures and programs

Nuclear Plant Aging Research (NPAR) program plan: Components, systems, and structures

International Nuclear Information System (INIS)

1987-09-01

The nuclear plant aging research described in this plan is intended to resolve issues related to the aging and service wear of equipment and systems and major components at commercial reactor facilities and their possible impact on plant safety. Emphasis has been placed on identification and characterization of the mechanisms of material and component degradation during service and evaluation of methods of inspection, surveillance, condition monitoring, and maintenance as means of mitigating such effects. Specifically, the goals of the program are as follows: (1) to identify and characterize aging and service wear effects which, if unchecked, could cause degradation of equipment, a systems, and major components and thereby impair plant safety; (2) to identify methods of inspection, surveillance, and monitoring, or of evaluating residual life of equipment, systems, and major components, which will ensure timely detection of significant aging effects prior to loss of safety function; and (3) to evaluate the effectiveness of storage, maintenance, repair, and replacement practices in mitigating the rate and extent of degradation caused by aging and service wear

Power plant cycle chemistry - a currently neglected power plant chemistry discipline

International Nuclear Information System (INIS)

Bursik, A.

2005-01-01

Power plant cycle chemistry seems to be a stepchild at both utilities and universities and research organizations. It is felt that other power plant chemistry disciplines are more important. The last International Power Cycle Chemistry Conference in Prague may be cited as an example. A critical review of the papers presented at this conference seems to confirm the above-mentioned statements. This situation is very unsatisfactory and has led to an increasing number of component failures and instances of damage to major cycle components. Optimization of cycle chemistry in fossil power plants undoubtedly results in clear benefits and savings with respect to operating costs. It should be kept in mind that many seemingly important chemistry-related issues lose their importance during forced outages of units practicing faulty plant cycle chemistry. (orig.)

Properties of parameter estimation techniques for a beta-binomial failure model. Final technical report

International Nuclear Information System (INIS)

Shultis, J.K.; Buranapan, W.; Eckhoff, N.D.

1981-12-01

Of considerable importance in the safety analysis of nuclear power plants are methods to estimate the probability of failure-on-demand, p, of a plant component that normally is inactive and that may fail when activated or stressed. Properties of five methods for estimating from failure-on-demand data the parameters of the beta prior distribution in a compound beta-binomial probability model are examined. Simulated failure data generated from a known beta-binomial marginal distribution are used to estimate values of the beta parameters by (1) matching moments of the prior distribution to those of the data, (2) the maximum likelihood method based on the prior distribution, (3) a weighted marginal matching moments method, (4) an unweighted marginal matching moments method, and (5) the maximum likelihood method based on the marginal distribution. For small sample sizes (N = or < 10) with data typical of low failure probability components, it was found that the simple prior matching moments method is often superior (e.g. smallest bias and mean squared error) while for larger sample sizes the marginal maximum likelihood estimators appear to be best

Methodology and application of surrogate plant PRA analysis to the Rancho Seco Power Plant: Final report

International Nuclear Information System (INIS)

Gore, B.F.; Huenefeld, J.C.

1987-07-01

This report presents the development and the first application of generic probabilistic risk assessment (PRA) information for identifying systems and components important to public risk at nuclear power plants lacking plant-specific PRAs. A methodology is presented for using the results of PRAs for similar (surrogate) plants, along with plant-specific information about the plant of interest and the surrogate plants, to infer important failure modes for systems of the plant of interest. This methodology, and the rationale on which it is based, is presented in the context of its application to the Rancho Seco plant. The Rancho Seco plant has been analyzed using PRA information from two surrogate plants. This analysis has been used to guide development of considerable plant-specific information about Rancho Seco systems and components important to minimizing public risk, which is also presented herein

Evaluation of a Kalman filter based power pressurizer instrument failure detection system implemented on a nuclear power plant training simulator

International Nuclear Information System (INIS)

Seegmiller, D.S.

1984-01-01

The usefulness of a nuclear power plant training simulator for developing and testing modern estimation and control applications for nuclear power plants is demonstrated. A Kalman filter based instrument failure detection technique for a pressurized water reactor pressurizer is implemented on the Department of Energy N Reactor Training Simulator. This real-time failure detection method computes the first two moments (mean and variance) of each element of a normalized filter innovations vector. Failed pressurizer instrumentation can be detected by comparing these moments to the known statistical properties of the steady state, linear Kalman fitler innovations sequence. The capabilities of the detection system are evaluated using simulated plant transients and instrument failures

Reliability of piping system components. Framework for estimating failure parameters from service data

International Nuclear Information System (INIS)

Nyman, R.; Hegedus, D.; Tomic, B.; Lydell, B.

1997-12-01

This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed 'PFCA'-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies

Reliability of piping system components. Framework for estimating failure parameters from service data

Energy Technology Data Exchange (ETDEWEB)

Nyman, R [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hegedus, D; Tomic, B [ENCONET Consulting GesmbH, Vienna (Austria); Lydell, B [RSA Technologies, Vista, CA (United States)

1997-12-01

This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed `PFCA`-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies. 63 refs, 30 tabs, 22 figs.

Component fragilities. Data collection, analysis and interpretation

International Nuclear Information System (INIS)

Bandyopadhyay, K.K.; Hofmayer, C.H.

1985-01-01

As part of the component fragility research program sponsored by the US NRC, BNL is involved in establishing seismic fragility levels for various nuclear power plant equipment with emphasis on electrical equipment. To date, BNL has reviewed approximately seventy test reports to collect fragility or high level test data for switchgears, motor control centers and similar electrical cabinets, valve actuators and numerous electrical and control devices, e.g., switches, transmitters, potentiometers, indicators, relays, etc., of various manufacturers and models. BNL has also obtained test data from EPRI/ANCO. Analysis of the collected data reveals that fragility levels can best be described by a group of curves corresponding to various failure modes. The lower bound curve indicates the initiation of malfunctioning or structural damage, whereas the upper bound curve corresponds to overall failure of the equipment based on known failure modes occurring separately or interactively. For some components, the upper and lower bound fragility levels are observed to vary appreciably depending upon the manufacturers and models. For some devices, testing even at the shake table vibration limit does not exhibit any failure. Failure of a relay is observed to be a frequent cause of failure of an electrical panel or a system. An extensive amount of additional fregility or high level test data exists

Probability of inadvertent operation of electrical components in harsh environments

International Nuclear Information System (INIS)

Knoll, A.

1989-01-01

Harsh environment, which means humidity and high temperature, may and will affect unsealed electrical components by causing leakage ground currents in ungrounded direct current systems. The concern in a nuclear power plant is that such harsh environment conditions could cause inadvertent operation of normally deenergized components, which may have a safety-related isolation function. Harsh environment is a common cause failure, and one way to approach the problem is to assume that all the unsealed electrical components will simultaneously and inadvertently energize as a result of the environmental common cause failure. This assumption is unrealistically conservative. Test results indicated that insulating resistences of any terminal block in harsh environments have a random distribution in the range of 1 to 270 kΩ, with a mean value ∼59 kΩ. The objective of this paper is to evaluate a realistic conditional failure probability for inadvertent operation of electrical components in harsh environments. This value will be used thereafter in probabilistic safety evaluations of harsh environment events and will replace both the overconservative common cause probability of 1 and the random failure probability used for mild environments

International Context Regarding Application of Single Failure Criterion For New Reactors

International Nuclear Information System (INIS)

Basic, I.; Vrbanic, I.

2016-01-01

The Single Failure Criterion (SFC) ensures reliable performance of safety systems in nuclear power plants in response to design basis initiating events. The SFC, basically, requires that the system must be capable of performing its task in the presence of any single failure. The capability of a system to perform its design function in the presence of a single failure could be threatened by a common cause failure such as a fire, flood, or human intervention or by any other cause with potential to induce multiple failures. When applied to plant's response to a postulated design-basis initiating event, the SFC usually represents a requirement that particular safety system performs its safety functions as designed under the conditions which can include: All failures caused by a single failure; All identifiable but non-detectable failures, including those in the non-tested components; All failures and spurious system actions that cause (or are caused by) the postulated event. The paper provides an overview of the regulatory design requirements for new reactors addressing Single Failure Criterion (SFC) in accordance to international best-practices, particularly considering the SCF relation to in-service testing, maintenance, repair, inspection and monitoring of systems, structures and components important to safety. The paper discusses the comparison of the current SFC requirements and guidelines published by the IAEA, WENRA, EUR and nuclear regulators in the United States, United Kingdom, Russia, Korea, Japan, China and Finland. Also, paper addresses the application of SFC requirements in design; considerations for testing, maintenance, repair, inspection and monitoring; allowable equipment outage times; exemptions to SFC requirements; and analysis for SFC application to two-, three- and four-train systems and applications for small and modular reactors. (author).

Failure mode and effect analysis on safety critical components of space travel

Directory of Open Access Journals (Sweden)

Kouroush Jenab

2015-07-01

Full Text Available Sending men to space has never been an ordinary activity, it requires years of planning and preparation in order to have a chance of success. The payoffs of reliable and repeatable space flight are many, including both Commercial and Military opportunities. In order for reliable and repeatable space flight to become a reality, catastrophic failures need to be detected and mitigated before they occur. It can be shown that small pieces of a design which seem ordinary can create devastating impacts if not designed and tested properly. This paper will address the use of a Failure Mode, Effects, and Criticality Analysis (FMECA with modified Risk Priority Number (RPN and its application to safety critical design components of shuttle liftoff. An example will be presented here which specifically focuses on the Solid Rocket Boosters (SRBs to illustrate the FMECA approach to reliable space travel.

Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, B.; Song, J.

1996-03-01

The Structural Aging Program is addressing the potential for degradation of concrete structural components and systems in nuclear power plants over time due to aging and aggressive environmental stressors. Structures are passive under normal operating conditions but play a key role in mitigating design-basis events, particularly those arising from external challenges such as earthquakes, extreme winds, fires and floods. Structures are plant-specific and unique, often are difficult to inspect, and are virtually impossible to replace. The importance of structural failures in accident mitigation is amplified because such failures may lead to common-cause failures of other components. Structural condition assessment and service life prediction must focus on a few critical components and systems within the plant. Components and systems that are dominant contributors to risk and that require particular attention can be identified through the mathematical formalism of a probabilistic risk assessment, or PRA. To illustrate, the role of structural degradation due to aging on plant risk is examined through the framework of a Level 1 seismic PRA of a nuclear power plant. Plausible mechanisms of structural degradation are found to increase the core damage probability by approximately a factor of two

Penstock failure detection system at the 'Valsan' hydro power plant

International Nuclear Information System (INIS)

Georgescu, A M; Coşoiu, C I; Alboiu, N; Hlevca, D; Tataroiu, R; Popescu, O

2012-01-01

'Valsan' is a small Hydro Power Plant, 5 MW, situated at about 160 km north of Bucharest, Romania, on the small 'Valsan' river in a remote mountainous area. It is equipped with a single Francis turbine. The penstock is located in the access shaft of the HPP. 'Hidroelectrica', the Romanian company that operates the HPP, was trying to implement a remote penstock failure detection system. Starting from a classic hydraulic problem, the authors of the paper derived a method for failure detection and localization on the pipe. The method assumes the existence of 2 flow meters and 2 pressure transducers at the inlet and outlet of the pressurized pipe. Calculations have to be based on experimental values measured in a permanent regime for different values of the flow rate. The method was at first tested on a pipe, in the Hydraulic Laboratory of the Technical University of Civil Engineering Bucharest. Pipe failure was modelled by opening of a valve on a tee branch of the analyzed pipe. Experimental results were found to be in good agreement with theoretical ones. The penstock of the 'Valsan' HPP, was modelled in EPANET, in order to: i) test the method at a larger scale; ii) get the right flow and pressure transducers that are needed to implement it. At the request of 'Hidroelectrica' a routine that computes the efficiency of the turbine was added to the monitoring software. After the system was implemented, another series of measurements were performed at the site in order to validate it. Failure was modelled by opening an existing valve on a branch of the penstock. Detection of the failure was correct and almost instantaneous, while failure location was accurate within 5% of the total penstock length.

Penstock failure detection system at the "Valsan" hydro power plant

Science.gov (United States)

Georgescu, A. M.; Coşoiu, C. I.; Alboiu, N.; Hlevca, D.; Tataroiu, R.; Popescu, O.

2012-11-01

"Valsan" is a small Hydro Power Plant, 5 MW, situated at about 160 km north of Bucharest, Romania, on the small "Valsan" river in a remote mountainous area. It is equipped with a single Francis turbine. The penstock is located in the access shaft of the HPP. "Hidroelectrica", the Romanian company that operates the HPP, was trying to implement a remote penstock failure detection system. Starting from a classic hydraulic problem, the authors of the paper derived a method for failure detection and localization on the pipe. The method assumes the existence of 2 flow meters and 2 pressure transducers at the inlet and outlet of the pressurized pipe. Calculations have to be based on experimental values measured in a permanent regime for different values of the flow rate. The method was at first tested on a pipe, in the Hydraulic Laboratory of the Technical University of Civil Engineering Bucharest. Pipe failure was modelled by opening of a valve on a tee branch of the analyzed pipe. Experimental results were found to be in good agreement with theoretical ones. The penstock of the "Valsan" HPP, was modelled in EPANET, in order to: i) test the method at a larger scale; ii) get the right flow and pressure transducers that are needed to implement it. At the request of "Hidroelectrica" a routine that computes the efficiency of the turbine was added to the monitoring software. After the system was implemented, another series of measurements were performed at the site in order to validate it. Failure was modelled by opening an existing valve on a branch of the penstock. Detection of the failure was correct and almost instantaneous, while failure location was accurate within 5% of the total penstock length.

A pragmatic approach to estimate alpha factors for common cause failure analysis

International Nuclear Information System (INIS)

Hassija, Varun; Senthil Kumar, C.; Velusamy, K.

2014-01-01

Highlights: • Estimation of coefficients in alpha factor model for common cause analysis. • A derivation of plant specific alpha factors is demonstrated. • We examine sensitivity of common cause contribution to total system failure. • We compare beta factor and alpha factor models for various redundant configurations. • The use of alpha factors is preferable, especially for large redundant systems. - Abstract: Most of the modern technological systems are deployed with high redundancy but still they fail mainly on account of common cause failures (CCF). Various models such as Beta Factor, Multiple Greek Letter, Binomial Failure Rate and Alpha Factor exists for estimation of risk from common cause failures. Amongst all, alpha factor model is considered most suitable for high redundant systems as it arrives at common cause failure probabilities from a set of ratios of failures and the total component failure probability Q T . In the present study, alpha factor model is applied for the assessment of CCF of safety systems deployed at two nuclear power plants. A method to overcome the difficulties in estimation of the coefficients viz., alpha factors in the model, importance of deriving plant specific alpha factors and sensitivity of common cause contribution to the total system failure probability with respect to hazard imposed by various CCF events is highlighted. An approach described in NUREG/CR-5500 is extended in this study to provide more explicit guidance for a statistical approach to derive plant specific coefficients for CCF analysis especially for high redundant systems. The procedure is expected to aid regulators for independent safety assessment

Investigations of inter-system common cause failures

International Nuclear Information System (INIS)

Nonclerca, P.; Gallois, M.; Vasseur, D.

2012-01-01

Intra-system common-cause failures (CCF) are widely studied and addressed in existing PSA models, but the information and studies that incorporate the potential for inter-system CCF is limited. However, the French Safety Authority has requested that EDF investigate the possibility of common-cause failure across system boundaries for Flamanville 3 (an EPR design). Also, the modeling of inter-system CCF, or the proof that their impact is negligible, would satisfy Capability Category III for one of the requirements in the ASME/ANS PRA standard in the U.S. EDF and EPRI have been working on a method to assess when it is necessary to take into account inter-system CCF in a PSA model between 2008 and 2010. This method is based both on the likelihood of inter-system CCF and on its demonstrated potential impact on CDF (core damage frequency). This method was first applied on pumps in different systems of the 900 MWe series plants. The second application concerned the motor-operated valves across different systems, using the same PSA model. This second application helped us refine the method, which was not optimal when the number of concerned components is very large. Since then, the method has been successfully applied on the pumps and 10 kV breakers of the EPR power plant in Flamanville. This paper describes the method and the results obtained in some of these studies. All studies have shown either that components in different systems, when they were not already part of a common cause failure group in the model, are not susceptible to common causes of failure, or that the potential for inter-system common-cause failure is negligible regarding the overall risk. (authors)

A survival analysis on critical components of nuclear power plants

International Nuclear Information System (INIS)

Durbec, V.; Pitner, P.; Riffard, T.

1995-06-01

Some tubes of heat exchangers of nuclear power plants may be affected by Primary Water Stress Corrosion Cracking (PWSCC) in highly stressed areas. These defects can shorten the lifetime of the component and lead to its replacement. In order to reduce the risk of cracking, a preventive remedial operation called shot peening was applied on the French reactors between 1985 and 1988. To assess and investigate the effects of shot peening, a statistical analysis was carried on the tube degradation results obtained from in service inspection that are regularly conducted using non destructive tests. The statistical method used is based on the Cox proportional hazards model, a powerful tool in the analysis of survival data, implemented in PROC PHRED recently available in SAS/STAT. This technique has a number of major advantages including the ability to deal with censored failure times data and with the complication of time-dependant co-variables. The paper focus on the modelling and a presentation of the results given by SAS. They provide estimate of how the relative risk of degradation changes after peening and indicate for which values of the prognostic factors analyzed the treatment is likely to be most beneficial. (authors). 2 refs., 3 figs., 6 tabs

System Function Evaluation due to Hardware Failure of NSSS Control Systems in the APR1400

International Nuclear Information System (INIS)

Kim, Juyoung; Ahn, Myunghoon; Kim, Woogoon; Yim, Hyeongsoon

2016-01-01

As the performance and failure modes of the control systems may affect the plant response to accidents or disturbances, an evaluation is done to identify potential control system failure modes resulting from single hardware failures. These failure modes are for use in the analytical evaluations that will be performed to assess the plant responses to various disturbances from the viewpoint of postulated system malfunctions. Failure modes that fall into any of the above categories will affect the performance of the control system and should be considered in the analytical evaluation of the NSSS responses to disturbances. An evaluation was performed to identify the failure modes of the NSSS Control Systems, caused by a hardware component, a common sensing device, and a common power supply. The multiple failure modes across the NSSS control Systems are limited by the improved design features, redundancy within each systems, and segmentation between systems. Also, the effects from the failure modes are expected to be acceptably terminated by the Plant Protection System. The failure modes derived through this evaluation will be further considered in the analytical evaluation of the NSSS responses to disturbances in order to identify the single failures which could create the most adverse conditions during a given transient

Enhanced Component Performance Study: Motor-Driven Pumps 1998–2014

Energy Technology Data Exchange (ETDEWEB)

Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-02-01

This report presents an enhanced performance evaluation of motor-driven pumps at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The motor-driven pump failure modes considered for standby systems are failure to start, failure to run less than or equal to one hour, and failure to run more than one hour; for normally running systems, the failure modes considered are failure to start and failure to run. An eight hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified in pump run hours per reactor year. Statistically significant decreasing trends were identified for standby systems industry-wide frequency of start demands, and run hours per reactor year for runs of less than or equal to one hour.

The qualification of electrical components and instrumentations relevant to safety; La qualificazione dei componenti elettrici e di strumentazione rilevanti per la sicurezza

Energy Technology Data Exchange (ETDEWEB)

Zambardi, F [ENEA - Direzione Sicurezza Nucleare e Protezione Sanitaria, Divisione Sistemi Elettrici e Strumentazione, Rome (Italy)

1989-03-15

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and peculiar aspects are discussed and the qualification by testing is taken into special account. Technical and organizational aspects related to a plant qualification program are also focused. The report ends with a look to the most significant research and development activities. (author)

Analysis of failure and maintenance experiences of motor operated valves in a Finnish nuclear power plant

International Nuclear Information System (INIS)

Simola, K.; Laakso, K.

1992-01-01

Operating experiences from 1981 up to 1989 of totally 104 motor operated closing valves (MOV) in different safety systems at TVO I and II nuclear power units were analysed in a systematic way. The qualitative methods used were failure mode and effects analysis (FMEA) and maintenance effects and criticality analysis (MECA). The failure descriptions were obtained from power plant's computerized failure reporting system. The reported 181 failure events were reanalysed and sorted according to specific classifications developed for the MOV function. Filled FMEA and MECA sheets on individual valves were stored in a microcomputer data base for further analyses. Analyses were performed for the failed mechanical and electrical valve parts, ways of detection of failure modes, failure effects, and repair and unavailability times

Enhanced Component Performance Study: Turbine-Driven Pumps 1998–2014

Energy Technology Data Exchange (ETDEWEB)

Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-11-01

This report presents an enhanced performance evaluation of turbine-driven pumps (TDPs) at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The TDP failure modes considered are failure to start (FTS), failure to run less than or equal to one hour (FTR=1H), failure to run more than one hour (FTR>1H), and normally running systems FTS and failure to run (FTR). The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified for TDP unavailability, for frequency of start demands for standby TDPs, and for run hours in the first hour after start. Statistically significant decreasing trends were identified for start demands for normally running TDPs, and for run hours per reactor critical year for normally running TDPs.

Failure analysis of motor bearing of sea water pump in nuclear power plant

International Nuclear Information System (INIS)

Bian Chunhua; Zhang Wei

2015-01-01

The motor bearing of sea water pump in Qinshan Phase II Nuclear Power plant broke after only one year's using. This paper introduces failure analysis process of the motor bearing. Chemical composition analysis, metallic phase analysis, micrographic examination, and hardness analysis, dimension analysis of each part of the bearing, as well as the high temperature and low temperature performance analysis of lubricating grease are performed. According to the analysis above mentioned, the failure mode of the bearing is wearing, and the reason of wearing is inappropriate installation of the bearing. (authors)

Periodic inspection of CANDU nuclear power plant containment components

International Nuclear Information System (INIS)

1989-09-01

This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

PSA methodology including new design, operational and safety factors, 'Level of recognition of phenomena with a presumed dominant influence upon operational safety' (failures of conventional as well as non-conventional passive components, dependent failures, influence of operator, fires and external threats, digital control, organizational factors)

International Nuclear Information System (INIS)

Jirsa, P.

2001-10-01

The document represents a specific type of discussion of existing methodologies for the creation and application of probabilistic safety assessment (PSA) in light of the EUR document summarizing requirements placed by Western European NPP operators on the future design of nuclear power plants. A partial goal of this discussion consists in mapping, from the PSA point of view, those selected design, operational and/or safety factors of future NPPs that may be entirely new or, at least, newly addressed. Therefore, the terms of reference for this stage were formulated as follows: Assess current level of knowledge and procedures in the analysis of factors and phenomena with a dominant influence upon operational safety of new generation reactors, especially in the following areas: (1) Phenomenology of failure types and mechanisms and reliability of conventional passive safety system components; (2) Phenomenology of failure types and mechanisms and reliability of non-conventional passive components of newly designed safety systems; (3) Phenomenology of types and mechanisms of dependent failures; (4) Human factor role in new generation reactors and its effect upon safety; (5) Fire safety and other external threats to new nuclear installations; (6) Reliability of the digital systems of the I and C system and their effect upon safety; and (7) Organizational factors in new nuclear installations. (P.A.)

Life-time management for mechanical components; Lebensdauermanagement mechanischer Komponenten

Energy Technology Data Exchange (ETDEWEB)

Roos, E. [Stuttgart Univ. (DE). Materialpruefungsanstalt (MPA)

2006-07-01

The safety and economic efficiency of industrial systems depend on the quality of components and systems. In the field of power generation, power plants should be safe and have high availability and minimum specific generation cost. Life management is essential for this. Depending on the safety relevance of systems, structures and components (SSC), this includes proofs of integrity, time-oriented or condition-oriented preventive maintenance, or just failure-oriented maintenance. (orig.)

Detection and mitigation of aging effects of nuclear power plant components

International Nuclear Information System (INIS)

Pachner, J.

1988-09-01

This paper describes the general principles of the methods for timely detection and mitigation of aging effects. These methods include condition monitoring, failure trending, system reliability monitoring, predictive maintenance and scheduled maintenance. In addition, developments of existing detection and mitigation methods needed to improve the capability for effective managing of nuclear power plant aging are discussed

The plant-specific impact of different pressurization rates in the probabilistic estimation of containment failure modes

International Nuclear Information System (INIS)

Ahn, Kwang Il; Yang, Joon Eon; Ha, Jae Joo

2003-01-01

The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities

A methodology for on-line fatigue life monitoring of Indian nuclear power plant components

International Nuclear Information System (INIS)

Mukhopadhyay, N.K.; Dutta, B.K.; Kushawaha, H.S.

1992-01-01

Fatigue is one of the most important aging effects of nuclear power plant components. Information about accumulation of fatigue helps in assessing structural degradation of the components. This assists in-service inspection and maintenance and may also support future life extension program of a plant. In the present report a methodology is being proposed for monitoring on line fatigue life of nuclear power plant components using available plant instrumentations. Major factors affecting fatigue life of a nuclear power plant components are the fluctuations of temperature, pressure and flow rate. Green's function technique is used in on line fatigue monitoring as computation time is much less than finite element method. A code has been developed which computes temperature and stress Green's functions in 2-D and axisymmetric structure by finite element method due to unit change in various fluid parameters. A post processor has also been developed which computes the temperature and stress responses using corresponding Green's functions and actual fluctuation in fluid parameters. In this post processor, the multiple site problem is solved by superimposing single site Green's function technique. It is also shown that Green's function technique is best suited for on line fatigue life monitoring of nuclear power plant components. (author). 6 refs., 43 figs

The development of specific reliability database for a Korean Nuclear Power Plant

International Nuclear Information System (INIS)

Park, S.K.; Park, B.L.; Kim, M.R.; Jeong, B.H.; Kwon, J.J.

2001-01-01

The object of this study is to develop reliability database for PSA application such as failure rate for safety related components, test and maintenance unavailability and common cause failure factors except for initiating event frequencies during the period of 10 years from 1990 to 1999. In this study we developed plant-specific reliability database for PSA (Probabilistic Safety Assessment) application and compared it with generic reliability database developed in the US such as EPRI-URD, IEEE-500, NUCLARR etc, in the component type basis. We have found that there are some general differences in the component failure rate and test and maintenance unavailability. We described the characteristics of differences for some important component types. We also analyzed the reasons for the differences in the aspect of maintenance terms such as maintenance policy and maintenance practice. We found that maintenance terms are important factors for the numbers of plant-specific reliability database. (author)

The condition monitoring system of turbine system components for nuclear power plants

International Nuclear Information System (INIS)

Ono, Shigetoshi

2013-01-01

The thermal and nuclear power plants have been imposed a stable supply of electricity. To certainly achieve this, we built the plant condition monitoring system based on the heat and mass balance calculation. If there are some performance changes on the turbine system components of their power plants, the heat and mass balance of the turbine system will change. This system has ability to detect the abnormal signs of their components by finding the changes of the heat and mass balance. Moreover we note that this system is built for steam turbine cycle operating with saturated steam conditions. (author)

Investigations of the Failure in Boilers Economizer Tubes Used in Power Plants

Science.gov (United States)

Moakhar, Roozbeh Siavash; Mehdipour, Mehrad; Ghorbani, Mohammad; Mohebali, Milad; Koohbor, Behrad

2013-09-01

In this study, failure of a high pressure economizer tube of a boiler used in gas-Mazut combined cycle power plants was studied. Failure analysis of the tube was accomplished by taking into account visual inspection, thickness measurement, and hardness testing as well as microstructural observations using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and x-ray diffraction (XRD). Optical microscopy images indicate that there is no phase transformation during service, and ferrite-pearlite remained. The results of XRD also revealed Iron sulfate (FeSO4) and Iron hydroxide sulfate (FeOH(SO4)) phases formed on the steel surface. A considerable amount of Sulfur was also detected on the outer surface of the tube by EDS analysis. Dew-point corrosion was found to be the principal reason for the failure of the examined tube while it has been left out-of-service.

Systematic generation of rules for nuclear power plant diagnostics

International Nuclear Information System (INIS)

Reifman, J.; Lee, J.C.

1988-01-01

The knowledge base of an expert system is generally represented by a set of heuristic rules derived from the expert's own experience and judgmental knowledge. These heuristic or production rules are cast as if (condition), then (consequence) statements, and represent, for nuclear power plant diagnostic systems, information connecting symptoms to failures. In this paper, the authors apply an entropy minimax pattern recognition algorithm to automate the process of extracting and encoding knowledge into a set of rules. Knowledge is extracted by recognizing patterns in plant parameters or symptoms associated with failures or transient events, and is encoded by casting the discovered patterns as production rules. The paper discusses how the proposed method can systematically generate rules that characterize failure of pressurizer components based on transient events analyzed with a pressurizer components based on transient events analyzed with a pressurizer water reactor simulator program

A pilot application of risk-based methods to establish in-service inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station

International Nuclear Information System (INIS)

Vo, T.; Gore, B.; Simonen, F.; Doctor, S.

1994-08-01

As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish in-service inspection plans for nuclear power plant components. This method uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot applications of this method. The specific systems addressed in this report are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants. To develop inspection plans, the acceptable level of risk from structural failure for important systems and components will be apportioned as a small fraction (i.e., 5%) of the total PRA-estimated risk for core damage. This process will determine target (acceptable) risk and target failure probability values for individual components. Inspection requirements will be set at levels to assure that acceptable failure probabilistics are maintained

Glycoprotein component of plant cell walls

International Nuclear Information System (INIS)

Cooper, J.B.; Chen, J.A.; Varner, J.E.

1984-01-01

The primary wall surrounding most dicotyledonous plant cells contains a hydroxyproline-rich glycoprotein (HRGP) component named extensin. A small group of glycopeptides solubilized from isolated cell walls by proteolysis contained a repeated pentapeptide glycosylated by tri- and tetraarabinosides linked to hydroxyproline and, by galactose, linked to serine. Recently, two complementary approaches to this problem have provided results which greatly increase the understanding of wall extensin. In this paper the authors describe what is known about the structure of soluble extensin secreted into the walls of the carrot root cells

Component aging illustrated in maintenance histories from IPRDS

International Nuclear Information System (INIS)

Kahl, W.K.; Borkowski, R.J.

1984-01-01

The centerpiece of the In-Plant Reliability Data (IPRD) program is a nuclear power plant component data base developed from corrective maintenance records. The primary objective of this program is to establish a comprehensive source of information for reliability analyses, including the compilation of maintenance histories for failure statistics. The IPRD program is currently a pilot effort that has 36 reactor-years of data on pumps, valves, diesel generators, batteries, battery chargers and inverters from five nuclear plants. The data are collected under the auspices of the Institute of Electrical and Electronics Engineers (IEEE) to certify anonymity of the participating plants, and is funded by the Nuclear Regulatory Commission

Component reliability data for use in probabilistic safety assessment

International Nuclear Information System (INIS)

1988-10-01

Generic component reliability data is indispensable in any probabilistic safety analysis. It is not realistic to assume that all possible component failures and failure modes modeled in a PSA would be available from the operating experience of a specific plant in a statistically meaningful way. The degree that generic data is used in PSAs varies from case to case. Some studies are totally based on generic data while others use generic data as prior information to be specialized by plant specific data. Most studies, however, finally use a combination where data for certain components come from generic data sources and others from Bayesian updating. The IAEA effort to compile a generic component reliability data base aimed at facilitating the use of data available in the literature and at highlighting pitfalls which deserve special consideration. It was also intended to complement the fault tree and event tree package (PSAPACK) and to facilitate its use. Moreover, it should be noted, that the IAEA has recently initiated a Coordinated Research Program in Reliability Data Collection, Retrieval and Analysis. In this framework the issues identified as most affecting the quality of existing data bases would be addressed. This report presents the results of a compilation made from the specialized literature and includes reliability data for components usually considered in PSA

Preventive maintenance-A countermeasure to plant aging

International Nuclear Information System (INIS)

Hlubek, W.

1985-01-01

The aging of power plants is caused by manifold and different influences. For instance, mechanical and thermal stress, radiation exposure, denting or wastage can considerably affect the aging of plant components and thus cause premature failures of components. In this presentation, the term 'Plant Aging' in nuclear power plants is to be understood more comprehensively than wear on components and material fatigue. In addition, nuclear power plants are to be adjusted to the advancing state of the science and technology (state-of-the-art) in order to guarantee safe operation at all times. The preventive maintenance - as a countermeasure to plant aging - comprises the systematic checks and servicing of the plant systems in operation and follows aging by inspection and tests. Experience with Rheinisch-Westfaelisches Elektrizitaetswerk AG (RWE) preventive maintenance program at the Biblis NNP (1300 MW, PWR) is discussed. The concept of an 'Integrated Maintenance System' as a means to avoid 'Plant Aging' is presented

Time-independent and time-dependent contributions to the unavailability of standby safety system components

International Nuclear Information System (INIS)

Lofgren, E.V.; Uryasev, S.; Samanta, P.

1997-01-01

The unavailability of standby safety system components due to failures in nuclear power plants is considered to involve a time-independent and a time-dependent part. The former relates to the component's unavailability from demand stresses due to usage, and the latter represents the component's unavailability due to standby-time stresses related to the environment. In this paper, data from the nuclear plant reliability data system (NPRDS) were used to partition the component's unavailability into the contributions from standby-time stress (i.e., due to environmental factors) and demand stress (i.e., due to usage). Analyses are presented of motor-operated valves (MOVs), motor-driven pumps (MDPs), and turbine-driven pumps (TDPs). MOVs fail predominantly (approx. 78 %) from environmental factors (standby-time stress failures). MDPs fail slightly more frequently from demand stresses (approx. 63 %) than standby-time stresses, while TDPs fail predominantly from standby-time stresses (approx. 78 %). Such partitions of component unavailability have many uses in risk-informed and performance-based regulation relating to modifications to Technical Specification, in-service testing, precise determination of dominant accident sequences, and implementation of maintenance rules

Filtering technique for detection and identification of measurement failures in nuclear power plants

International Nuclear Information System (INIS)

Racz, A.

1989-11-01

The basic requirement of the safe operation of nuclear power plants (NPP) is to have reliable information on all quantities that can be measured, monitored or controlled during the operation. Kalman filtering techniques have been applied for prompt detection and identification of failures in the measurement systems used in NPPs. Mathematical basis of Kalman filtering and various models applied to failure detection are overviewed. The applicability of some models are evaluated by real results of NPP measurements. A sample system for an NPP is suggested, based on several numerical tests. (R.P.) 23 refs.; 40 figs.; 2 tabs

Microstructures, Forming Limit and Failure Analyses of Inconel 718 Sheets for Fabrication of Aerospace Components

Science.gov (United States)

Sajun Prasad, K.; Panda, Sushanta Kumar; Kar, Sujoy Kumar; Sen, Mainak; Murty, S. V. S. Naryana; Sharma, Sharad Chandra

2017-04-01

Recently, aerospace industries have shown increasing interest in forming limits of Inconel 718 sheet metals, which can be utilised in designing tools and selection of process parameters for successful fabrication of components. In the present work, stress-strain response with failure strains was evaluated by uniaxial tensile tests in different orientations, and two-stage work-hardening behavior was observed. In spite of highly preferred texture, tensile properties showed minor variations in different orientations due to the random distribution of nanoprecipitates. The forming limit strains were evaluated by deforming specimens in seven different strain paths using limiting dome height (LDH) test facility. Mostly, the specimens failed without prior indication of localized necking. Thus, fracture forming limit diagram (FFLD) was evaluated, and bending correction was imposed due to the use of sub-size hemispherical punch. The failure strains of FFLD were converted into major-minor stress space ( σ-FFLD) and effective plastic strain-stress triaxiality space ( ηEPS-FFLD) as failure criteria to avoid the strain path dependence. Moreover, FE model was developed, and the LDH, strain distribution and failure location were predicted successfully using above-mentioned failure criteria with two stages of work hardening. Fractographs were correlated with the fracture behavior and formability of sheet metal.

Canadian programs on understanding and managing aging degradation of nuclear power plant components

International Nuclear Information System (INIS)

Chadha, J.A.; Pachner, J.

1989-06-01

Maintaining adequate safety and reliability of nuclear power plants and nuclear power plant life assurance and life extension are growing in importance as nuclear plants get older. Age-related degradation of plant components is complex and not fully understood. This paper provides an overview of the Canadian approach and the main activities and their results towards understanding and managing age-related degradation of nuclear power plant components, structures and systems. A number of pro-active programs have been initiated to anticipate, detect and mitigate potential aging degradation at an early stage before any serious impact on plant safety and reliability. These programs include Operational Safety Management Program, Nuclear Plant Life Assurance Program, systematic plant condition assessment, refurbishment and upgrading, post-service examination and testing, equipment qualification, research and development, and participation in the IAEA programs on safety aspects of nuclear power plant aging and life extension. A regulatory policy on nuclear power plants is under development and will be based on the domestic as well as foreign and international studies and experience

The material concept in German light water reactors. Contribution to plant safety economic efficiency and failure provision; Das Werkstoffkonzept in deutschen Leichtwasserreaktoren. Beitrag zur Anlagensicherheit, Wirtschaftlichkeit und Schadensvorsorge

Energy Technology Data Exchange (ETDEWEB)

Ilg, Ulf [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Philippsburg; Koenig, Guenter [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Neckarwestheim; Erve, Manfred [AREVA NP GmbH, Erlangen (Germany)

2008-07-01

In the design and construction stage of nuclear power plants relevant decisions may affect the service life of a component, and thus influence safety and availability of the plant. The German ''basic safety concept'' has an important effect on the quality of the BOL (begin of life) status. Materials selection and qualification are of significant importance for the component lifetime and the profitability of the plant. Examples for the implementation of this concept are demonstrated for the steam generator tubing material Incoloy 800, the inside-plated ferritic compound tubes as control rod drive mechanism nozzle through the RPV head of BWR plants that are not susceptible for corrosion enhanced cracking that was observed for Inconel 600 tubing. A fundamental failure analysis of crack formation in Ti stabilized austenitic pipes of BWR plants found since 1993 were definitely identified as intergranular stress corrosion caused by a local sensitization of the welding process induced overheated structured in the heat affected zone. This allowed target-oriented mitigation measures. The safety culture implemented in German nuclear plants in connection with the break preclusion or integrity concept, respectively, including a continuous actualization with respect to the state-of-the art are the technical prerequisites for damage precaution and possible life time extension.

Optimal test intervals of standby components based on actual plant-specific data

International Nuclear Information System (INIS)

Jones, R.B.; Bickel, J.H.

1987-01-01

Based on standard reliability analysis techniques, both under testing and over testing affect the availability of standby components. If tests are performed too often, unavailability is increased since the equipment is being used excessively. Conversely if testing is performed too infrequently, the likelihood of component unavailability is also increased due to the formation of rust, heat or radiation damage, dirt infiltration, etc. Thus from a physical perspective, an optimal test interval should exist which minimizes unavailability. This paper illustrates the application of an unavailability model that calculates optimal testing intervals for components with a failure database. (orig./HSCH)

Instrument failure monitoring in nuclear power systems

International Nuclear Information System (INIS)

Tylee, J.L.

1982-01-01

Methods of monitoring dynamic systems for instrument failures were developed and evaluated. In particular, application of these methods to nuclear power plant components is addressed. For a linear system, statistical tests on the innovations sequence of a Kalman filter driven by all system measurements provides a failure detection decision and identifies any failed sensor. This sequence (in an unfailed system) is zero-mean with calculable covariance; hence, any major deviation from these properties is assumed to be due to an instrument failure. Once a failure is identified, the failed instrument is replaced with an optimal estimate of the measured parameter. This failure accommodation is accomplished using optimally combined data from a bank of accommodation Kalman filters (one for each sensor), each driven by a single measurement. Using such a sensor replacement allows continued system operation under failed conditions and provides a system operator with information otherwise unavailable. To demonstrate monitor performance, a liner failure monitor was developed for the pressurizer in the Loss-of-Fluid Test (LOFT) reactor plant. LOFT is a small-scale pressurized water reactor (PWR) research facility located at the Idaho National Engineering Laboratory. A linear, third-order model of the pressurizer dynamics was developed from first principles and validated. Using data from the LOFT L6 test series, numerous actual and simulated water level, pressure, and temperature sensor failures were employed to illustrate monitor capabilities. Failure monitor design was applied to nonlinear dynamic systems by replacing all monitor linear Kalman filters with extended Kalman filters. A nonlinear failure monitor was derived for LOFT reactor instrumentation. A sixth-order reactor model, including descriptions of reactor kinetics, fuel rod heat transfer, and core coolant dynamics, was obtained and verified with test data

Common cause failure investigations using the European Reliability Data System

International Nuclear Information System (INIS)

Games, A.M.; Breewood, M.; Amendola, A.; Keller, A.Z.

1984-01-01

The European Reliability Data System (ERDS) has provided data for use in investigations into common cause failures (CCFs) in nuclear power plants. These investigations have been made on two levels, at a system and inter-system level. Data have been used from the Component Event Data Bank and from the Licensee Event Report Files, both part of the ERDS. The two studies required different methodologies although both commenced with a temporal sorting procedure for the failure events. The studies demonstrated that different types of common cause failure necessitate different search algorithms, and thus a data search must be closely related to an appropriate CCF classification system, which in the first instance would not be based on causes of failure. (author)

NDT: Replication avoids unnecessary replacement of power plant components

International Nuclear Information System (INIS)

Neubauer, B.; Wedel, U.

1984-01-01

Effective fracture prevention for components operating at high temperatures can be achieved without sacrificing useful life. This is done by nondestructive-metallographic examination at crack-susceptible locations of the components. Creep microcracks approximately one micron in size can be detected. RWTUV experience shows that, in general, the components need not be replaced or repaired until these microcracks have grown to form small creep macrocracks. The long prewarning period before macrocracks form provides assurance of safe operation for the full useful life of the components tested. The economic benefit achieved is considerable. Replication techniques have been widely applied by the authors in operating power plants since 1977. This nondestructive-evaluation method involves polishing small areas of selected piping-system components, preparing replicas of the polished areas, and examining the replicas under microscope for evidence of cavities, microcracks, or macrocracks

TNO experience on sodium cleaning of large plant components by vacuum distillation

Energy Technology Data Exchange (ETDEWEB)

Smit, C Ch [MT-TNO Dept. 50-MW Sodium Component Test Facility, Hengelo (Netherlands)

1978-08-01

The Intermediate Heat Exchanger and Steam generators developed within the framework of the SNR-programme are being tested in the 50 MW Test facility at Hengelo - The Netherlands. The facility was designed and built by Neratoom, and is operated by TNO, the Dutch Organisation for Applied Scientific Research. Sodium technology work, such as reported in this paper, is done in close cooperation with Neratoom and with TNO-laboratories at Apeldoorn, where several smaller sodium rigs and other facilities are available. The operation and maintenance of a large sodium test facility and sodium rigs lead to frequent cleaning of small plant components, test sections and sampling devices. The choice of method usually depends on the size of the component and the cleaning quality needed. The results are predictable and satisfactory. For large components, however, the situation is different. Although the basic cleaning methods using alcohol and moist gas are well-known, and procedures for the cleaning of small components are available, complete cleaning of tight crevices and threaded bolds cannot be guaranteed, and consequently the requalification procedure needs to include a complete disassembly and inspection of the cleaned component. For large components this policy cannot always be followed. In those cases for instance where an in-between internal inspection is required, or where only small modifications of the test object are necessary, other possibilities have to be considered. For this reason some work has been done to develop reliable vacuum distillation procedures for large components, based on the cleaning experience with small plant components. The results of these procedures applied to large plant components are reported in this paper.

TNO experience on sodium cleaning of large plant components by vacuum distillation

International Nuclear Information System (INIS)

Smit, C.Ch.

1978-01-01

The Intermediate Heat Exchanger and Steam generators developed within the framework of the SNR-programme are being tested in the 50 MW Test facility at Hengelo - The Netherlands. The facility was designed and built by Neratoom, and is operated by TNO, the Dutch Organisation for Applied Scientific Research. Sodium technology work, such as reported in this paper, is done in close cooperation with Neratoom and with TNO-laboratories at Apeldoorn, where several smaller sodium rigs and other facilities are available. The operation and maintenance of a large sodium test facility and sodium rigs lead to frequent cleaning of small plant components, test sections and sampling devices. The choice of method usually depends on the size of the component and the cleaning quality needed. The results are predictable and satisfactory. For large components, however, the situation is different. Although the basic cleaning methods using alcohol and moist gas are well-known, and procedures for the cleaning of small components are available, complete cleaning of tight crevices and threaded bolds cannot be guaranteed, and consequently the requalification procedure needs to include a complete disassembly and inspection of the cleaned component. For large components this policy cannot always be followed. In those cases for instance where an in-between internal inspection is required, or where only small modifications of the test object are necessary, other possibilities have to be considered. For this reason some work has been done to develop reliable vacuum distillation procedures for large components, based on the cleaning experience with small plant components. The results of these procedures applied to large plant components are reported in this paper

Evolving inspection technologies for reliable condition assessment of components and plants

International Nuclear Information System (INIS)

Baldev Raj

1994-01-01

Condition assessment of components and plants are being done regularly in many an industry. The methodologies adopted are being continuously refined. However, each of these methodologies are being applied in isolation, without realizing the synergistic advantage we derive when a global approach is taken for condition assessment. Developments in a variety of fields, that have a definite bearing on the reliability of condition assessment, are not applied (or even thought that they could be applied) together. The possible impact of evolving technologies in enhancing the efficiency of condition assessment of components and plants are discussed. (author). 11 refs

Estimation of the common cause failure probabilities of the components under mixed testing schemes

International Nuclear Information System (INIS)

Kang, Dae Il; Hwang, Mee Jeong; Han, Sang Hoon

2009-01-01

For the case where trains or channels of standby safety systems consisting of more than two redundant components are tested in a staggered manner, the standby safety components within a train can be tested simultaneously or consecutively. In this case, mixed testing schemes, staggered and non-staggered testing schemes, are used for testing the components. Approximate formulas, based on the basic parameter method, were developed for the estimation of the common cause failure (CCF) probabilities of the components under mixed testing schemes. The developed formulas were applied to the four redundant check valves of the auxiliary feed water system as a demonstration study for their appropriateness. For a comparison, we estimated the CCF probabilities of the four redundant check valves for the mixed, staggered, and non-staggered testing schemes. The CCF probabilities of the four redundant check valves for the mixed testing schemes were estimated to be higher than those for the staggered testing scheme, and lower than those for the non-staggered testing scheme.

Detecting and mitigating aging in component cooling water systems

International Nuclear Information System (INIS)

Lofaro, R.J.

1991-01-01

The time-dependent effects of aging on component cooling water (CCW) systems in nuclear power plants has been studied and documented as part of a research program sponsored by the US Nuclear Regulatory Commission. It was found that age related degradation leads to failures in the CCW system which can result in an increase in system unavailability, if not properly detected and mitigated. To identify effective methods of managing this degradation, information on inspection, monitoring, and maintenance practices currently available was obtained from various operating plants and reviewed. The findings were correlated with the most common aging mechanisms and failure modes and a compilation of aging detection and mitigation practices was formulated. This paper discusses the results of this work

Detecting and mitigating aging in component cooling water systems

International Nuclear Information System (INIS)

Lofaro, R.J.; Aggarwal, S.

1992-01-01

The time-dependent effects of aging on component cooling water (CCW) systems in nuclear power plants has been studied and documented as part of a research program sponsored by the US Nuclear Regulatory Commission. It was found that age related degradation leads to failures in the CCW system which can result in an increase in system unavailability, if not properly detected and mitigated. To identify effective methods of managing this degradation, information on inspection, monitoring, and maintenance practices currently available was obtained from various operating plants and reviewed. The findings were correlated with the most common aging mechanisms and failure modes, and a compilation of aging detection and mitigation practices was formulated. This paper discusses the results of this work

Root cause analysis in support of reliability enhancement of engineering components

International Nuclear Information System (INIS)

Kumar, Sachin; Mishra, Vivek; Joshi, N.S.; Varde, P.V.

2014-01-01

Reliability based methods have been widely used for the safety assessment of plant system, structures and components. These methods provide a quantitative estimation of system reliability but do not give insight into the failure mechanism. Understanding the failure mechanism is a must to avoid the recurrence of the events and enhancement of the system reliability. Root cause analysis provides a tool for gaining detailed insights into the causes of failure of component with particular attention to the identification of fault in component design, operation, surveillance, maintenance, training, procedures and policies which must be improved to prevent repetition of incidents. Root cause analysis also helps in developing Probabilistic Safety Analysis models. A probabilistic precursor study provides a complement to the root cause analysis approach in event analysis by focusing on how an event might have developed adversely. This paper discusses the root cause analysis methodologies and their application in the specific case studies for enhancement of system reliability. (author)

Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

Energy Technology Data Exchange (ETDEWEB)

Lydell, Bengt (Scandpower Risk Management Inc., Houston, TX (US)); Olsson, Anders (Relcon Scandpower AB, Stockholm (SE))

2008-01-15

This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning

Prediction of accident sequence probabilities in a nuclear power plant due to earthquake events

International Nuclear Information System (INIS)

Hudson, J.M.; Collins, J.D.

1980-01-01

This paper presents a methodology to predict accident probabilities in nuclear power plants subject to earthquakes. The resulting computer program accesses response data to compute component failure probabilities using fragility functions. Using logical failure definitions for systems, and the calculated component failure probabilities, initiating event and safety system failure probabilities are synthesized. The incorporation of accident sequence expressions allows the calculation of terminal event probabilities. Accident sequences, with their occurrence probabilities, are finally coupled to a specific release category. A unique aspect of the methodology is an analytical procedure for calculating top event probabilities based on the correlated failure of primary events

Application of fatigue monitoring system in PWR nuclear power plant

International Nuclear Information System (INIS)

Piao Lei

2014-01-01

Fatigue failure is one form of equipment failure of nuclear power plant, influencing equipment lifetime and lifetime extension. Fatigue monitoring system can track real thermal transient at fatigue sensitive components, establish a basis for fatigue analyses based on realistic operating loads, identify unexpected operational transients, optimize the plant behavior by improved operating modes, provide supporting data for lifetime management, enhance security of plant and reduce economical loss. Fatigue monitoring system has been applied in many plants and is required to be applied in Generation-III nuclear power plant. It is necessary to develop the fatigue monitoring system with independent intellectual property rights and improve the competitiveness of domestic Generation-III nuclear power technology. (author)

Reliability models for a nonrepairable system with heterogeneous components having a phase-type time-to-failure distribution

International Nuclear Information System (INIS)

Kim, Heungseob; Kim, Pansoo

2017-01-01

This research paper presents practical stochastic models for designing and analyzing the time-dependent reliability of nonrepairable systems. The models are formulated for nonrepairable systems with heterogeneous components having phase-type time-to-failure distributions by a structured continuous time Markov chain (CTMC). The versatility of the phase-type distributions enhances the flexibility and practicality of the systems. By virtue of these benefits, studies in reliability engineering can be more advanced than the previous studies. This study attempts to solve a redundancy allocation problem (RAP) by using these new models. The implications of mixing components, redundancy levels, and redundancy strategies are simultaneously considered to maximize the reliability of a system. An imperfect switching case in a standby redundant system is also considered. Furthermore, the experimental results for a well-known RAP benchmark problem are presented to demonstrate the approximating error of the previous reliability function for a standby redundant system and the usefulness of the current research. - Highlights: • Phase-type time-to-failure distribution is used for components. • Reliability model for nonrepairable system is developed using Markov chain. • System is composed of heterogeneous components. • Model provides the real value of standby system reliability not an approximation. • Redundancy allocation problem is used to show usefulness of this model.

Manufacture of piping components for nuclear power plants

International Nuclear Information System (INIS)

Bartecek, R.

1983-01-01

Hammer forging of hollow forging ingots, extrusion and elestroslag remelting may be used for the manufacture of large pipes. Technologies have been developed for the manufacture of elbows based on various types of forming. These procedures mainly include the hydraulic pressing of elbows from tubes and the pressing of symmetrical halves of elbows with subsequent welding. The hammer forging of valves, cross pieces, etc., has been replaced by forging and pressing. In order to prevent failures from occurring in the pipes during operation of nuclear power plants, pipes are being made of larger forgings, which reduces the number of welds. This improves the quality of the pipes, reduces production and assembly costs and is metal-saving. (E.S.)

Recommendations for managing equipment aging in nuclear power plants

International Nuclear Information System (INIS)

Gunther, W.E.; Subudhi, M.; Aggarwal, S.K.

1992-01-01

Research conducted under the auspices of the US NRC's Nuclear Plant Aging Research (NPAR) Program has resulted in a large database of component and system operating, maintenance, and testing information. This database has been used to determine the susceptibility to aging of selected components, and the potential for equipment aging to impact plant safety and availability. it has also identified methods for detecting and mitigating component and system aging. This paper describes the research recommendations on electrical components which could be applied to maintenance, testing, and inspection activities to detect and mitigate the effects of aging prior to equipment failures

Contemporary statistical procedures (Parametric Empirical Bayes) and nuclear plant event rates

International Nuclear Information System (INIS)

Gaver, D.P.; Worledge, D.H.

1985-01-01

The conduct of a nuclear power plant probabilistic risk assessment (PRA) recognizes that each of a great many vital components and systems is subject to failure. One aspect of the PRA procedure is to quantify individual item failure propensity, often in terms of the failure rate parameter of an exponential distribution or Poisson process, and then to combine rates so as to effectively infer the probability of plant failure, e.g., core damage. The formal method of combination of such rates involves use of fault-tree analysis. The defensibility of the final fault-tree result depends both upon the adequacy of the failure representations of its components, and upon the correctness and inclusiveness of the fault tree logic. This paper focuses upon the first issue, in particular, upon contemporary proposals for deriving estimates of individual rates. The purpose of the paper is to present, in basically non-mathematical terms, the essential nature of some of these proposals, and an assessment of how they might fit into, and contribute positively to, a more defensible or trustworthy PRA process

Common cause failure data collection and analysis for safety-related components of TRIGA SSR-14MW Pitesti, Romania

International Nuclear Information System (INIS)

Radu, G.; Mladin, D.

2003-01-01

This paper presents a study performed on the set of common cause failures (CCF) of safety-related components of the research reactor TRIGA SSR-14 MW Pitesti. The data collected cover a period of 20 years, from 1979 to 2000. The sources of data are Shift Supervisor Reports, Work Authorizations, and Reactor Log Books. Events collected are analyzed by failure mode and degrees of failure. Qualitative analysis of root causes, coupling factors and corrective actions and quantitative analysis of CCF events are studied. The objective of this work is to develop qualitative insights in the nature of the reported events and to build a site-specific common cause events database. (author)

Evaluation of the Latest Generic Data for PSA Applications of Domestic Nuclear Plants

International Nuclear Information System (INIS)

Hwang, Seok Won; Oh, Ji Yong; Lee, Byung Sik

2009-01-01

Generic data of domestic PSAs have mostly referred to 'Advanced Light Water Reactor (ALWR) Utility Requirements Document (URD) 'issued by EPRI. Generally, current data of domestic PSA have been customized with the generic and plant specific data through Bayesian analysis. The generic reference has established by collecting US nuclear plant practices from mid 1980s to early 1990s. Over the decade, US plants had showed low performances and capabilities in operation. On the other hand, the current domestic nuclear plants shows world class performance in operation and maintenance compared with the corresponding US nuclear plants in URD. Therefore, it is necessary to apply proper generic sources which can represent the current domestic plant performances and status. In 2007, the latest generic source (NUREG/CR-6928) is published by US NRC, which deals with new types of failure modes and analysis methods. A fundamental improvement in NUREG/CR-6928 compared with previous data source is the distinction between standby and alternating/running component basic events, which shows different failure mechanisms. Significant differences were also noted running failure events occurred within and beyond the first hour for emergency diesel generators, cooling units, and pumps. This was done because the historical perspective on running failure rates indicated approximately a factor of 15 differences between the two failure rates for several component types. ALWR URD uses lognormal distribution in the estimation of failure rates. On the contrary, NUREG/CR-6928 uses beta and gamma distributions for demand and running failures, respectively. This work has proposed an approach to the application of NUREG/CR-6928 to current PSA practice by comparing it with URD data. Moreover, this attempt results in eliciting substantial insights of the establishment of the domestic generic database

Seismic PSA of nuclear power plants a case study

International Nuclear Information System (INIS)

Hari Prasad, M.; Dubey, P.N.; Reddy, G.R.; Saraf, R.K.; Ghosh, A.K.

2006-07-01

Seismic Probabilistic Safety Assessment (Seismic PSA) analysis is an external event PSA analysis. The objective of seismic PSA for the plants is to examine the existence of plant vulnerabilities against postulated earthquakes by numerically assessing the plant safety and to take appropriate measures to enhance the plant safety. Seismic PSA analysis integrates the seismic hazard analysis, seismic response analysis, seismic fragility analysis and system reliability/ accident sequence analysis. In general, the plant consists of normally operating and emergency standby systems and components. The failure during an earthquake (induced directly by excessive inertial stresses or indirectly following the failure of some other item) of an operating component will lead to a change in the state of the plant. In that case, various scenarios can follow depending on the initiating event and the status of other sub-systems. The analysis represents these possible chronological sequences by an event tree. The event trees and the associated fault trees model the sub-systems down to the level of individual components. The procedure has been applied for a typical Indian nuclear power plant. From the internal event PSA level I analysis significant contribution to the Core Damage Frequency (CDF) was found due to the Fire Water System. Hence, this system was selected to establish the procedure of seismic PSA. In this report the different elements that go into seismic PSA analysis have been discussed. Hazard curves have been developed for the site. Fragility curve for the seismically induced failure of Class IV power has been developed. The fragility curve for fire-water piping system has been generated. Event tree for Class IV power supply has been developed and the dominating accident sequences were identified. CDF has been estimated from these dominating accident sequences by convoluting hazard curves of initiating event and fragility curves of the safety systems. (author)

Analysis methods for structure reliability of piping components

International Nuclear Information System (INIS)

Schimpfke, T.; Grebner, H.; Sievers, J.

2004-01-01

In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour (BMWA) GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The long-term objective of this development is to provide failure probabilities of passive components for probabilistic safety analysis of nuclear power plants. Up to now the code can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents some of the results of a benchmark analysis in the frame of the European project NURBIM (Nuclear Risk Based Inspection Methodologies for Passive Components). (orig.)

Defense-in-depth for common cause failure of nuclear power plant safety system software