WorldWideScience

Sample records for plant component failures

  1. 4. Nuclear power plant component failures

    International Nuclear Information System (INIS)

    1990-01-01

    Nuclear power plant component failures are dealt with in relation to reliability in nuclear power engineering. The topics treated include classification of failures, analysis of their causes and impacts, nuclear power plant failure data acquisition and processing, interdependent failures, and human factor reliability in nuclear power engineering. (P.A.). 8 figs., 7 tabs., 23 refs

  2. LWR nuclear power plant component failures

    International Nuclear Information System (INIS)

    Schmidt, W.H.

    1980-10-01

    An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

  3. Generic nuclear power plant component failure data bank

    International Nuclear Information System (INIS)

    Araujo Goes, A.G. de; Gibelli, S.M.O.

    1988-11-01

    This report consist in the development of a generic nuclear power plant component failure data bank. This data bank was implemented in a PC-XT microcomputer, IBM compatible, using the Open Access II program. Generic failure data tables for Westinghouse nuclear power plants and for general PWR power plants are presented. They are the final product of a research which included a preselection and a selection of data collected from the available sources in the library of CNEN (National Nuclear Energy Commission) and from the CIN/CNEN (Neclear Information Center). Futhermore, a proposal of evaluating models of average failure rates of pumps and valves are also presented. Through the electronic data bank one can easily have a generic view of failure rate ranges as well as failure models foe a certain component. It is very importante to develop procedures to collect and store generic failure data that can be quickly accessed, in order to update the Probabilistic Safety Study of Angra-1 and to used in studies which may have component failures of nuclear power plant safety systems. In the future, data specialization can be achieved by means of statistical calculations involving specific data collected from the operational experience of Angra-1 nuclear power plant and the generic data bank. (author) [pt

  4. Estimation of component failure rates for PSA on nuclear power plants 1982-1997

    International Nuclear Information System (INIS)

    Kirimoto, Yukihiro; Matsuzaki, Akihiro; Sasaki, Atsushi

    2001-01-01

    Probabilistic safety assessment (PSA) on nuclear power plants has been studied for many years by the Japanese industry. The PSA methodology has been improved so that PSAs for all commercial LWRs were performed and used to examine for accident management.On the other hand, most data of component failure rates in these PSAs were acquired from U.S. databases. Nuclear Information Center (NIC) of Central Research Institute of Electric Power Industry (CRIEPI) serves utilities by providing safety- , and reliability-related information on operation and maintenance of the nuclear power plants, and by evaluating the plant performance and incident trends. So, NIC started a research study on estimating the major component failure rates at the request of the utilities in 1988. As a result, we estimated the hourly-failure rates of 47 component types and the demand-failure rates of 15 component types. The set of domestic component reliability data from 1982 to 1991 for 34 LWRs has been evaluated by a group of PSA experts in Japan at the Nuclear Safety Research Association (NSRA) in 1995 and 1996, and the evaluation report was issued in March 1997. This document describes the revised component failure rate calculated by our re-estimation on 49 Japanese LWRs from 1982 to 1997. (author)

  5. Detection of instrument or component failures in a nuclear plant by Luenberger observers

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Colley, R.W.; Alexandro, F.J.; Clark, R.N.

    1985-01-01

    A diagnostic system, which will distinguish between instrument failures (flowmeters, etc.) and component failures (valves, filters, etc.) that show the same symptoms, has been developed for nuclear Plants using Luenberger observers. Luenberger observers are online computer based modules constructed following the technology of Clark [3]. A seventh order model of an FFTF subsystem was constructed using the Advanced Continuous Simulation Language (ACSL) and was used to show through simulation that Luenberger observers can be applied to nuclear systems

  6. Generic component failure data base

    International Nuclear Information System (INIS)

    Eide, S.A.; Calley, M.B.

    1992-01-01

    This report discusses comprehensive component generic failure data base which has been developed for light water reactor probabilistic risk assessments. The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) was used to generate component failure rates. Using this approach, most of the failure rates are based on actual plant data rather then existing estimates

  7. Failure modes of safety-related components at fires on nuclear power plants

    International Nuclear Information System (INIS)

    Aaslund, A.

    2000-03-01

    Probabilistic assessment methods can be used to identify specific plant vulnerabilities. Application of such methods can also facilitate selection among system design alternatives available for safety enhancements. The quality of assessment results is however strongly dependent on realistic and accurate input data for modelling of system component behaviour and failure modes during conditions to be assessed. Use of conservative input data may not lead to results providing guidance on safety upgrades. Adequate input data for probabilistic assessments seems to be lacking for at least failure modes of some electrical components when exposed to a fire. This report presents an attempt to improve the situation with respect to such input data. In order to take advantage of information in existing documentation of fire incident occurrences some of the lessons learned from the fire at Browns Ferry Nuclear Power Plant on March 22, 1975 are discussed in this report. Also a summary of results from different fire tests of electrical cables presented in a fire risk analysis report is a part of the references. The failure modes used to describe fire-induced damage are 'open circuit' and 'hot short' which seems to be commonly accepted terms within the branch. Definitions of the terms are included in the report. Effects of the failure modes when occurring in some of the channels of the reactor protection system are discussed with respect to the existing design of the reactor protection system at Ringhals 2 nuclear power unit. Experiences from the Browns Ferry fire and results from fire tests of electrical cables indicate that the dominating failure mode for electrical cables is 'open circuit'. An 'open circuit' failure leads to circuit disjunction and loss of continuity. The circuit can no longer transmit its signal or power. When affecting channels of the reactor protection system an 'open circuit' failure can cause extensive inadvertent actions of safety related equipment

  8. Component failure data handbook

    International Nuclear Information System (INIS)

    Gentillon, C.D.

    1991-04-01

    This report presents generic component failure rates that are used in reliability and risk studies of commercial nuclear power plants. The rates are computed using plant-specific data from published probabilistic risk assessments supplemented by selected other sources. Each data source is described. For rates with four or more separate estimates among the sources, plots show the data that are combined. The method for combining data from different sources is presented. The resulting aggregated rates are listed with upper bounds that reflect the variability observed in each rate across the nuclear power plant industry. Thus, the rates are generic. Both per hour and per demand rates are included. They may be used for screening in risk assessments or for forming distributions to be updated with plant-specific data

  9. Evaluation of Component Failure Data of the Operating Nuclear Power Plants in Korea Based on NUREG/CR-6928

    International Nuclear Information System (INIS)

    Jeon, Hojun; Na, Janghwan; Shin, Taeyoung

    2014-01-01

    This paper focuses on ensuring the quality of component failure data. When performing data analysis in PSA, we have customized the component failure data based on Bayesian analysis using plant specific experiences and the generic data of Advanced Light Water Reactor Utility Requirements Document (ALWR URD). However, ALWR URD was established by collecting US nuclear power plant (NPP) practices from mid 1980s to early 1990s. We analyzed the component failure data using the raw data of component failures in Pressurized Water Reactor (PWR) plants by 2012. This paper presents the results from analyzing the component failure data based on the new generic data and the latest specific failure data. We also compare the new component failure data to the existing data of PSA models, and evaluate the risk impacts by applying the new data to the PSA models of reference NPPs in this paper. To apply the new generic data source to PSA models, we reviewed and compared NUREG/CR-6928 and the existing generic data source, ALWR URD. In addition, we analyzed the component failure data generated from 16 PWR plants by the end of 2012, and performed the Bayesian update with these raw data based on the new generic data source of NUREG/CR-6928. Also, we reviewed the PSA models of the reference NPP, and identified some important components to CDF. The failure data of the major components decreased in general by applying the new generic data and the latest plant specific data. As a result, the CDF of the reference NPP decreased over 30% compared to the value of the existing CDF

  10. 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)

  11. 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...

  12. 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)

  13. 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

  14. 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)

  15. 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.)

  16. Comparison of Tritium Component Failure Rate Data

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2004-01-01

    Published failure rate values from the US Tritium Systems Test Assembly, the Japanese Tritium Process Laboratory, the German Tritium Laboratory Karlsruhe, and the Joint European Torus Active Gas Handling System have been compared. This comparison is on a limited set of components, but there is a good variety of data sets in the comparison. The data compared reasonably well. The most reasonable failure rate values are recommended for use on next generation tritium handling system components, such as those in the tritium plant systems for the International Thermonuclear Experimental Reactor and the tritium fuel systems of inertial fusion facilities, such as the US National Ignition Facility. These data and the comparison results are also shared with the International Energy Agency cooperative task on fusion component failure rate data

  17. 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

  18. 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)

  19. 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

  20. Assessment of missiles generated by pressure component failure and its application to recent gas-cooled nuclear plant design

    International Nuclear Information System (INIS)

    Tulacz, J.; Smith, R.E.

    1980-01-01

    Methods for establishing characteristics of missiles following pressure barrier rupture have been reviewed in order to enable evaluation of structural response to missile impact and to aid the design of barriers to protect essential plant on gas cooled nuclear plant against unacceptable damage from missile impact. Methods for determining structural response of concrete barriers to missile impact have been reviewed and some methods used for assessing the adequacy of steel barriers on gas-cooled nuclear plant have been described. The possibility of making an incredibility case for some of the worst missiles based on probability arguments is briefly discussed. It is shown that there may be scope for such arguments but there are difficulties in quantifying some of the probability factors. (U.K.)

  1. 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)

  2. IPRDS: component histories and nuclear plant aging

    International Nuclear Information System (INIS)

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

    1984-01-01

    A comprehensive assessment of nuclear power plant component operating histories, maintenance histories, and design and fabrication details is essential to understanding aging phenomena. As part of the In-Plant Reliability Data System (IPRDS), an attempt is being made to collect and analyze such information from a sampling of US nuclear power plants. Utilizing the IPRDS, one can reconstruct the failure history of the components and gain new insight into the causes and modes of failures resulting from normal or premature aging. This information assembled from the IPRDS can be combined with operating histories and postservice component inspection results for cradle-to-grave assessments of component aging under operating conditions. A comprehensive aging assessment can then be used to provide guidelines for improving the detection, monitoring, and mitigation of aging-related failures

  3. Distributions of component failure rates estimated from LER data

    International Nuclear Information System (INIS)

    Atwood, C.L.

    1985-01-01

    Past analyses of Licensee Event Report (LER) data have noted that component failure rates vary from plant to plant, and have estimated the distributions by two-parameter gamma distributions. In this study, a more complicated distributional form is considered, a mixture of gammas. This could arise if the plants' failure rates cluster into distinct groups. The method was applied to selected published LER data for diesel generators, pumps, valves, and instrumentation and control assemblies. The improved fits from using a mixture rather than a single gamma distribution were minimal, and not statistically significant. There seem to be two possibilities: either explanatory variables affect the failure rates only in a gradual way, not a qualitative way; or, for estimating individual component failure rates, the published LER data have been analyzed to the limit of resolution. 9 refs

  4. Distributions of component failure rates, estimated from LER data

    International Nuclear Information System (INIS)

    Atwood, C.L.

    1985-01-01

    Past analyses of Licensee Event Report (LER) data have noted that component failure rates vary from plant to plant, and have estimated the distributions by two-parameter γ distributions. In this study, a more complicated distributional form is considered, a mixture of γs. This could arise if the plants' failure rates cluster into distinct groups. The method was applied to selected published LER data for diesel generators, pumps, valves, and instrumentation and control assemblies. The improved fits from using a mixture rather than a single γ distribution were minimal, and not statistically significant. There seem to be two possibilities: either explanatory variables affect the failure rates only in a gradual way, not a qualitative way; or, for estimating individual component failure rates, the published LER data have been analyzed to the limit of resolution

  5. 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.)

  6. 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.

  7. Analysis of failed nuclear plant components

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1993-01-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power-generating stations since 1974. The considerations involved in working with an analyzing radioactive components 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 service. 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

  8. Analysis of failed nuclear plant components

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1992-07-01

    Argonne National Laboratory has conducted analyses of failed components from nuclear power generating stations since 1974. The considerations involved in working with and analyzing radioactive components 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 service. The failures discussed are (a) intergranular stress corrosion cracking of core spray injection piping in a boiling water reactor, (b) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressure water reactor, (c) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (d) failure of pump seal wear rings by nickel leaching in a boiling water reactor

  9. Investigation of valve failure problems in LWR power plants

    International Nuclear Information System (INIS)

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

  10. 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

  11. IPRDS - Component histories and nuclear plant aging

    International Nuclear Information System (INIS)

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

    1984-01-01

    A comprehensive assessment of nuclear power plant component operating histories, maintenance histories, and design and fabrication details is essential to understanding aging phenomena. As part of the In-Plant Reliability Data System (IPRDS), an attempt is being made to collect and analyze such information from a sampling of U.S. nuclear power plants. Utilizing the IPRDS, one can reconstruct the failure history of the components and gain new insight into the causes and modes of failures resulting from normal or premature aging. This information assembled from the IPRDS can be combined with operating histories and postservice component inspection results for ''cradle-to-grave'' assessments of component aging under operating conditions. A comprehensive aging assessment can then be used to provide guidelines for improving the detection, monitoring, and mitigation of aging-related failures. The examples chosen for this paper illustrate two aging-related areas: the effects of an improved preventive maintenance policy in mitigating aging of a feedwater pump and the identification of reoccuring failures in parts of diesel generators

  12. 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

  13. 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 .

  14. 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

  15. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  16. 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.

  17. 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)

  18. Failure cause and failure rate evaluation on pumps of BWR plants in PSA. Hypothesis testing for typical or plant specific failure rate of pumps

    International Nuclear Information System (INIS)

    Sanada, Takahiro; Nakamura, Makoto

    2009-01-01

    In support of domestic nuclear industry effort to gather and analyze failure data of components concerning nuclear power plants, Nuclear Information Archives (NUCIA) are published for useful information to help PSA. This report focuses on NUCIA pertaining to pumps in domestic nuclear power plants, and provides the reliable estimation on failure rate of pumps resulting from failure cause analysis and hypothesis testing of classified and plant specific failure rate of pumps for improving quality in PSA. The classified and plant specific failure rate of pumps are estimated by analyzing individual domestic nuclear power plant's data of 26 Boiling Water Reactors (BWRs) concerning functionally structurally classified pump failures reported from beginning of commercial operation to March 31, 2007. (author)

  19. Failure trend analysis for safety related components of Korean standard NPPs

    International Nuclear Information System (INIS)

    Choi, Sun Yeong; Han, Sang Hoon

    2005-01-01

    The component reliability data of Korean NPP that reflects the plant specific characteristics is required necessarily for PSA of Korean nuclear power plants. We have performed a project to develop the component reliability database (KIND, Korea Integrated Nuclear Reliability Database) and S/W for database management and component reliability analysis. Based on the system, we have collected the component operation data and failure/repair data during from plant operation date to 2002 for YGN 3, 4 and UCN 3, 4 plants. Recently, we provided the component failure rate data for UCN 3, 4 standard PSA model from the KIND. We evaluated the components that have high-ranking failure rates with the component reliability data from plant operation date to 1998 and 2000 for YGN 3,4 and UCN 3, 4 respectively. We also identified their failure mode that occurred frequently. In this study, we analyze the component failure trend and perform site comparison based on the generic data by using the component reliability data which is extended to 2002 for UCN 3, 4 and YGN 3, 4 respectively. We focus on the major safety related rotating components such as pump, EDG etc

  20. 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

  1. Component failures that lead to manual shutdowns

    International Nuclear Information System (INIS)

    1979-01-01

    The data for this report are taken from a population of thirty-five LWRs, al of which differ appreciably in size, design, and age. Appendix A provides a graphical display of the number of manual shutdowns per operating year as a function of plant age, with the frequency adjusted to reflect plant availability

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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.)

  8. 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)

  9. 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

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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.)

  12. 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.)

  13. 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)

  14. Effect of Component Failures on Economics of Distributed Photovoltaic Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lubin, Barry T. [Univ. of Hartford, West Hartford, CT (United States)

    2012-02-02

    This report describes an applied research program to assess the realistic costs of grid connected photovoltaic (PV) installations. A Board of Advisors was assembled that included management from the regional electric power utilities, as well as other participants from companies that work in the electric power industry. Although the program started with the intention of addressing effective load carrying capacity (ELCC) for utility-owned photovoltaic installations, results from the literature study and recommendations from the Board of Advisors led investigators to the conclusion that obtaining effective data for this analysis would be difficult, if not impossible. The effort was then re-focused on assessing the realistic costs and economic valuations of grid-connected PV installations. The 17 kW PV installation on the University of Hartford's Lincoln Theater was used as one source of actual data. The change in objective required a more technically oriented group. The re-organized working group (changes made due to the need for more technically oriented participants) made site visits to medium-sized PV installations in Connecticut with the objective of developing sources of operating histories. An extensive literature review helped to focus efforts in several technical and economic subjects. The objective of determining the consequences of component failures on both generation and economic returns required three analyses. The first was a Monte-Carlo-based simulation model for failure occurrences and the resulting downtime. Published failure data, though limited, was used to verify the results. A second model was developed to predict the reduction in or loss of electrical generation related to the downtime due to these failures. Finally, a comprehensive economic analysis, including these failures, was developed to determine realistic net present values of installed PV arrays. Two types of societal benefits were explored, with quantitative valuations developed

  15. 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

  16. 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

  17. 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

  18. 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.

  19. Intelligent Component Monitoring for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Tsoukalas, Lefteri

    2010-01-01

    Reliability and economy are two major concerns for a nuclear power generation system. Next generation nuclear power reactors are being developed to be more reliable and economic. An effective and efficient surveillance system can generously contribute toward this goal. Recent progress in computer systems and computational tools has made it necessary and possible to upgrade current surveillance/monitoring strategy for better performance. For example, intelligent computing techniques can be applied to develop algorithm that help people better understand the information collected from sensors and thus reduce human error to a new low level. Incidents incurred from human error in nuclear industry are not rare and have been proven costly. The goal of this project is to develop and test an intelligent prognostics methodology for predicting aging effects impacting long-term performance of nuclear components and systems. The approach is particularly suitable for predicting the performance of nuclear reactor systems which have low failure probabilities (e.g., less than 10 -6 year -1 ). Such components and systems are often perceived as peripheral to the reactor and are left somewhat unattended. That is, even when inspected, if they are not perceived to be causing some immediate problem, they may not be paid due attention. Attention to such systems normally involves long term monitoring and possibly reasoning with multiple features and evidence, requirements that are not best suited for humans.

  20. 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.

  1. 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...

  2. 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)

  3. 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

  4. 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.)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. Diagnostic technology of PWR plant equipment failures

    International Nuclear Information System (INIS)

    Nakamura, Tetsuo; Tanaka, Mamoru; Okamachi, Masao; Taguchi, Shozo; Nagashima, Kazuhiro; Ishikawa, Satoshi

    1985-01-01

    To confirm the soundness of the important facilities in a nuclear power plant contributes to the reliability of the plant operations and improvement of its operation rate. For this purpose, the following diagnostic techniques have been developed. (1). Vibration and loose parts monitoring: Detection of abnormal structural vibrations in the reactor, estimation of its mode, detection of loose parts in the primary system, and estimation of the position and energy of their collisions against the reactor vessel or the like. (2). Valve leak monitoring: Detection of leaks from primary valves in the primary cooling boundary, such as the pressurizer relief valve and safety valve, and estimation of the form of the leaks. (3). Detector noise response diagnosis: Diagnosis of degradation of principal process detectors during plant operation. Furthermore, a diagnostic system incorporating the above diagnostic technology applicable to actual plants has been experimentally manufactured and successfully verified. (author)

  10. 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.

  11. 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.)

  12. Polyphophoinositides components of plant nuclear membranes

    International Nuclear Information System (INIS)

    Hendrix, K.W.; Boss, W.F.

    1987-01-01

    The polyphosphoinositides, phosphatidylinositol monophosphate (PIP) and phosphatidylinositol bisphosphate (PIP 2 ), have been shown to be important components in signal transduction in many animal cells. Recently, these lipids have been found to be associated with plasma membrane but not microsomal membrane isolated from fusogenic wild carrot cells; however, in that study the lipids of the nuclear membrane were not analyzed. Since polyphosphoinositides had been shown to be associated with the nuclear membranes as well as the plasma membrane in some animal cells, it was important to determine whether they were associated with plant nuclear membranes as well. Cells were labeled for 18h with [ 3 H] inositol and the nuclei were isolated by a modification of the procedure of Saxena et al. Preliminary lipid analyses indicate lower amount of PIP and PIP 2 in nuclear membranes compared to whole protoplasts. This suggests that the nuclear membranes of carrot cells are not enriched in PIP and PIP 2 ; however, the Triton X-100 used during the nuclear isolation procedure may have affected the recovery of the lipids. Experiments are in progress to determine the effects of Triton X-100 on lipid extraction

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. Evaluation of premature failure of a gas turbine component

    CSIR Research Space (South Africa)

    Dedekind, MO

    1996-01-03

    Full Text Available A case study of certain gas turbine stator vanes which fail prematurely is presented, with a view to determining whether operational procedure might have caused the failures. The engines had been operated from a ‘hot-and-high’ environment...

  1. Lessons learned from fatique failures in major FWR components

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    This paper evaluates the field fatigue failure experience and describes the lessons learned that can be employed in managing fatigue damage at the sites of these failures and at other susceptible sites. Fatigue damage has resulted in cracks on the inside surfaces of vessels and piping, and in some cases, through-wall cracks resulting in coolant leakage. All of the fatigue failures resulted from conditions or stressors that were not accounted for in the original design analyses. In some cases, it has proven difficult to discover fatigue cracks using conventional inservice inspection methods; several cracks were detected because of leakage. Supplementary monitoring and inspection techniques such as fatigue monitoring, acoustic emission monitoring, and time-of-flight-diffraction ultrasonic testing can be used to assist in identifying susceptible sites, estimating crack growth, and sizing existing fatigue cracks. It is important to identify the root cause of failures because once the stressors and degradation mechanisms are known, changes in operating procedures and designs can be implemented to mitigate future fatigue damage

  2. Component failure experiments of the seal casing for the multiple-row bolted connection of the containment equipment hatch of the nuclear power plant Philippsburg II; Experimente zum Versagen des Dichtkastens fuer die Materialschleusenverschraubung im Containment des KKW Philippsburg II

    Energy Technology Data Exchange (ETDEWEB)

    Messemer, G.

    2008-06-15

    For the weakness analysis of the containment under internal pressure a question that arose was at what pressure the seal casing for the equipment hatch will fail. For this investigation model strips of the seal casing cross section were manufactured as tensile specimens. Variations of the seal plates welding seams were considered and tested. The type of failure and the associated deformations and resulting forces were measured for 6 specimens. (orig.)

  3. 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

  4. 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)

  5. Reliability of Wind Turbine Components-Solder Elements Fatigue Failure

    DEFF Research Database (Denmark)

    Kostandyan, Erik; Sørensen, John Dalsgaard

    2012-01-01

    on the temperature mean and temperature range. Constant terms and model errors are estimated. The proposed methods are useful to predict damage values for solder joint in power electrical components. Based on the proposed methods it is described how to find the damage level for a given temperature loading profile....... The proposed methods are discussed for application in reliability assessment of Wind Turbine’s electrical components considering physical, model and measurement uncertainties. For further research it is proposed to evaluate damage criteria for electrical components due to the operational temperature...

  6. 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)

  7. 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

  8. 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.

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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

  15. Undesired Plant-Derived Components in Food

    NARCIS (Netherlands)

    Dusemund, Birgit; Rietjens, Ivonne M.C.M.; Abraham, Klaus; Cartus, Alexander; Schrenk, Dieter

    2017-01-01

    Among the various chemical compounds, the class of natural plant-derived substances in the modern food chain is generating increasing concern. Adverse effects encountered may be various and pose risks of acute, subchronic, or chronic toxicity. The underlying mechanisms of toxicity may be

  16. 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.

  17. 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)

  18. 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...

  19. Structural mechanics of nuclear plant components

    International Nuclear Information System (INIS)

    Roche, R.

    1986-10-01

    Sound structural analysis are needed for designing safe and reliable components, hence his play is very important in nuclear industry. This report is a provisional writing on the good practice in structural mechanics. Emphasis is put on non elastic analysis, damage appraisal, fatigue, fracture mechanics and also on elevated temperature behaviour [fr

  20. CONSIDERATIONS OVER A BIOGAS PLANT COMPONENTS

    Directory of Open Access Journals (Sweden)

    Mariana DUMITRU

    2014-04-01

    Full Text Available This paper starts from the conviction that one of the main environmental problems of today’s society is the continuously increasing production of organic wastes. In many countries, sustainable waste management have become major political priorities in order to reduce pollution and greenhouse gas emissions and to avoid, as much as possible, global climate changes. This problem becomes more and more present in our country too. Production of biogas through anaerobic digestion of animal manure and slurries as well as of a wide range of digestible organic wastes, converts these substrates into renewable energy and offers a natural fertiliser for agriculture. That is why we consider that biogas plants will be more and more used in the future. In this paper we show the different stages which must be operated in a biogas plant and the problems which can be met in each of them.

  1. 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

  2. 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

  3. 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

  4. 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

  5. Monitoring ageing of components in nuclear plants

    International Nuclear Information System (INIS)

    Fritz, M.R.

    1992-01-01

    There are several mechanisms of ageing or damage in nuclear components, of which the best known can be classified into three categories: generalized damage mechanisms (wear, corrosion, erosion,...), local damage phenomena (fatigue, corrosion,...) and material property degradations. For ageing evaluation, the first requirement is a good understanding of the damage mechanisms and the determination of the kinetic laws and major influencing factors. When these factors are measurable physical parameters, ageing monitoring and periodic evaluation of damage level become possible. From the set of tools available for ageing evaluation, four are presented here in more detail: the transient logging procedure, the defect injuriousness analysis, the fatigue meter, the probabilistic approach to structural integrity. (author)

  6. Monitoring ageing of components in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, M R [FRAMATOME, Paris (France)

    1992-07-01

    There are several mechanisms of ageing or damage in nuclear components, of which the best known can be classified into three categories: generalized damage mechanisms (wear, corrosion, erosion,...), local damage phenomena (fatigue, corrosion,...) and material property degradations. For ageing evaluation, the first requirement is a good understanding of the damage mechanisms and the determination of the kinetic laws and major influencing factors. When these factors are measurable physical parameters, ageing monitoring and periodic evaluation of damage level become possible. From the set of tools available for ageing evaluation, four are presented here in more detail: the transient logging procedure, the defect injuriousness analysis, the fatigue meter, the probabilistic approach to structural integrity. (author)

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. Determination of the remaining operational life of power plant components

    International Nuclear Information System (INIS)

    Eiden, H.; Vorwerk, K.; Graeff, D.; Hoff, E.

    1983-01-01

    The proceedings volume presents, in full wording, eight papers read at a TUEV Rheinland meeting in Johannesburg, South Africa, in August 1982. Subjects: Layout, quality assurance, service life analysis etc. of power plant components. (RW) [de

  13. 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

  14. 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.)

  15. 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)

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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...

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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...

  13. 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

  14. 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.)

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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)

  20. 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)

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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).

  6. 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)

  7. 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)

  8. 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

  9. 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)

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. Some current engineering topics in nuclear power plant components

    International Nuclear Information System (INIS)

    Amana, M.

    1977-01-01

    An analysis based on the principle of fracture mechanics, is presented for several engineering problems occuring in nuclear power plant components. The specific problems covered are: underclad cracking; stress corrosion cracking; cracks in HAZ of nozzle weld; feedwater nozzle corner crack; shift of transition temperature due to neutron irradiation; LWR local-ECC thermal shock experiment; and design and material selection of RPV in terms of fracture mechanics. (B.R.H.)

  15. Nuclear Power Plant Mechanical Component Flooding Fragility Experiments Status

    Energy Technology Data Exchange (ETDEWEB)

    Pope, C. L. [Idaho State Univ., Pocatello, ID (United States); Savage, B. [Idaho State Univ., Pocatello, ID (United States); Johnson, B. [Idaho State Univ., Pocatello, ID (United States); Muchmore, C. [Idaho State Univ., Pocatello, ID (United States); Nichols, L. [Idaho State Univ., Pocatello, ID (United States); Roberts, G. [Idaho State Univ., Pocatello, ID (United States); Ryan, E. [Idaho State Univ., Pocatello, ID (United States); Suresh, S. [Idaho State Univ., Pocatello, ID (United States); Tahhan, A. [Idaho State Univ., Pocatello, ID (United States); Tuladhar, R. [Idaho State Univ., Pocatello, ID (United States); Wells, A. [Idaho State Univ., Pocatello, ID (United States); Smith, C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-24

    This report describes progress on Nuclear Power Plant mechanical component flooding fragility experiments and supporting research. The progress includes execution of full scale fragility experiments using hollow-core doors, design of improvements to the Portal Evaluation Tank, equipment procurement and initial installation of PET improvements, designation of experiments exploiting the improved PET capabilities, fragility mathematical model development, Smoothed Particle Hydrodynamic simulations, wave impact simulation device research, and pipe rupture mechanics research.

  16. Advanced targeted monitoring of high temperature components in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E.; Maile, K.; Jovanovic, A. [MPA Stuttgart (Germany)

    1998-12-31

    The article presents the idea of targeted monitoring of high-temperature pressurized components in fossil-fueled power plants, implemented within a modular software system and using, in addition to pressure and temperature data, also displacement and strain measurement data. The concept has been implemented as a part of a more complex company-oriented Internet/Intranet system of MPA Stuttgart (ALIAS). ALIAS enables to combine smoothly the monitoring results with those of the off-line analysis, e. g. sensitivity analyses, comparison with preceding experience (case studies), literature search, search in material databases -(experimental and standard data), nonlinear FE-analysis, etc. The concept and the system have been implemented in real plant conditions several power plants in Germany and Europe: one of these applications and its results are described more in detail in the presentation. (orig.) 9 refs.

  17. Advanced targeted monitoring of high temperature components in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E; Maile, K; Jovanovic, A [MPA Stuttgart (Germany)

    1999-12-31

    The article presents the idea of targeted monitoring of high-temperature pressurized components in fossil-fueled power plants, implemented within a modular software system and using, in addition to pressure and temperature data, also displacement and strain measurement data. The concept has been implemented as a part of a more complex company-oriented Internet/Intranet system of MPA Stuttgart (ALIAS). ALIAS enables to combine smoothly the monitoring results with those of the off-line analysis, e. g. sensitivity analyses, comparison with preceding experience (case studies), literature search, search in material databases -(experimental and standard data), nonlinear FE-analysis, etc. The concept and the system have been implemented in real plant conditions several power plants in Germany and Europe: one of these applications and its results are described more in detail in the presentation. (orig.) 9 refs.

  18. 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

  19. 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

  20. 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.)

  1. 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.)

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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.)

  10. 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.

  11. 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

  12. IR-360 nuclear power plant safety functions and component classification

    Energy Technology Data Exchange (ETDEWEB)

    Yousefpour, F., E-mail: fyousefpour@snira.co [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of); Shokri, F.; Soltani, H. [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of)

    2010-10-15

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  13. IR-360 nuclear power plant safety functions and component classification

    International Nuclear Information System (INIS)

    Yousefpour, F.; Shokri, F.; Soltani, H.

    2010-01-01

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  14. Countermeasure technologies against materials deterioration of nuclear power plant components

    International Nuclear Information System (INIS)

    2004-09-01

    This report was tentative safety standard on countermeasure technologies against materials deterioration of nuclear power plant components issued in 2004 on the base of the testing data obtained until March 2004, which was to be applied for technical evaluation for lifetime management of aged plants and preventive maintenance or repair of neutron irradiated components such as core shrouds and jet pumps. In order to prevent stress corrosion cracks (SCCs) of austenitic stainless steel welds of reactor components, thermal surface modification using laser beams was used on neutron irradiated materials with laser cladding or surface melting process methods by limiting heat input according to amount of accumulated helium so as to prevent crack initiation caused by helium bubble growth and coalescence. Laser cladding method of laser welding using molten sleeve set inside pipe surface to prevent SCCs of nickel-chromium-iron alloy welds, alloy 690 cladding method using tungsten inert gas (TIG) welding to prevent SCCs of nickel-chromium-iron alloy welds for dissimilar joints of pipes, and laser surface solid solution heat treatment method of laser irradiation on surfaces to prevent SCCs of austenitic stainless steel welds were also included as repair technologies. (T. Tanaka)

  15. 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

  16. EPRI research on component aging and nuclear plant life extension

    International Nuclear Information System (INIS)

    Sliter, G.E.; Carey, J.J.

    1985-01-01

    This paper first describes several research efforts sponsored by the Electric Power Research Institute (EPRI) that examine the aging degradation of organic materials and the nuclear plant equipment in which they appear. This research includes a compendium of material properties characterizing the effects of thermal and radiation aging, shake table testing to evaluate the effects of aging on the seismic performance of electrical components, and a review of condition monitoring techniques applicable to electrical equipment. Also described is a long-term investigation of natural versus artificial aging using reactor buildings as test beds. The paper then describes how the equipment aging research fits into a broad-scoped EPRI program on nuclear plant life extension. The objective of this program is to provide required information, technology, and guidelines to enable utilities to significantly extend operating life beyond the current 40-year licensed term

  17. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Choi, Jae Boong [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2002-03-15

    This project focuses on developing reliable life evaluation technology for nuclear power plant components, and is divided into two parts, development of a life evaluation system for nuclear pressure vessels 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 in this project: defect assessment method for steam generator tubes, development of fatigue monitoring system, assessment of corroded pipes, domestic round robin analysis for constructing P-T limit curve for RPV, development of probabilistic integrity assessment technique, effect of aging on strength of dissimilar welds, applicability of LBB to cast stainless steel, and development of probabilistic piping fracture mechanics.

  18. 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)

  19. Project of mechanical components for nuclear power plants

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do; Farias Brito David, D. de

    1984-01-01

    The equipment foreseen to be part of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design of the components. The design and calculation's concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities are described. (Author) [pt

  20. Design and structural calculation of nuclear power plant mechanical components

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do

    1986-01-01

    The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt

  1. 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)

  2. 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

  3. 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)

  4. FAILPROB-A Computer Program to Compute the Probability of Failure of a Brittle Component; TOPICAL

    International Nuclear Information System (INIS)

    WELLMAN, GERALD W.

    2002-01-01

    FAILPROB is a computer program that applies the Weibull statistics characteristic of brittle failure of a material along with the stress field resulting from a finite element analysis to determine the probability of failure of a component. FAILPROB uses the statistical techniques for fast fracture prediction (but not the coding) from the N.A.S.A. - CARES/life ceramic reliability package. FAILPROB provides the analyst at Sandia with a more convenient tool than CARES/life because it is designed to behave in the tradition of structural analysis post-processing software such as ALGEBRA, in which the standard finite element database format EXODUS II is both read and written. This maintains compatibility with the entire SEACAS suite of post-processing software. A new technique to deal with the high local stresses computed for structures with singularities such as glass-to-metal seals and ceramic-to-metal braze joints is proposed and implemented. This technique provides failure probability computation that is insensitive to the finite element mesh employed in the underlying stress analysis. Included in this report are a brief discussion of the computational algorithms employed, user instructions, and example problems that both demonstrate the operation of FAILPROB and provide a starting point for verification and validation

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. 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.

  12. Probabilistic approaches to life prediction of nuclear plant structural components

    International Nuclear Information System (INIS)

    Villain, B.; Pitner, P.; Procaccia, H.

    1996-01-01

    In the last decade there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to aging mechanisms, mainly on major passive structural components such as steam generators, pressure vessels and piping in nuclear plants. Because there can be numerous uncertainties involved in a assessment of the performance of these structural components, probabilistic methods. The benefits of a probabilistic approach are the clear treatment of uncertainly and the possibility to perform sensitivity studies from which it is possible to identify and quantify the effect of key factors and mitigative actions. They thus provide information to support effective decisions to optimize In-Service Inspection planning and maintenance strategies and for realistic lifetime prediction or reassessment. The purpose of the paper is to discuss and illustrate the methods available at EDF for probabilistic component life prediction. This includes a presentation of software tools in classical, Bayesian and structural reliability, and an application on two case studies (steam generator tube bundle, reactor pressure vessel). (authors)

  13. Probabilistic approaches to life prediction of nuclear plant structural components

    International Nuclear Information System (INIS)

    Villain, B.; Pitner, P.; Procaccia, H.

    1996-01-01

    In the last decade there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to aging mechanisms, mainly on major passive structural components such as steam generators, pressure vessels and piping in nuclear plants. Because there can be numerous uncertainties involved in an assessment of the performance of these structural components, probabilistic methods provide an attractive alternative or supplement to more conventional deterministic methods. The benefits of a probabilistic approach are the clear treatment of uncertainty and the possibility to perform sensitivity studies from which it is possible to identify and quantify the effect of key factors and mitigative actions. They thus provide information to support effective decisions to optimize In-Service Inspection planning and maintenance strategies and for realistic lifetime prediction or reassessment. The purpose of the paper is to discuss and illustrate the methods available at EDF for probabilistic component life prediction. This includes a presentation of software tools in classical, Bayesian and structural reliability, and an application on two case studies (steam generator tube bundle, reactor pressure vessel)

  14. Safety aspects of nuclear power plant component aging

    International Nuclear Information System (INIS)

    Conte, M.; Deletre, G.; Henry, J.Y.

    1988-01-01

    The safety of nuclear plants depends on the capacity of the systems they are composed to perform the functions they were designed for. The identification and understanding of phenomena liable to degrade this operational capacity thus constitute one of the safety problems for which allowance must be made at the earliest stage of a project. Aging, a natural and hence unavoidable process affecting all the components of an installation, was identified at a very early stage as being one of these phenomena. The investigation and implementation of solutions to the safety problems associated to aging make it necessary to: defining the domain in which the consequences of aging are to be evaluated, identifying the parameters involved, identifying the components sensitive to these parameters, understanding the mechanisms which govern its evolution. The results of qualification tests, and of tests and checks carried out at different stages of construction and operation, as well as allowance for operating experience, constitute the necessary basis for establishing or improving the regulatory requirements. The procedures for validating components and systems of the installation are also drawn up on the basis of these tests. Finally, the actions initiated within the scope of research and development programmes supply the additional data necessary for such validation, and provide the indispensable support for knowledge improvement

  15. 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

  16. 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)

  17. 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)

  18. Single-failure-proof cranes for nuclear power plants

    International Nuclear Information System (INIS)

    Porse, L.

    1979-05-01

    NRC has licensed reactors on the basis that the safe handling of critical loads can be accomplished by adding safety features to the handling equipment, by adding special features to the structures and areas over which the critical load is carried, or by a combination of the two. When reliance for the safe handling of critical loads is placed on the crane system itself, the system should be designed so that a single failure will not result in the loss of the capability of the system to safely retain the load. Features of the design, fabrication, installation, inspection, testing, and operation of single-failure-proof overhead crane handling systems that are used for handling critical loads are identified. These features are limited to the hoisting system and to braking systems for trolley and bridge. Other load-bearing items such as girders should be conservatively designed but need not be considered single failure proof

  19. 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.

  20. 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.

  1. 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.

  2. Failure analysis of boiler tubes in lakhra coal power plant

    International Nuclear Information System (INIS)

    Shah, A.; Baluch, M.M.; Ali, A.

    2010-01-01

    Present work deals with the failure analysis of a boiler tube in Lakhra fluidized bed combustion power station. Initially, visual inspection technique was adopted to analyse the fractured surface. Detailed microstructural investigations of the busted boiler tube were carried out using light optical microscope and scanning electron microscope. The hardness tests were also performed. A 50 percent decrease in hardness of intact portion of the tube material and from area adjacent to failure was measured, which was found to be in good agreement with the wall thicknesses measured of the busted boiler tube i.e. 4 mm and 2 mm from unaffected portion and ruptured area respectively. It was concluded that the major cause of failure of boiler tube is erosion of material which occurs due the coal particles strike at the surface of the tube material. Since the temperature of boiler is not maintained uniformly. The variations in boiler temperature can also affect the material and could be another reason for the failure of the tube. (author)

  3. Actin is an essential component of plant gravitropic signaling pathways

    Science.gov (United States)

    Braun, Markus; Hauslage, Jens; Limbach, Christoph

    2003-08-01

    A role of the actin cytoskeleton in the different phases of gravitropism in higher plant organs seems obvious, but experimental evidence is still inconclusive and contradictory. In gravitropically tip-growing rhizoids and protonemata, however, it is well documented that actin is an essential component of the tip-growth machinery and is involved either in the cellular mechanisms that lead to gravity sensing and in the processes of the graviresponses that result in the reorientation of the growth direction. All these processes depend on a complexly organized and highly dynamic organization of actin filaments whose diverse functions are coordinated by numerous associated proteins. Actin filaments and myosins mediate the transport of secretory vehicles to the growing tip and precisely control the delivery of cell wall material. In addition, both cell types use a very efficient actomyosin-based system to control and correct the position of their statoliths and to direct sedimenting statoliths to confined graviperception sites at the plasma membrane. The studies presented in this paper provide evidence for the essential role of actin in plant gravity sensing and the gravitropic responses. A unique actin-organizing center exists in the tip of characean rhizoids and protonemata which is associated with and dynamically regulated by a specific set of actin-dynamizing proteins. It is concluded that this highly dynamic apical actin array is an essential prerequisite for gravity sensing and gravity-oriented tip growth.

  4. 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...

  5. 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.

  6. Effects of air blast on power plant structures and components

    International Nuclear Information System (INIS)

    Kot, C.A.; Valentin, R.A.; McLennan, D.A.; Turula, P.

    1978-10-01

    The effects of air blast from high explosives detonation on selected power plant structures and components are investigated analytically. Relying on a synthesis of state of the art methods estimates of structural response are obtained. Similarly blast loadings are determined from compilations of experimental data reported in the literature. Plastic-yield line analysis is employed to determine the response of both concrete and steel flat walls (plates) under impulsive loading. Linear elastic theory is used to investigate the spalling of concrete walls and mode analysis methods predict the deflection of piping. The specific problems considered are: the gross deformation of reinforced concrete shield and containment structures due to blast impulse, the spalling of concrete walls, the interaction or impact of concrete debris with steel containments and liners, and the response of exposed piping to blast impulse. It is found that for sufficiently close-in detonations and/or large explosive charge weights severe damage or destruction will result. This is particularly true for structures or components directly exposed to blast impulse

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. Impact of few failure data on the opportunistic replacement policy for multi-component systems

    International Nuclear Information System (INIS)

    Laggoune, Radouane; Chateauneuf, Alaa; Aissani, Djamil

    2010-01-01

    In continuous operating units, the production loss is often very large during the system shut down. Their economic profitability is conditioned by the implementation of suitable maintenance policy that could increase the availability and reduce the operating costs. In this paper, an opportunistic replacement policy is proposed for multi-component series system in the context of data uncertainty, where the expected total cost per unit time is minimized under general lifetime distribution. When the system is down, either correctively or preventively, the opportunity to replace preventively non-failed components is considered. To deal with the problem of the small size of failure data samples, the Bootstrap technique is applied, in order to model the uncertainties in parameter estimates. The Weibull parameters are considered as random variables rather than just deterministic point estimates. A solution procedure based on Monte Carlo simulations with informative search method is proposed and applied to the optimization of preventive maintenance plan for a hydrogen compressor in an oil refinery.

  14. Natural versus artificial aging of nuclear power plant components

    International Nuclear Information System (INIS)

    Shaw, M.T.

    1992-01-01

    This program seeks to understand the aging of polymeric materials, in cables and other components in nuclear reactor containment, by comparing aging processes for a variety of materials under natural conditions with those under the accelerated laboratory conditions used in qualification. The first five-year phase has been completed in what is planned as a long-term study of up to 40 years. Data from the program can be used as a basis of forecasting more realistic lifetimes in reactor service. The program is of critical importance for utilities both for the safe operation of plants and for minimizing the cost of periodic replacements upon expiration of originally predicted qualified life. The first five-year period has involved the selection and acquisition of test specimens, their preparation for placement in the containment, the selection of plants and locations for the specimens, the establishment of methods for monitoring radiation and temperature levels at each site, development of plans for scheduled removals, test method development, and testing of the specimens by physical and mechanical methods. Specimens have been subjected to short-term accelerated aging, as well as to reactor containment aging for up to five years. They consist of many types of polymers in products of several different manufacturers. Environmental conditions cover a wide range of temperature and radiation levels at 17 locations in 9 reactors of participating utilities. Initial results, which include tests of special cases subject to 8 years of reactor aging at Northeast Utilities, indicate several instances of changes having statistical significance in density or tensile properties due to containment service, but none of these changes are large enough to be of any concern. 12 refs., 19 figs., 21 tabs

  15. 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.)

  16. 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)

  17. 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)

  18. 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

  19. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

    International Nuclear Information System (INIS)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E.

    1994-09-01

    This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

  20. 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

  1. Analysis of GMO Plum Plant Culture in System Operations Failure

    Science.gov (United States)

    Mercado, Dianne

    2017-01-01

    GMO plum trees are being evaluated at the Kennedy Space Center as a possible candidate for future space crops. Previously conducted horticultural testing compared the performance of several plum genotypes in controlled environment chambers, resulting in a down-selection to the NASA-11 genotype. Precursory studies determined the water use requirements to sustain the plants as well as the feasibility of grafting non-GMO plum scions onto GMO plum rootstocks of NASA-5, NASA-10, and NASA-11 genotypes. This study follows the growth and horticultural progress of plum trees and in-vitro cultures from August 2017 to November 2017, and provides supplemental support for future GMO plum studies. The presence of Hurricane Irma in early September 2017 resulted in the plants undergoing material deterioration from major changes to their overall horticultural progress.

  2. Schizophrenia as failure of left hemispheric dominance for the phonological component of language.

    Science.gov (United States)

    Angrilli, Alessandro; Spironelli, Chiara; Elbert, Thomas; Crow, Timothy J; Marano, Gianfranco; Stegagno, Luciano

    2009-01-01

    T. J. Crow suggested that the genetic variance associated with the evolution in Homo sapiens of hemispheric dominance for language carries with it the hazard of the symptoms of schizophrenia. Individuals lacking the typical left hemisphere advantage for language, in particular for phonological components, would be at increased risk of the typical symptoms such as auditory hallucinations and delusions. Twelve schizophrenic patients treated with low levels of neuroleptics and twelve matched healthy controls participated in an event-related potential experiment. Subjects matched word-pairs in three tasks: rhyming/phonological, semantic judgment and word recognition. Slow evoked potentials were recorded from 26 scalp electrodes, and a laterality index was computed for anterior and posterior regions during the inter stimulus interval. During phonological processing individuals with schizophrenia failed to achieve the left hemispheric dominance consistently observed in healthy controls. The effect involved anterior (fronto-temporal) brain regions and was specific for the Phonological task; group differences were small or absent when subjects processed the same stimulus material in a Semantic task or during Word Recognition, i.e. during tasks that typically activate more widespread areas in both hemispheres. We show for the first time how the deficit of lateralization in the schizophrenic brain is specific for the phonological component of language. This loss of hemispheric dominance would explain typical symptoms, e.g. when an individual's own thoughts are perceived as an external intruding voice. The change can be interpreted as a consequence of "hemispheric indecision", a failure to segregate phonological engrams in one hemisphere.

  3. Schizophrenia as failure of left hemispheric dominance for the phonological component of language.

    Directory of Open Access Journals (Sweden)

    Alessandro Angrilli

    Full Text Available BACKGROUND: T. J. Crow suggested that the genetic variance associated with the evolution in Homo sapiens of hemispheric dominance for language carries with it the hazard of the symptoms of schizophrenia. Individuals lacking the typical left hemisphere advantage for language, in particular for phonological components, would be at increased risk of the typical symptoms such as auditory hallucinations and delusions. METHODOLOGY/PRINCIPAL FINDINGS: Twelve schizophrenic patients treated with low levels of neuroleptics and twelve matched healthy controls participated in an event-related potential experiment. Subjects matched word-pairs in three tasks: rhyming/phonological, semantic judgment and word recognition. Slow evoked potentials were recorded from 26 scalp electrodes, and a laterality index was computed for anterior and posterior regions during the inter stimulus interval. During phonological processing individuals with schizophrenia failed to achieve the left hemispheric dominance consistently observed in healthy controls. The effect involved anterior (fronto-temporal brain regions and was specific for the Phonological task; group differences were small or absent when subjects processed the same stimulus material in a Semantic task or during Word Recognition, i.e. during tasks that typically activate more widespread areas in both hemispheres. CONCLUSIONS/SIGNIFICANCE: We show for the first time how the deficit of lateralization in the schizophrenic brain is specific for the phonological component of language. This loss of hemispheric dominance would explain typical symptoms, e.g. when an individual's own thoughts are perceived as an external intruding voice. The change can be interpreted as a consequence of "hemispheric indecision", a failure to segregate phonological engrams in one hemisphere.

  4. 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)

  5. Communication failure: basic components, contributing factors, and the call for structure.

    Science.gov (United States)

    Dayton, Elizabeth; Henriksen, Kerm

    2007-01-01

    Communication is a taken-for-granted human activity that is recognized as important once it has failed. Communication failures are a major contributor to adverse events in health care. The components and processes of communication converge in an intricate manner, creating opportunities for misunderstanding along the way. When a patient's safety is at risk, providers should speak up (that is, initiate a message) to draw attention to the situation before harm is caused. They should also clearly explain (encode) and understand (decode) each other's diagnosis and recommendations to ensure well coordinated delivery of care. Beyond basic dyadic communication exchanges, an intricate web of individual, group, and organizational factors--more specifically, cognitive workload, implicit assumptions, authority gradients, diffusion of responsibility, and transitions of care--complicate communication. More structured and explicitly designed forms of communication have been recommended to reduce ambiguity, enhance clarity, and send an unequivocal signal, when needed, that a different action is required. Read-backs, Situation-Background-Assessment-Recommendation, critical assertions, briefings, and debriefings are seeing increasing use in health care. CODA: Although structured forms of communication have good potential to enhance clarity, they are not fail-safe. Providers need to be sensitive to unexpected consequences regarding their use.

  6. On the major ductile fracture methodologies for failure assessment of nuclear reactor components

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Andrade, Arnaldo H.P. de; Landes, John D.

    1996-01-01

    In structures like nuclear reactor components there is a special concern with the loads that may occur under postulated accident conditions. These loads can cause the stresses to go well beyond the linear elastic limits, requiring the use of ductile fracture mechanics methods to the prediction of the structure behavior. Since the use of numerical methods to apply EPFM concepts is expensive and time consuming, the existence of analytical engineering procedures are of great relevance. The lack of precision in detail, as compared with numerical nonlinear analyses, is compensated by the possibility of quick failure assessments. This is a determinant factor in situations where a systematic evaluation of a large range of geometries and loading conditions is necessary, like in thr application of the Leak-Before-Break (LBB) concept on nuclear piping. This paper outlines four ductile fracture analytical methods, pointing out positive and negative aspects of each one. The objective is to take advantage of this critical review to conceive a new methodology, one that would gather strong points of the major existent methods and would try to eliminate some of their drawbacks. (author)

  7. 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

  8. 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...

  9. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  10. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  11. 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.

  12. 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.

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. Operational planning optimization of steam power plants considering equipment failure in petrochemical complex

    International Nuclear Information System (INIS)

    Luo, Xianglong; Zhang, Bingjian; Chen, Ying; Mo, Songping

    2013-01-01

    Highlights: ► We develop a systematic programming methodology to address equipment failure. ► We classify different operation conditions into real periods and virtual periods. ► The formulated MILP models guarantee cost reduction and enough operation safety. ► The consideration of reserving operation redundancy is effective. - Abstract: One or more interconnected steam power plants (SPPs) are constructed in a petrochemical complex to supply utility energy to the process. To avoid large economic penalties or process shutdowns, these SPPs should be flexible and reliable enough to meet the process energy requirement under varying conditions. Unexpected utility equipment failure is inevitable and difficult to be predicted. Most of the conventional methods are based on the assumption that SPPs do not experience any kind of equipment failure. Unfortunately, a process shutdown cannot be avoided when equipment fails unexpectedly. In this paper, a systematic methodology is presented to minimize the total cost under normal conditions while reserving enough flexibility and safety for unexpected equipment failure conditions. The proposed method transforms the different conditions into real periods to indicate normal scenarios and virtual periods to indicate unexpected equipment failure scenarios. The optimization strategy incorporating various operation redundancy scheduling, the transition constraints from equipment failure conditions to normal conditions, and the boiler load increase behavior modeling are presented to save cost and guarantee operation safety. A detailed industrial case study shows that the proposed systematic methodology is effective and practical in coping with equipment failure conditions with only few additional cost penalties

  20. 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)

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

    International Nuclear Information System (INIS)

    Tian Lu

    2012-01-01

    This paper briefly describes the development of digital I and C system in nuclear power plant, and analyses the viewpoints of NRC and other nuclear safety authorities on Software Common Cause Failure (SWCCF). In view of the SWCCF issue introduced by the digitized platform adopted in nuclear power plant safety system, this paper illustrated a diversified defence strategy for computer software and hardware. A diversified defence-in-depth solution is provided for digital safety system of nuclear power plant. Meanwhile, analysis on problems may be faced during application of nuclear safety license are analyzed, and direction of future nuclear safety I and C system development are put forward. (author)

  2. 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.

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. Quality assurance during the manufacture of nuclear power plant components

    International Nuclear Information System (INIS)

    Mueller, J.

    1976-01-01

    Apart from the special requirements of quality assurance in the production of components for the nuclear industry, in particular nuclear power stations, the author discusses special methods of quality control in the testing of welded joints. (TK) [de

  8. 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.)

  9. NUCLEBRAS' installations for tests of nuclear power plants components

    International Nuclear Information System (INIS)

    Vasconcelos Paiva, I.P. de; Horta, J.A.L.; Avelar Esteves, F. de; Pinheiro, R.B.

    1983-05-01

    The reasons for NUCLEBRAS' Nuclear Technology Development Center to implement a laboratory for supporting Brazilian manufacturers, giving to them the means for performing functional tests of industrial products, are presented. A brief description of the facilities under construction: the Components Test Loop and the Facility for Testing N.P.P. Components under Accident Conditions, and of other already in operation, is given, as well as its objectives and main technical characteristics. Some test results already obtained are also presented. (Author) [pt

  10. Nuclebras' installations for performance tests of nuclear power plants components

    International Nuclear Information System (INIS)

    Vasconcelos Paiva, I.P. de; Avelar Esteves, F. de; Horta, J.A.L.; Resende, M.F.R.; Pinheiro, R.B.

    1984-01-01

    The reasons for Nuclebras' Nuclear Technology Development Center to implement a laboratory for supporting Brazilian manufactures, giving to them the means for performing functional tests of industrial products, are presented. A brief description of facilities under construction: the components Test Loop and Facility for Testing N.P.P. components under Accident conditions, and other already in operation, as well as its objectives and main technical characteristics. Some test results had already obtained are also presented. (Author) [pt

  11. 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)

  12. 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

  13. 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)

  14. [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.

  15. Transgenic plants as vital components of integrated pest management

    NARCIS (Netherlands)

    Kos, Martine; van Loon, J.J.A.; Dicke, M.; Vet, L.E.M.

    2009-01-01

    Although integrated pest management (IPM) strategies have been developed worldwide, further improvement of IPM effectiveness is required. The use of transgenic technology to create insect-resistant plants can offer a solution to the limited availability of highly insect-resistant cultivars.

  16. Analysis of chemical components from plant tissue samples

    Science.gov (United States)

    Laseter, J. L.

    1972-01-01

    Information is given on the type and concentration of sterols, free fatty acids, and total fatty acids in plant tissue samples. All samples were analyzed by gas chromatography and then by gas chromatography-mass spectrometry combination. In each case the mass spectral data was accumulated as a computer printout and plot. Typical gas chromatograms are included as well as tables describing test results.

  17. Plutonium Finishing Plant (PFP) HVAC System Component Index; FINAL

    International Nuclear Information System (INIS)

    DICK, J.D.

    1999-01-01

    This document identities the components, design media, procedures and defines the critical characteristics of Commercial Grade Items necessary to ensure the HVAC system provides these functions. This document lists safety class (SC) and safety significant (SS) components for the Heating Ventilation Air Conditioning (HVAC) and specifies the critical characteristics for Commercial Grade Items (CGI), as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to properly perform its safety function. There may be several manufacturers or models that meet the critical characteristics for any one item

  18. Determination of plant components degradation using ultrasonic C-scan

    International Nuclear Information System (INIS)

    Mohamad Pauzi Ismail; Suhairy Sani; Abdul Nassir Ibrahim

    2002-01-01

    C-scan Ultrasonic Inspection technique is increasingly used for the assessment of plant integrity. Due to the advancement of the equipment, Probability of Detection (POD) of this technique increased significantly as compared with the conventional techniques. Thus in many cases, the technique is accepted by engineers to be used to replace the conventional inspection methods such as visual inspections, thickness gauging and ultrasonic B-Scan. Thickness gauging and ultrasonic B-scan is still widely used by industries. However, both techniques have their own disadvantages. The most notable disadvantages of these techniques are related to the reliability of readings given by the equipment. In addition to this, thickness gauge would only provide data at certain points and B-scan would only provide data for certain lines. This paper presents and discusses results of C-scan measurement performed in power generation, chemical and petro-chemical plants. Due to its high accuracy, results from these measurements were used to establish the true condition of plant and to calculate its remaining safe life. Results presented in this paper include those related to corrosion, erosion and lamination in acid and gas pipelines, finger sludge catcher, steam drums in vessels and piping and electron beam machine. (Author)

  19. 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)

  20. 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.

  1. A Further Study of Productive Failure in Mathematical Problem Solving: Unpacking the Design Components

    Science.gov (United States)

    Kapur, Manu

    2011-01-01

    This paper replicates and extends my earlier work on productive failure in mathematical problem solving (Kapur, doi:10.1007/s11251-009-9093-x, 2009). One hundred and nine, seventh-grade mathematics students taught by the same teacher from a Singapore school experienced one of three learning designs: (a) traditional lecture and practice (LP), (b)…

  2. The impact of feedwater and condensate return excursions on boiler system component failures

    Energy Technology Data Exchange (ETDEWEB)

    Esmacher, Mel J. [GE Water and Process Technologies, The Woodlands, TX (United States); Rossi, Anthony [GE Water and Process Technologies, Trevose, PA (United States)

    2010-02-15

    During boiler operation, the transport of contaminants in boiler feedwater or condensate return via hardness excursions or transport of metal oxides due to corrosion can cause fouling and subsequent tube failure due to under-deposit corrosion or overheating. Case histories are reviewed and suitable corrective actions discussed. (orig.)

  3. Impact of the specialization from failures data in probability safety analysis for process plants

    International Nuclear Information System (INIS)

    Ribeiro, Antonio C.O.; Melo, P.F. Frutuoso e

    2005-01-01

    Full text: The aim of this paper is to show the Bayesian inference in reliability studies, which are used to failures, rates updating in safety analyses. It is developed the impact of its using in quantitative risk assessments (QRA) for industrial process plants. With this approach we find a structured and auditable way of showing the difference between an industrial installation with a good project and maintenance structure from another one that shows a low level of quality in these areas. In general the evidence from failures rates and as follow the frequency of occurrence from scenarios, which the risks taken in account in ERA, are taken from generics data banks, instead of, the installation in analysis. The use of this methodology in probabilistic safety analysis (PSA) for nuclear plants is commonly used when you need to find the final fault tree event evaluation applied to a scenario, but it is not showed in a PSA level III. (author)

  4. 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

  5. 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

  6. Qualification of electronic components for use in nuclear power plants

    International Nuclear Information System (INIS)

    Zorrilla, J.; Antonaccio, E; Luraschi, C.; Rodriguez, F.; Ranalli, J.; Ponce, M.; Dotro, R.; Guinda, J.

    2013-01-01

    There are a large number of instrument subjected to different service condition in a NPP. For instance different instruments can be found working in environment where the dose rate goes from negligible levels up to very harsh radiation levels. When technical specification and or equipment purchasing should be carried out it is possible to find the total leak of qualified instrument. In this context there is a need of dedicated qualification. In this work two different radiation resistance for two different I&C equipment/component were studies. The first I&C equipment was an LVDT (liner variable differential transformer). This equipment was tested while it was actuated in a strong gamma field in order to evaluate possible electromagnetic interferences a number of cycles equivalent to one year of service. After that the component was subjected to accelerated radiation aging and then actuated test under gamma field were carried out. The second I&C component to be tested was an (author)

  7. Systematic analysis and prevention of human originated common cause failures in relation to maintenance activities at Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Laakso, K. [VTT Industrial Systems, Espoo (Finland)

    2006-12-15

    inspections and functional testing. Such a planning case shall include a need evaluation of both the component and system level testing at the end of work. In addition, the use of condition monitoring information helps in reducing the uncertainty about equipment operability after intrusive maintenance or modifications. An increased use of condition information for a situation specific steering of preventive maintenance actions could also help to avoid unnecessary predetermined preventive maintenance actions which have the potential to cause unnecessary failures. A more agile adjustment of work orders to correspond the new conditions identified by maintenance personnel locally during work was also found necessary to reduce human CCFs caused by e.g. missing testing or inspections. The results emphasize the responsibility and requirements of versatility and specialisation of the planning and performance of maintenance and operability verification which brings this work to a knowledge work. For instance, flexibility is needed for the adaptation to specific conditions of the equipment and work as well as adhering to the rules and procedures is required. And mostly both a specialist knowledge of the equipment and a functional overview of the system are required. The event and error analyses of the multiple and single errors would help in training the maintenance, operation and technical personnel to identify better error mechanisms and prevent undetected human CCFs and errors, too. An earlier and better defined examination of the CCF risks as a part of the failure reporting and repair and modification processes would also help to identify, investigate and prevent CCFs. Generally, operability verification of the work objects in plant equipment should be planned and implemented better as an integral part of the plant maintenance process requiring knowledge of both the maintenance and operation branches. Methods for analysis of maintenance history information, examples of presentation

  8. Systematic analysis and prevention of human originated common cause failures in relation to maintenance activities at Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Laakso, K.

    2006-12-01

    inspections and functional testing. Such a planning case shall include a need evaluation of both the component and system level testing at the end of work. In addition, the use of condition monitoring information helps in reducing the uncertainty about equipment operability after intrusive maintenance or modifications. An increased use of condition information for a situation specific steering of preventive maintenance actions could also help to avoid unnecessary predetermined preventive maintenance actions which have the potential to cause unnecessary failures. A more agile adjustment of work orders to correspond the new conditions identified by maintenance personnel locally during work was also found necessary to reduce human CCFs caused by e.g. missing testing or inspections. The results emphasize the responsibility and requirements of versatility and specialisation of the planning and performance of maintenance and operability verification which brings this work to a knowledge work. For instance, flexibility is needed for the adaptation to specific conditions of the equipment and work as well as adhering to the rules and procedures is required. And mostly both a specialist knowledge of the equipment and a functional overview of the system are required. The event and error analyses of the multiple and single errors would help in training the maintenance, operation and technical personnel to identify better error mechanisms and prevent undetected human CCFs and errors, too. An earlier and better defined examination of the CCF risks as a part of the failure reporting and repair and modification processes would also help to identify, investigate and prevent CCFs. Generally, operability verification of the work objects in plant equipment should be planned and implemented better as an integral part of the plant maintenance process requiring knowledge of both the maintenance and operation branches. Methods for analysis of maintenance history information, examples of presentation

  9. A Study on Estimating the Next Failure Time of Compressor Equipment in an Offshore Plant

    Directory of Open Access Journals (Sweden)

    SangJe Cho

    2016-01-01

    Full Text Available The offshore plant equipment usually has a long life cycle. During its O&M (Operation and Maintenance phase, since the accidental occurrence of offshore plant equipment causes catastrophic damage, it is necessary to make more efforts for managing critical offshore equipment. Nowadays, due to the emerging ICTs (Information Communication Technologies, it is possible to send health monitoring information to administrator of an offshore plant, which leads to much concern on CBM (Condition-Based Maintenance. This study introduces three approaches for predicting the next failure time of offshore plant equipment (gas compressor with case studies, which are based on finite state continuous time Markov model, linear regression method, and their hybrid model.

  10. 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)

  11. Three-dimensional testing of power plant components

    International Nuclear Information System (INIS)

    Martin, A.

    1989-01-01

    Industrial photogrammetry is a dimensional checking procedure whose main advantages are the fast acquisition of the basic data (image), contactless inspection, and independent data processing. As a result of these basic characteristics, photogrammetry is particularly well suited to the maintenance of nuclear power plants. Since 1983, Framatome has employed photogrammetry in a number of cases for 3D dimensional checks and inspections of systems for repair purposes. To this day, e.g., the tube plates of steam generators have been inspected, the dimensional stability of the support rings in steam generators have been checked, and the alignment pins of fuel elements have been examined in this way. (orig.) [de

  12. Use of neural networks to monitor power plant components

    International Nuclear Information System (INIS)

    Ikonomopoulos, A.; Tsoukalas, L.H.

    1992-01-01

    A new methodology is presented for nondestructive evaluation (NDE) of check valve performance and degradation. Artificial neural network (ANN) technology is utilized for processing frequency domain signatures of check valves operating in a nuclear power plant (NPP). Acoustic signatures obtained from different locations on a check valve are transformed from the time domain to the frequency domain and then used as input to a pretrained neural network. The neural network has been trained with data sets corresponding to normal operation, therefore establishing a basis for check valve satisfactory performance. Results obtained from the proposed methodology demonstrate the ability of neural networks to perform accurate and quick evaluations of check valve performance

  13. Creep property testing of energy power plant component material

    International Nuclear Information System (INIS)

    Nitiswati, Sri; Histori; Triyadi, Ari; Haryanto, Mudi

    1999-01-01

    Creep testing of SA213 T12 boiler piping material from fossil plant, Suralaya has been done. The aim of the testing is to know the creep behaviour of SA213 T12 boiler piping material which has been used more than 10 yeas, what is the material still followed ideal creep curve (there are primary stage, secondary stage, and tertiary stage). This possibility could happened because the material which has been used more than 10 years usually will be through ageing process because corrosion. The testing was conducted in 520 0C, with variety load between 4% until 50% maximum allowable load based on strength of the material in 520 0C

  14. Failure rate evaluation for different components operating in sodium, based on operating experience of the RAPSODIE and the PHENIX reactors and the test loops

    International Nuclear Information System (INIS)

    Boisseau, J.; Dorey, J.; Hedin, F.; Le Floch, C.

    1982-01-01

    The failure rates of the following components, valves operating in sodium, mechanical and electromagnetic pumps, and heat exchangers including intermediate heat exchangers, cold traps, steam generators, are evaluated by analysing the main incidents which occurred on these components. Therefore, this paper contains an evaluation of the operating experience of components working in sodium and of the reliability of these components

  15. 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

  16. 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

  17. 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.

  18. 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

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

    OpenAIRE

    Kouroush Jenab; Joseph Pineau

    2015-01-01

    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 impa...

  20. 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

  1. Seismic fragility of nuclear power plant components (Phase 2): A fragility handbook on eighteen components

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.K.; Hofmayer, C.H.; Kassir, M.K.; Shteyngart, S.

    1991-06-01

    Fragility estimates of seven equipment classes were published in earlier reports. This report presents fragility analysis results from eleven additional equipment categories. The fragility levels are expressed in probabilistic terms. For users' convenience, this concluding report includes a summary of fragility results of all eighteen equipment classes. A set of conversion factors based on judgment is recommended for use of the information for early vintage equipment. The knowledge gained in conducting the Component Fragility Program and similar other programs is expected to provide a new direction for seismic verification and qualification of equipment. 15 refs., 12 tabs

  2. Proof of integrity and ageing management of mechanical components in nuclear power plants

    International Nuclear Information System (INIS)

    Roos, E.; Herter, K.-H.; Kockelmann, H.; Schuler, X.

    2005-01-01

    Demands and requirements for a safe operation of mechanical components during the whole operation life time (plant life management) to assure aging phenomena (aging management) and to prove the integrity (prove of integrity, e.g. in order to exclude large breaks) can be found in guidelines, codes and standards. In the present paper a general concept to proof the integrity as part of the ageing management of pressurized components and systems is presented. The concept is based on the actual material characteristics, the actual as-built configurations and the design of the components and systems including the knowledge of possible failure mechanism during operation. An important part of the assessment is the leak before break behavior and the break preclusion concept. Based on essential research results the developed procedures and methodologies for the assessment of the critical crack sizes as well as the critical loading conditions are reported and discussed. In detail the following aspects have to be treated: (a) evaluation of the as-built status of quality (design, construction, material, fabrication; results of recurrent non destructive examinations up to now, operational experience); (b) determination of the relevant loading conditions by means of in-service monitoring (monitoring of the mode of operation, the water chemistry, the mechanical and thermal stresses, the dynamic loading), emergency and faulted condition loads as specified; (c) evaluation of the actual status of quality with respect to the relevant loading conditions (stress analysis-limitation of the stresses; fatigue analysis-determination of the usage factor; fracture mechanics analysis-determination of crack growth, critical crack sizes and loading conditions); (d) evaluation and extent of the in-service monitoring and recurrent inspections to guarantee the succeeding operation (recurrent non destructive examination - minimum detectable flaw sizes, examination area, examination intervals; leak

  3. Cost analysis of small hydroelectric power plants components and preliminary estimation of global cost

    International Nuclear Information System (INIS)

    Basta, C.; Olive, W.J.; Antunes, J.S.

    1990-01-01

    An analysis of cost for each components of Small Hydroelectric Power Plant, taking into account the real costs of these projects is shown. It also presents a global equation which allows a preliminary estimation of cost for each construction. (author)

  4. Lithuanian requirements for ageing management of systems and components important to safety of nuclear power plant

    International Nuclear Information System (INIS)

    Ramanauskiene, A.

    2000-01-01

    In this paper the Lithuanian requirements for ageing management of systems and components important to safety of Ignalina nuclear power plant (two RBMK-1500 water-cooled graphite moderated channel-type power reactors) are presented

  5. General requirements for pressure-retaining systems and components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1991-11-01

    This standard specifies the general requirements for the design, fabrication and installation of pressure-retaining systems, components, and their supports in CANDU nuclear power plants. (16 figs., 2 tabs., 25 refs.)

  6. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae [Sungkwunkwan Univ., Seoul (Korea, Republic of); Lee, Hyang Beom [Soongsil Univ., Seoul (Korea, Republic of); Shin, Young Kil [Kunsan National Univ., Gunsan (Korea, Republic of); Chung, Hyun Jo [Wonkwang Univ., Iksan (Korea, Republic of); Park, Ik Keun; Park, Eun Soo [Seoul National University of Technology, Seoul (Korea, Republic of)

    2001-03-15

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing.

  7. Development of life evaluation technology for nuclear power plant components

    International Nuclear Information System (INIS)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae; Lee, Hyang Beom; Shin, Young Kil; Chung, Hyun Jo; Park, Ik Keun; Park, Eun Soo

    2001-03-01

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing

  8. An approach to safety problems relating to ageing of nuclear power plant components

    International Nuclear Information System (INIS)

    Conte, M.; Deletre, G.; Henry, J.Y.; Le Meur, M.

    1989-10-01

    The safety of nuclear power plants, in France, is discussed. The attention is focused on the ageing phenomena, as a potential cause of the degradation of the systems functional capabilities. The allowance for ageing in design and its importance on safety, are analyzed. The understanding of phenomena relating to ageing and the components surveillance, are considered. As the effective ageing on the components of nuclear power plants is not fully understood, technical improvements and more accurate analysis are required

  9. Secondary prevention- an essential component of the comprehensive rehabilitation of patients with heart failure

    Directory of Open Access Journals (Sweden)

    Pop Dana

    2017-12-01

    Full Text Available Heart failure is currently a real public health problem due to the extremely high morbidity and mortality of this disease. In this context, cardiovascular prevention measures should be implemented as early as possible. In addition to classic prevention measures, a number of extremely important specific recommendations should be considered: informing patients about their underlying disease, identifying the cardiovascular and non-cardiovascular factors that have led to cardiac decompensation, reducing daily salt consumption, monitoring body weight, forbidding smoking and recreational substances, conducting a regular exercise program under supervision, and increasing adherence to treatment.

  10. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho

    2016-01-01

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B 4 C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  11. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B{sub 4}C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  12. [Herbalism, botany and components analysis study on original plants of frankincense].

    Science.gov (United States)

    Sun, Lei; Xu, Jimin; Jin, Hongyu; Tian, Jingai; Lin, Ruichao

    2011-01-01

    In order to clarify original plants of traditional Chinese medicine (TCM) frankincense, a GC method for determination essential oils and a HPLC method for determination boswellic acids were carried out together with analysis of herbalism, botany, components and pharmacology papers of frankincense. It was concluded that original plants of TCM frankincense include at least Boswellia sacra, B. papyrifera and B. serrata.

  13. 76 FR 27669 - Automotive Components Holdings, LLC, a Subsidiary of Ford Motor Company, Saline Plant Division...

    Science.gov (United States)

    2011-05-12

    ... Holdings, LLC, a Subsidiary of Ford Motor Company, Saline Plant Division, Including Workers Whose Wages Were Reported Under Ford Company, Visteon, MSX International, W.J. O'Neil Company, and Unibar, Saline... workers of Automotive Components Holdings, LLC, a Subsidiary of Ford Motor Company, Saline Plant Division...

  14. Remaining life assessment and plant life extension in high temperature components of power and petrochemical plant

    International Nuclear Information System (INIS)

    Fleming, A.

    2003-01-01

    This paper explains the reasons why plant life can so easily be extended beyond the original design life. It details the means by which plant life extension is normally achieved, a structured plan for achieving such plant life extension at reasonable cost and some of the key techniques used in assessing the remaining life and discusses the simple repair options available. (author)

  15. Dead or Alive? Using Membrane Failure and Chlorophyll a Fluorescence to Predict Plant Mortality from Drought.

    Science.gov (United States)

    Guadagno, Carmela R; Ewers, Brent E; Speckman, Heather N; Aston, Timothy Llewellyn; Huhn, Bridger J; DeVore, Stanley B; Ladwig, Joshua T; Strawn, Rachel N; Weinig, Cynthia

    2017-09-01

    Climate models predict widespread increases in both drought intensity and duration in the next decades. Although water deficiency is a significant determinant of plant survival, limited understanding of plant responses to extreme drought impedes forecasts of both forest and crop productivity under increasing aridity. Drought induces a suite of physiological responses; however, we lack an accurate mechanistic description of plant response to lethal drought that would improve predictive understanding of mortality under altered climate conditions. Here, proxies for leaf cellular damage, chlorophyll a fluorescence, and electrolyte leakage were directly associated with failure to recover from drought upon rewatering in Brassica rapa (genotype R500) and thus define the exact timing of drought-induced death. We validated our results using a second genotype (imb211) that differs substantially in life history traits. Our study demonstrates that whereas changes in carbon dynamics and water transport are critical indicators of drought stress, they can be unrelated to visible metrics of mortality, i.e. lack of meristematic activity and regrowth. In contrast, membrane failure at the cellular scale is the most proximate cause of death. This hypothesis was corroborated in two gymnosperms ( Picea engelmannii and Pinus contorta ) that experienced lethal water stress in the field and in laboratory conditions. We suggest that measurement of chlorophyll a fluorescence can be used to operationally define plant death arising from drought, and improved plant characterization can enhance surface model predictions of drought mortality and its consequences to ecosystem services at a global scale. © 2017 American Society of Plant Biologists. All Rights Reserved.

  16. 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

  17. 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

  18. Correlated seed failure as an environmental veto to synchronize reproduction of masting plants.

    Science.gov (United States)

    Bogdziewicz, Michał; Steele, Michael A; Marino, Shealyn; Crone, Elizabeth E

    2018-07-01

    Variable, synchronized seed production, called masting, is a widespread reproductive strategy in plants. Resource dynamics, pollination success, and, as described here, environmental veto are possible proximate mechanisms driving masting. We explored the environmental veto hypothesis, which assumes that reproductive synchrony is driven by external factors preventing reproduction in some years, by extending the resource budget model of masting with correlated reproductive failure. We ran this model across its parameter space to explore how key parameters interact to drive seeding dynamics. Next, we parameterized the model based on 16 yr of seed production data for populations of red (Quercus rubra) and white (Quercus alba) oaks. We used these empirical models to simulate seeding dynamics, and compared simulated time series with patterns observed in the field. Simulations showed that resource dynamics and reproduction failure can produce masting even in the absence of pollen coupling. In concordance with this, in both oaks, among-year variation in resource gain and correlated reproductive failure were necessary and sufficient to reproduce masting, whereas pollen coupling, although present, was not necessary. Reproductive failure caused by environmental veto may drive large-scale synchronization without density-dependent pollen limitation. Reproduction-inhibiting weather events are prevalent in ecosystems, making described mechanisms likely to operate in many systems. © 2018 The Authors New Phytologist © 2018 New Phytologist Trust.

  19. How insects overcome two-component plant chemical defence: plant β-glucosidases as the main target for herbivore adaptation.

    Science.gov (United States)

    Pentzold, Stefan; Zagrobelny, Mika; Rook, Fred; Bak, Søren

    2014-08-01

    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-component chemical defence. These adaptations include host plant choice, non-disruptive feeding guilds and various physiological adaptations as well as metabolic enzymatic strategies of the insect’s digestive system. Furthermore, insect adaptations often act in combination, may exist in both generalists and specialists, and can act on different classes of defence compounds. We discuss how generalist and specialist insects appear to differ in their ability to use these different types of adaptations: in generalists, adaptations are often inducible, whereas in specialists they are often constitutive. Future studies are suggested to investigate in detail how insect adaptations act in combination to overcome plant chemical defences and to allow ecologically relevant conclusions.

  20. Method for estimating failure probabilities of structural components and its application to fatigue problem of internally cooled superconductors

    International Nuclear Information System (INIS)

    Shibui, M.

    1989-01-01

    A new method for fatigue-life assessment of a component containing defects is presented such that a probabilistic approach is incorporated into the CEGB two-criteria method. The present method assumes that aspect ratio of initial defect, proportional coefficient of fatigue crack growth law and threshold stress intensity range are treated as random variables. Examples are given to illustrate application of the method to the reliability analysis of conduit for an internally cooled cabled superconductor (ICCS) subjected to cyclic quench pressure. The possible failure mode and mechanical properties contributing to the fatigue life of the thin conduit are discussed using analytical and experimental results. 9 refs., 9 figs

  1. Service life monitoring of the main components at the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Hahn, J.; Vincour, D.

    2007-01-01

    Knowledge and experience gained from the introduction and periodical implementation of life assessment of the major components of the Temelin nuclear power plant is summarized. The initial Soviet technical design of the plant did not incorporate lifetime monitoring and evaluation, therefore it was completed with demonstrative strength and lifetime calculations from Czech companies. Moreover, a Westinghouse primary circuit diagnosis and monitoring system, including the monitoring of temperature and pressure cycles for low-cycle fatigue evaluation, was installed at the plant. The DIALIFE code for the calculation of mainly the low-cycle fatigue of the key pressure components, was developed and installed subsequently as a superstructure to the monitoring system. (author)

  2. 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)

  3. Reliability Analysis of Fatigue Failure of Cast Components for Wind Turbines

    DEFF Research Database (Denmark)

    Rafsanjani, Hesam Mirzaei; Sørensen, John Dalsgaard

    2015-01-01

    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...

  4. Systematic Characterization of Component Failures for the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Petersen, P.I.

    1999-01-01

    A fusion reactor will be a fairly complex system consisting of many components. All the components are required to work in order to produce a plasma and control it. Some of the components will be large, and for economic reasons there will not be spares for all components. It is therefore important to have a system whereby troubles are communicated, recorded and analyzed. Such a trouble report system has been in place at the DIII-D tokamak facility for many years. The purpose of the system is to easily facilitate communication between the people that discover problems and those that fix the problems. The trouble sheets are logged into a computer database that is used to characterize the kind of problems that the facility experiences, and determine which equipment, software, or human errors are causing significant downtime. The information is also used to evaluate whether sufficient maintenance is done to the equipment and to provide a basis for replacing it. The original system was based on paper forms. About a year ago the system was changed to a web-based system. In the new system a trouble report is filled out using a web browser, and the information is emailed to the repair personnel and managers as soon as the form is submitted through the web. The paper will discuss the problems experienced at the DIII-D facility, and how the information is used to adjust the preventive maintenance schedule

  5. Reliability optimization for series systems under uncertain component failure rates in the design phase

    NARCIS (Netherlands)

    Ge, Q.; Peng, H.; van Houtum, G.J.J.A.N.; Adan, I.J.B.F.

    2018-01-01

    We develop an optimization model to determine the reliability design of critical components in a serial system. The system is under a service contract, and a penalty cost has to be paid by the OEM when the total system down time exceeds a predetermined level, which complicates the evaluation of the

  6. 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)

  7. Development of Web-Based Common Cause Failure (CCF) Database Module for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Hyun-Gyo; Hwang, Seok-Won; Shin, Tae-young

    2015-01-01

    Probabilistic safety assessment (PSA) has been used to identify risk vulnerabilities and derive the safety improvement measures from construction to operation stages of nuclear power plants. In addition, risk insights from PSA can be applied to improve the designs and operation requirements of plants. However, reliability analysis methods for quantitative PSA evaluation have essentially inherent uncertainties, and it may create a distorted risk profiles because of the differences among the PSA models, plant designs, and operation status. Therefore, it is important to ensure the quality of the PSA model so that analysts identify design vulnerabilities and utilize risk information. Especially, the common cause failure (CCF) has been pointed out as one of major issues to be able to cause the uncertainty related to the PSA analysis methods and data because CCF has a large influence on the PSA results. Organization for economic cooperation and development /nuclear energy agent (OECD/NEA) has implemented an international common cause failure data exchange (ICDE) project for the CCF quality assurance through the development of the detailed analysis methodologies and data sharing. However, Korea Hydro and Nuclear Power company (KHNP) does not have the basis for the data gathering and analysis for CCF analyses. In case of methodology, the Alpha Factor parameter estimation, which can analyze uncertainties and estimate an interface factor (Impact Vector) with an ease, is ready to be applied rather than the Multi Greek Letter (MGL) method. This article summarizes the development of the plant-specific CCF database (DB) module considering the raw data collection and the analysis procedure based on the CCF parameter calculation method of ICDE. Although the portion affected by CCF in the PSA model is quite a large, the development efforts of the tools to collect and analyze data were insufficient. Currently, KHNP intends to improve PSA quality and ensure CCF data reliability by

  8. Development of Web-Based Common Cause Failure (CCF) Database Module for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun-Gyo; Hwang, Seok-Won; Shin, Tae-young [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Probabilistic safety assessment (PSA) has been used to identify risk vulnerabilities and derive the safety improvement measures from construction to operation stages of nuclear power plants. In addition, risk insights from PSA can be applied to improve the designs and operation requirements of plants. However, reliability analysis methods for quantitative PSA evaluation have essentially inherent uncertainties, and it may create a distorted risk profiles because of the differences among the PSA models, plant designs, and operation status. Therefore, it is important to ensure the quality of the PSA model so that analysts identify design vulnerabilities and utilize risk information. Especially, the common cause failure (CCF) has been pointed out as one of major issues to be able to cause the uncertainty related to the PSA analysis methods and data because CCF has a large influence on the PSA results. Organization for economic cooperation and development /nuclear energy agent (OECD/NEA) has implemented an international common cause failure data exchange (ICDE) project for the CCF quality assurance through the development of the detailed analysis methodologies and data sharing. However, Korea Hydro and Nuclear Power company (KHNP) does not have the basis for the data gathering and analysis for CCF analyses. In case of methodology, the Alpha Factor parameter estimation, which can analyze uncertainties and estimate an interface factor (Impact Vector) with an ease, is ready to be applied rather than the Multi Greek Letter (MGL) method. This article summarizes the development of the plant-specific CCF database (DB) module considering the raw data collection and the analysis procedure based on the CCF parameter calculation method of ICDE. Although the portion affected by CCF in the PSA model is quite a large, the development efforts of the tools to collect and analyze data were insufficient. Currently, KHNP intends to improve PSA quality and ensure CCF data reliability by

  9. 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

  10. Seismic fragility of nuclear power plant components: Phase 2, Motor control center, switchboard, panelboard and power supply

    International Nuclear Information System (INIS)

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

    1987-12-01

    In Phase I of the Component Fragility Program, Brookhaven National Laboratory (BNL) has developed a procedure to establish the seismic fragility of nuclear power plant equipment by use of existing test data and demonstrated its application by considering two equipment pieces. In Phase II of the program, BNL has collected additional test data, and has further advanced and is applying the methodology to determine the fragility levels of selected essential equipment categories. The data evaluation of four equipment families, namely, motor control center, switchboard, panelboard and power supply has been completed. Fragility levels have been determined for various failure modes of each equipment class and the deterministic results are presented in terms of test response spectra. In addition, the test data have been analyzed for determination of the respective probabilistic fragility levels. To this end, a single g-value has been selected to approximately represent the test vibration level and a statistical analysis has been performed with the g-values corresponding to a particular failure mode. The zero period acceleration and the average spectral acceleration over a frequency range of interest are separately used as the single g-value. The resulting parameters are presented in terms of a median value, an uncertainty coefficient and a randomness coefficient. Ultimately, each fragility level is expressed in terms of a single descriptor called an HCLPF value corresponding to a high (95%) confidence of a low (5%) probability of failure. The important observations made in the process of data analysis are included in this report

  11. 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

  12. 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

  13. Responsiveness of cold tolerant chickpea characteristics in fall and spring planting: II. yield and yield components

    Directory of Open Access Journals (Sweden)

    ahmad nezami

    2009-06-01

    Full Text Available Previous research in Mashhad collection chickpeas (MCC has shown that there are some cold tolerant genotypes for fall planting in the highlands. To obtain more detailed information about the reaction of these genotypes to fall and spring planting, the yield and yield component responses of 33 chickpea genotypes (32 cold tolerant genotypes and one susceptible genotypes to four planting dates (28 Sep., 16 Oct., 2 Nov., and 7 Mar. were evaluated in 2000-2001 growing season. The experiment was conducted at the experimental field of college of agriculture, Ferdowsi University of Mashhad as a split plot design with two replications. The planting dates were imposed as main plot and chickpea genotypes as subplot. Effects of planting date and genotype on percent of plant survival (PPS after winter, number. of pod per plant, 100 seed weight, yield and Harvest Index (HI were significant (p

  14. Design and fabrication of stainless steel components for long life of spent fuel reprocessing plants

    International Nuclear Information System (INIS)

    Natarajan, R.; Ramkumar, P.; Sundararaman, V.; Kamachi Mudali, U.; Baldev Raj; Shanmugam, K.

    2010-01-01

    Reprocessing of spent nuclear fuels based on the PUREX process is the proven process with many commercial plants operating satisfactorily worldwide. The process medium being nitric acid, austenitic stainless steel is the material of construction as it is the best commercially available material for meeting the conditions in the reprocessing plants. Because of the high radiation fields, contact maintenance of equipment and systems of these plants are very time consuming and costly unlike other chemical process plants. Though the plants constructed in the early years required extensive shut downs for replacement of equipment and systems within the first fifteen years of operation itself, development in the field of stainless steel metallurgy and fabrication techniques have made it possible to design the present day plants for an operating life period of forty years. A review of the operational experience of the PUREX process based aqueous reprocessing plants has been made in this paper and reveals that life limiting failures of equipment and systems are mainly due to corrosion while a few are due to stresses. Presently there are no standards for design specification of materials and fabrication of reprocessing plants like the nuclear power plants, where well laid down ASTM and ASME codes and standards are available which are based on the large scale operational feedbacks on pressure vessels for conventional and nuclear industries. (author)

  15. 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

  16. Screening tests of representative nuclear power plant components exposed to secondary environments created by fires

    International Nuclear Information System (INIS)

    Jacobus, M.J.

    1986-06-01

    This report presents results of screening tests to determine component survivability in secondary environments created by fires, specifically increased temperatures, increased humidity, and the presence of particulates and corrosive vapors. Additionally, chloride concentrations were measured in the exhaust from several of the tests used to provide fire environments. Results show actual failure or some indication of failure for strip chart recorders, electronic counters, an oscilloscope amplifier, and switches and relays. The chart recorder failures resulted from accumulation of particulates on the pen slider mechanisms. The electronic counter experienced leakage current failures on circuit boards after the fire exposure and exposure to high humidity. The oscillosocpe amplifier experienced thermal-related drift as high as 20% before thermal protective circuitry shut the unit down. In some cases, switches and relays experienced high contact resistances with the low voltages levels used for the mesurements. Finally, relays tested to thermal failure experienced various failures, all at temperatures ranging from 150 0 C to above 350 0 C. The chloride measurements show that most of the hydrogen chloride generated in the test fires is combined with particulate by the time it reaches the exhaust duct, indicating that hydrogen chloride condensation may be less likely than small scale data implies. 13 refs., 36 figs

  17. 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)

  18. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre

    International Nuclear Information System (INIS)

    2007-05-01

    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

  19. Failure analysis of a boiler tube in USC coal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, N.H.; Kim, S.; Choe, B.H.; Yoon, K.B.; Kwon, D.I. [Kangnung National University, Kangnung (Republic of Korea)

    2009-10-15

    This paper presents failure analysis of final superheater tube in ultra-supercritical (USC) coal power plant. Visual inspection was performed to find out the characteristics of fracture of the as-received material. And the micro-structural changes such as grain growth and carbide coarsening was examined by scanning electron microscope. Detailed microscopic studies were made to find out the behavior of the scale exfoliation on the waterside of tubes. From those investigations, the creep rupture may be caused by the softened structure induced by carbide coarsening and accelerated by the metal temperature increase by the impediment of heat transfer due to voids.

  20. 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

  1. Deep knowledge expert system for diagnosis of multiple-failure severe transients in nuclear power plant

    International Nuclear Information System (INIS)

    Martin, R.P.; Nassersharif, B.

    1987-01-01

    TAMUS (Transient Analysis of MUltiple-failure Simulations) is a prototype expert system which is the result of a project investigating and implementing event confidence-levels (used by reactor safety experts in reactor transient analysis) in the form of an expert system. Currently, TAMUS is designed to diagnose reactor transients by analyzing simulated sensor and plant thermal hydraulic information from a system simulation. TAMUS uses a knowledge base of existing emergency nuclear plant operating guidelines and detailed thermal-hydraulic calculating results correlated to confidence-levels. TAMUS can diagnose a number of reactor transients (for example, loss-of-coolant accidents, steam-generator-tube ruptures, loss-of-offsite power, etc.). Future work includes the expansion of the knowledge base and improvement of the deep-knowledge qualitative models

  2. 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

  3. 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.

  4. Maintenance service for major component of PWR plant. Replacement of pressurizer safe end weld

    International Nuclear Information System (INIS)

    Miyoshi, Yoshiyuki; Kobayashi, Yuki; Yamamoto, Kazuhide; Ueda, Takeshi; Suda, Naoki; Shintani, Takashi

    2017-01-01

    In October 2016, MHI completed the replacement of safe end weld of pressurizer (Pz) of Ringhals unit 3, which was the first maintenance work for main component of pressurized water reactor (PWR) plant in Europe. For higher reliability and longer lifetime of PWR plant, MHI has conducted many kinds of maintenance works of main components of PWR plants in Japan against stress corrosion cracking due to aging degradation. Technical process for replacement of Pz safe end weld were established by MHI. MHI has experienced the work for 21 PWR units in Japan. That of Ringhals unit 3 was planned and conducted based on the experiences. In this work, Alloy 600 used for welds of nozzles of Pz was replaced with Alloy 690. Alloy 690 is more corrosive-resistant than Alloy 600. Specially designed equipment and technical process were developed and established by MHI to replace safe end weld of Pz and applied for the Ringhals unit 3 as a first application in Europe. The application had been performed in success and achieved the planned replacement work duration and total radiation dose by using sophisticated machining and welding equipment designed to meet the requirements to be small, lightweight and remote-controlled and operating by well skilled MHI personnel experienced in maintenance activities for major components of PWR plant in Japan. The success shows that the experience, activities and technology developed in Japan for main components of PWR plant shall be applicable to contribute reliable operations of nuclear power plants in Europe and other countries. (author)

  5. Offshore Wind Power Plant Technology Catalogue - Components of wind power plants, AC collection systems and HVDC systems

    DEFF Research Database (Denmark)

    Das, Kaushik; Antonios Cutululis, Nicolaos

    2017-01-01

    Traditionally, Offshore Wind Power Plants (OWPPs) are connected through many com-ponents as shown in the figure 1. An OWPP consists of controllable, variable speed Wind Turbines (WTs). These WTs are connected through Medium Voltage (MV) sub-marine cables typically at voltage level of upto 33-66 k...... for the cables as well reduce the power losses through them....

  6. Tests of qualification of national components of nuclear power plants under design basis accident

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1990-01-01

    With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

  7. Application of some pattern recognition methods for the early detection of failures at NPP components by means of noise diagnosis

    International Nuclear Information System (INIS)

    Weiss, F.P.

    1985-01-01

    The automation of the decisions on normality or abnormality of the plant condition being based on automated measurements is an essential step for the integration of noise diagnostics into the control and safety system of a nuclear power plant. By reason of the stochastic character of noise diagnostic measuring quantities principles of statistical pattern recognition are used in order automatically to get a decision on the plant condition. Four different pattern recognition methods complementing each other have been developed respectively tested with data from a WWER-440 type reactor. These four methods are included in a specially written software package. According to stationarity, correlation and probability distribution type of the state describing features and according to the necessary detection sensitivity to failures either the decorrelation method, the cluster method, the Parzen method or the distribution test of Wilcoxon, Mann and Whitney has to be applied. The efficiency and the limits of the investigated methods are discussed in detail. In context with the surveillance of the WWER-440 core by means of the power spectral densities of neutron flux fluctuations it could as well experimentally as theoretically be shown that the logarithmic power spectral densities follow a Gaussian probability distribution. (author)

  8. Nuclear power plant systems, structures and components and their safety classification

    International Nuclear Information System (INIS)

    2000-01-01

    The assurance of a nuclear power plant's safety is based on the reliable functioning of the plant as well as on its appropriate maintenance and operation. To ensure the reliability of operation, special attention shall be paid to the design, manufacturing, commissioning and operation of the plant and its components. To control these functions the nuclear power plant is divided into structural and functional entities, i.e. systems. A systems safety class is determined by its safety significance. Safety class specifies the procedures to be employed in plant design, construction, monitoring and operation. The classification document contains all documentation related to the classification of the nuclear power plant. The principles of safety classification and the procedures pertaining to the classification document are presented in this guide. In the Appendix of the guide, examples of systems most typical of each safety class are given to clarify the safety classification principles

  9. 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

  10. 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.

  11. Review On Feasibility of Using Satellite Imaging for Risk Management of Derailment Related Turnout Component Failures

    Science.gov (United States)

    Dindar, Serdar; Kaewunruen, Sakdirat; Osman, Mohd H.

    2017-10-01

    One of the emerging significant advances in engineering, satellite imaging (SI) is becoming very common in any kind of civil engineering projects e.g., bridge, canal, dam, earthworks, power plant, water works etc., to provide an accurate, economical and expeditious means of acquiring a rapid assessment. Satellite imaging services in general utilise combinations of high quality satellite imagery, image processing and interpretation to obtain specific required information, e.g. surface movement analysis. To extract, manipulate and provide such a precise knowledge, several systems, including geographic information systems (GIS) and global positioning system (GPS), are generally used for orthorectification. Although such systems are useful for mitigating risk from projects, their productiveness is arguable and operational risk after application is open to discussion. As the applicability of any novel application to the railway industry is often measured in terms of whether or not it has gained in-depth knowledge and to what degree, as a result of errors during its operation, this novel application generates risk in ongoing projects. This study reviews what can be achievable for risk management of railway turnouts thorough satellite imaging. The methodology is established on the basis of other published articles in this area and the results of applications to understand how applicable such imagining process is on railway turnouts, and how sub-systems in turnouts can be effectively traced/operated with less risk than at present. As a result of this review study, it is aimed that the railway sector better understands risk mitigation in particular applications.

  12. 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

  13. 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

  14. Controversy Associated With the Common Component of Most Transgenic Plants – Kanamycin Resistance Marker Gene

    OpenAIRE

    Jelenić, Srećko

    2003-01-01

    Plant genetic engineering is a powerful tool for producing crops resistant to pests, diseases and abiotic stress or crops with improved nutritional value or better quality products. Currently over 70 genetically modified (GM) crops have been approved for use in different countries. These cover a wide range of plant species with significant number of different modified traits. However, beside the technology used for their improvement, the common component of most GM crops is the neomycin phosp...

  15. A multi-criteria decision making system for damage assessment of critical components in power plants

    International Nuclear Information System (INIS)

    Jovanovic, A.; Auerkari, P.; Brear, J.M.

    1996-01-01

    A multi-criteria decision making tool for engineering applications has been developed in the European project BE5935. The tool has been developed and applied in the area of power plants, primarily for the decisions regarding the inspection and maintenance planning in the area of power plants. Practical application of the methodology and of the software is shown here for the damage assessment of critical components. (authors)

  16. 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.

  17. Advancements in the design of safety-related systems and components of the MARS nuclear plant

    International Nuclear Information System (INIS)

    Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

    1992-01-01

    In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

  18. Towards a more consolidated approach to material data management in life assessment of power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, A.; Maile, K. [MPA Stuttgart (Germany)

    1998-12-31

    The presentation discusses the necessity of having a more consolidated (unified, possibly `European`) framework for all (not only pure experimental) material data needed for optimized life management and assessment of high-temperature and other components in power and process plants. After setting the main requirements for such a system, a description of efforts done in this direction at MPA Stuttgart in the area of high-temperature components in power plants is given. Furthermore, a reference to other relevant efforts elsewhere is made and an example of practical application of the proposed solution described (optimized material selection and life assessment of high-temperature piping). (orig.) 10 refs.

  19. Towards a more consolidated approach to material data management in life assessment of power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, A; Maile, K [MPA Stuttgart (Germany)

    1999-12-31

    The presentation discusses the necessity of having a more consolidated (unified, possibly `European`) framework for all (not only pure experimental) material data needed for optimized life management and assessment of high-temperature and other components in power and process plants. After setting the main requirements for such a system, a description of efforts done in this direction at MPA Stuttgart in the area of high-temperature components in power plants is given. Furthermore, a reference to other relevant efforts elsewhere is made and an example of practical application of the proposed solution described (optimized material selection and life assessment of high-temperature piping). (orig.) 10 refs.

  20. Application of the Safety Classification of Structures, Systems and Components in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2016-04-01

    This publication describes how to complete tasks associated with every step of the classification methodology set out in IAEA Safety Standards Series No. SSG-30, Safety Classification of Structures, Systems and Components in Nuclear Power Plants. In particular, how to capture all the structures, systems and components (SSCs) of a nuclear power plant to be safety classified. Emphasis is placed on the SSCs that are necessary to limit radiological releases to the public and occupational doses to workers in operational conditions This publication provides information for organizations establishing a comprehensive safety classification of SSCs compliant with IAEA recommendations, and to support regulators in reviewing safety classification submitted by licensees

  1. 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

  2. Applications of cathodic protection for the protection of aqueous and soil corrosion of power plant components

    International Nuclear Information System (INIS)

    Sinha, A.K.; Mitra, A.K.; Bhakta, U.C.; Sanyal, S.K.

    2000-01-01

    Power plant components exposed to environments such as water and soil are susceptible to severe corrosion. Many times the effect of corrosion in power plant components can be catastrophic. The problem is aggravated for underground pipelines due to additional factors such as large network of pipelines, proximity to earth mat, high voltage transmission lines, corrosive chemicals, inadequate approach etc. Other components such as condenser water boxes, internals of pipelines, clarifier bridge structures, cooling water inlet gates and pipes etc. which are in continuous contact with water, are subjected to severe corrosion. The nature and locations of all such components are at places which are not accessible for routine maintenance and hence they require long term reliable protection against corrosion. Experience has shown that anti-corrosive coatings are inadequate in preventing corrosion and due to their location regular maintenance coatings are also not feasible. Under such circumstances the applications of cathodic protection provides a long term solution the design of cathodic protection, for such applications differs from the commonly employed cathodic protection for cross-country pipelines and submerged structures due to other complexities in the plant region and maintenance of the applied system. The present paper intends to discuss the applications of cathodic protection with suitable anti-corrosive coatings for protection of various power plant components and the specific features of each type of application. (author)

  3. Engineering failure assessment methods applied to pressurized components; Bruchmechanische Bewertung druckfuehrender Komponenten mittels ingenieurmaessiger Bewertungsverfahren

    Energy Technology Data Exchange (ETDEWEB)

    Zerbst, U.; Beeck, F.; Scheider, I.; Brocks, W. [GKSS-Forschungszentrum Geesthacht GmbH (Germany). Inst. fuer Werkstofforschung

    1998-11-01

    Under the roof of SINTAP (Structural Integrity Assessment Procedures for European Industry), a European BRITE-EURAM project, a study is being carried out into the possibility of establishing on the basis of existing models a standard European flaw assessment method. The R6 Routine and the ETM are important, existing examples in this context. The paper presents the two methods, explaining their advantages and shortcomes as well as common features. Their applicability is shown by experiments with two pressure vessels subject to internal pressure and flawed by a surface crack or a through-wall crack, respectively. Both the R6 Routine and the ETM results have been compared with results of component tests carried out in the 1980s at TWI and are found to yield acceptable conservative, i.e. sufficiently safe, lifetime predictions, as they do not give lifetime assessments which unduly underestimate the effects of flaws under operational loads. (orig./CB) [Deutsch] Gegenwaertig wird im Rahmen von SINTAP (Structural Integrity Assessment Procedures for European Industries), einem europaeischen BRITE-EURAM-Projekt geprueft, inwieweit auf der Grundlage vorhandener Modelle eine einheitliche europaeische Fehlerbewertungsmethode erstellt werden kann. Eine zentrale Stellung kommt dabei Verfahren wie der R6-Routine und dem ETM zu. In der vorliegenden Arbeit wurden beide Methoden vorgestellt, wobei ihre Vor- und Nachteile, aber auch ihre Gemeinsamkeiten herausgearbeitet wurden. Die Anwendung wurde an zwei innendruckbelasteten Behaeltern mit Oberflaechen- bzw. wanddurchdringendem Riss demonstriert. Sowohl R6-Routine als auch ETM ergaben im Vergleich mit am TWI zu Beginn der 80er Jahre durchgefuehrten Bauteilexperimenten eine vertretbare konservative Vorhersage, d.h. eine nicht allzu grosse Unterschaetzung der ertragbaren Last der Bauteile. (orig.)

  4. Tonometry revisited: perfusion-related, metabolic, and respiratory components of gastric mucosal acidosis in acute cardiorespiratory failure.

    Science.gov (United States)

    Jakob, Stephan M; Parviainen, Ilkka; Ruokonen, Esko; Kogan, Alexander; Takala, Jukka

    2008-05-01

    Mucosal pH (pHi) is influenced by local perfusion and metabolism (mucosal-arterial pCO2 gradient, DeltapCO2), systemic metabolic acidosis (arterial bicarbonate), and respiration (arterial pCO2). We determined these components of pHi and their relation to outcome during the first 24 h of intensive care. We studied 103 patients with acute respiratory or circulatory failure (age, 63+/-2 [mean+/-SEM]; Acute Physiology and Chronic Health Evaluation II score, 20+/-1; Sequential Organ Failure Assessment score, 8+/-0). pHi, and the effects of bicarbonate and arterial and mucosal pCO2 on pHi, were assessed at admission, 6, and 24 h. pHi was reduced (at admission, 7.27+/-0.01) due to low arterial bicarbonate and increased DeltapCO2. Low pHi (or=7.32 at admission; P=0.061) was associated with an increased DeltapCO2 in 59% of patients (mortality, 47% vs. 4% for patients with low pHi and normal DeltapCO2; P=0.0003). An increased versus normal DeltapCO2, regardless of pHi, was associated with increased mortality at admission (51% vs. 5%; Pacidosis. Inadequate tissue perfusion may persist despite stable hemodynamics and contributes to poor outcome.

  5. Material Selection for an Ultra High Strength Steel Component Based on the Failure Criteria of CrachFEM

    International Nuclear Information System (INIS)

    Kessler, L.; Beier, Th.; Werner, H.; Horstkott, D.; Dell, H.; Gese, H.

    2005-01-01

    An increasing use of combining more than one process step is noticed for coupling crash simulations with the results of forming operations -- mostly by inheriting the forming history like plastic strain and material hardening. Introducing a continuous failure model allows a further benefit of these coupling processes; it sometimes can even be the most attractive result of such a work. In this paper the algorithm CrachFEM for fracture prediction has been used to generate more benefit of the successive forming and crash simulations -- especially for ultra high strength steels. The choice and selection of the material grade in combination with the component design can therefore be done far before the prototyping might show an unsuccessful crash result; and in an industrial applicable manner

  6. 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.

  7. Modulating wind power plant output using different frequency modulation components for damping grid oscillations

    DEFF Research Database (Denmark)

    2017-01-01

    A method, controller, wind power plant, and computer program product are disclosed for operating a wind power plant comprising a plurality of wind turbines, the wind power plant producing a plant power output. The method comprises receiving a modulation request signal indicating a requested...... modulation of the plant power output, the requested modulation specifying a modulation frequency. The method further comprises generating a respective power reference signal for each of at least two wind turbines of the plurality of wind turbines selected to fulfill the requested modulation, Each generated...... power reference signal includes a respective modulation component corresponding to a portion of the requested modulation and having a frequency different than the modulation frequency....

  8. Life Cycle Management Managing the Aging of Critical Nuclear Plant Components

    International Nuclear Information System (INIS)

    Meyer, Theodore A.; Elder, G. Gary; Llovet, Ricardo

    2002-01-01

    Life Cycle Management is a structured process to manage equipment aging and long-term equipment reliability for nuclear plant Systems, Structures and Components (SSCs). The process enables the identification of effective repair, replace, inspect, test and maintenance activities and the optimal timing of the activities to maximize the economic value to the nuclear plant. This paper will provide an overview of the process and some of the tools that can be used to implement the process for the SSCs deemed critical to plant safety and performance objectives. As nuclear plants strive to reduce costs, extend life and maximize revenue, the LCM process and the supporting tools summarized in this paper can enable development of a long term, cost efficient plan to manage the aging of the plant SSCs. (authors)

  9. Structural materials requirements for in-vessel components of fusion power plants

    International Nuclear Information System (INIS)

    Schaaf, B. van der

    2000-01-01

    The economic production of fusion energy is determined by principal choices such as using magnetic plasma confinement or generating inertial fusion energy. The first generation power plants will use deuterium and tritium mixtures as fuel, producing large amounts of highly energetic neutrons resulting in radiation damage in materials. In the far future the advanced fuels, 3 He or 11 B, determine power plant designs with less radiation damage than in the first generation. The first generation power plants design must anticipate radiation damage. Solid sacrificing armour or liquid layers could limit component replacements costs to economic levels. There is more than radiation damage resistance to determine the successful application of structural materials. High endurance against cyclic loading is a prominent requirement, both for magnetic and inertial fusion energy power plants. For high efficiency and compactness of the plant, elevated temperature behaviour should be attractive. Safety and environmental requirements demand that materials have low activation potential and little toxic effects under both normal and accident conditions. The long-term contenders for fusion power plant components near the plasma are materials in the range from innovative steels, such as reduced activation ferritic martensitic steels, to highly advanced ceramic composites based on silicon carbide, and chromium alloys. The steels follow an evolutionary path to basic plant efficiencies. The competition on the energy market in the middle of the next century might necessitate the riskier but more rewarding development of SiCSiC composites or chromium alloys

  10. [Features of calcium crystals and calcium components in 54 plant species in salinized habitats of Tianjin].

    Science.gov (United States)

    Xu, Jing-Jing; Ci, Hua-Cong; He, Xing-Dong; Xue, Ping-Ping; Zhao, Xue-Lai; Guo, Jian-Tan; Gao, Yu-Bao

    2012-05-01

    Plant calcium (Ca) is composed of dissociated Ca2+ and easily soluble, slightly soluble, and hard soluble combined Ca salts. The hard soluble Ca salts can often engender Ca crystals. To understand the Ca status in different growth form plants in salinized habitats, 54 plant species were sampled from the salinized habitats in Tianjin, with the Ca crystals examined by microscope and the Ca components determined by sequential fractionation procedure. More Ca crystals were found in 38 of the 54 plant species. In 37 of the 38 plant species, drusy and prismatic Ca oxalate crystals dominated, whereas the cystolith of Ca carbonate crystal only appeared in the leaves of Ficus carica of Moraceae. The statistics according to growth form suggested that deciduous arbors and shrubs had more Ca oxalate crystal, liana had lesser Ca oxalate crystal, and herbs and evergreen arbors had no Ca oxalate crystal. From arbor, shrub, liana to herb, the concentration of HCl-soluble Ca decreased gradually, while that of water soluble Ca was in adverse. The concentration of water soluble Ca in herbs was significantly higher than that in arbors and shrubs. This study showed that in salinized habitats, plant Ca crystals and Ca components differed with plant growth form, and the Ca oxalate in deciduous arbors and shrubs played an important role in withstanding salt stress.

  11. Structural health monitoring of power plant components based on a local temperature measurement concept

    International Nuclear Information System (INIS)

    Rudolph, Juergen; Bergholz, S.; Hilpert, R.; Jouan, B.; Goetz, A.

    2012-01-01

    The fatigue assessment of power plant components based on fatigue monitoring approaches is an essential part of the integrity concept and modern lifetime management. It is comparable to structural health monitoring approaches in other engineering fields. The methods of fatigue evaluation of nuclear power plant components based on realistic thermal load data measured on the plant are addressed. In this context the Fast Fatigue Evaluation (FFE) and Detailed Fatigue Calculation (DFC) of nuclear power plant components are parts of the three staged approach to lifetime assessment and lifetime management of the AREVA Fatigue Concept (AFC). The three stages Simplified Fatigue Estimation (SFE), Fast Fatigue Evaluation (FFE) and Detailed Fatigue Calculation (DFC) are characterized by increasing calculation effort and decreasing degree of conservatism. Their application is case dependent. The quality of the fatigue lifetime assessment essentially depends on one hand on the fatigue model assumptions and on the other hand on the load data as the basic input. In the case of nuclear power plant components thermal transient loading is most fatigue relevant. Usual global fatigue monitoring approaches rely on measured data from plant instrumentation. As an extension, the application of a local fatigue monitoring strategy (to be described in detail within the scope of this paper) paves the way of delivering continuously (nowadays at a frequency of 1 Hz) realistic load data at the fatigue relevant locations. Methods of qualified processing of these data are discussed in detail. Particularly, the processing of arbitrary operational load sequences and the derivation of representative model transients is discussed. This approach related to realistic load-time histories is principally applicable for all fatigue relevant components and ensures a realistic fatigue evaluation. (orig.)

  12. 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.

  13. 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

  14. Service experience with AISI type 316 steel components in CEGB Midlands Region power plant

    International Nuclear Information System (INIS)

    Plastow, B.; Bagnall, B.I.; Yeldham, D.E.

    1979-01-01

    The service performance of AISI Type 316 steel components in sections up to 100 mm thick in Power Plant of the Midlands Region of the C.E.G.B. is reviewed. A comparison is drawn between the satisfactory performance of components whose dimensional stability is not critical and the difficulties experienced when rapid rates of change of temperature cause distortion in thick section components. Weldment manufacture and performance are reviewed and both are considered to be satisfactory. In general the material has performed well and the difficulties due to distortion have been overcome by imposing operating regimes which limit rates of temperature change. (author)

  15. Safety philosophy and design principles for systems and components of nuclear power plant: external event

    International Nuclear Information System (INIS)

    Lopes, J.P.G.

    1986-01-01

    In nuclear power plants, some systems and components are designed to withstand external impacts. Such systems and components are those which have to perform their functions even during and after the occurrences of an earthquake, for example, fulfilling the safety objectives and avoiding the release of radioactive material to the environment. The aim of this report is to introduce the safety philosophy and design principles for systems/components to perform their functions during and after the occurrence of an earthquake, as applied by NUCLEN for Angra 2 and 3. (Author) [pt

  16. A code for simulation of human failure events in nuclear power plants: SIMPROC

    International Nuclear Information System (INIS)

    Gil, Jesus; Fernandez, Ivan; Murcia, Santiago; Gomez, Javier; Marrao, Hugo; Queral, Cesar; Exposito, Antonio; Rodriguez, Gabriel; Ibanez, Luisa; Hortal, Javier; Izquierdo, Jose M.; Sanchez, Miguel; Melendez, Enrique

    2011-01-01

    Over the past years, many Nuclear Power Plant organizations have performed Probabilistic Safety Assessments to identify and understand key plant vulnerabilities. As part of enhancing the PSA quality, the Human Reliability Analysis is essential to make a realistic evaluation of safety and about the potential facility's weaknesses. Moreover, it has to be noted that HRA continues to be a large source of uncertainty in the PSAs. Within their current joint collaborative activities, Indizen, Universidad Politecnica de Madrid and Consejo de Seguridad Nuclear have developed the so-called SIMulator of PROCedures (SIMPROC), a tool aiming at simulate events related with human actions and able to interact with a plant simulation model. The tool helps the analyst to quantify the importance of human actions in the final plant state. Among others, the main goal of SIMPROC is to check the Emergency Operating Procedures being used by operating crew in order to lead the plant to a safe shutdown plant state. Currently SIMPROC is coupled with the SCAIS software package, but the tool is flexible enough to be linked to other plant simulation codes. SIMPROC-SCAIS applications are shown in the present article to illustrate the tool performance. The applications were developed in the framework of the Nuclear Energy Agency project on Safety Margin Assessment and Applications (SM2A). First an introductory example was performed to obtain the damage domain boundary of a selected sequence from a SBLOCA. Secondly, the damage domain area of a selected sequence from a loss of Component Cooling Water with a subsequent seal LOCA was calculated. SIMPROC simulates the corresponding human actions in both cases. The results achieved shown how the system can be adapted to a wide range of purposes such as Dynamic Event Tree delineation, Emergency Operating Procedures and damage domain search.

  17. Revising and Updating the Plant Science Components of the Connecticut Vocational Agriculture Curriculum.

    Science.gov (United States)

    Connecticut Univ., Storrs. Dept. of Educational Leadership.

    This curriculum guide provides the plant science components of the vocational agriculture curriculum for Regional Vocational Agriculture Centers. The curriculum is divided into exploratory units for students in the 9th and 10th grades and specialized units for students in grades 11 and 12. The five exploratory units are: agricultural pest control;…

  18. Thermal damage of power plants components and their reparation. Aspects of welding engineering

    International Nuclear Information System (INIS)

    Kautz, H.R.; Zurn, H.E.D.

    1993-01-01

    In the last years, the technology of power plants has been developed. With the recommendation in environmental protection, the research is focussed on gaseous effluents purification . In case of were an accident, the welding engineering might repair the components. 47 refs

  19. 75 FR 65514 - Automotive Components Holdings, LLC, A Subsidiary of Ford Motor Company, Saline Plant Division...

    Science.gov (United States)

    2010-10-25

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-72,029] Automotive Components Holdings, LLC, A Subsidiary of Ford Motor Company, Saline Plant Division, Saline, MI; Notice of Affirmative Determination Regarding Application for Reconsideration By application sent to this office on April 8, 2010, the...

  20. Current advances in screening for bioactive components from medicinal plants by affinity ultrafiltration mass spectrometry.

    Science.gov (United States)

    Chen, Guilin; Huang, Bill X; Guo, Mingquan

    2018-05-21

    Medicinal plants have played an important role in maintaining human health for thousands of years. However, the interactions between the active components in medicinal plants and some certain biological targets during a disease are still unclear in most cases. To conduct the high-throughput screening for small active molecules that can interact with biological targets, which is of great theoretical significance and practical value. The ultrafiltration mass spectrometry (UF-LC/MS) is a powerful bio-analytical method by combining affinity ultrafiltration and liquid chromatography-mass spectrometry (LC/MS), which could rapidly screen and identify small active molecules that bind to biological targets of interest at the same time. Compared with other analytical methods, affinity UF-LC/MS has the characteristics of fast, sensitive and high throughput, and is especially suitable for the complicated extracts of medicinal plants. In this review, the basic principle, characteristics and some most recent challenges in UF-LC/MS have been demonstrated. Meanwhile, the progress and applications of affinity UF-LC/MS in the discovery of the active components from natural medicinal plants and the interactions between small molecules and biological target proteins are also briefly summarised. In addition, the future directions for UF-LC/MS are also prospected. Affinity UF-LC/MS is a powerful tool in studies on the interactions between small active molecules and biological protein targets, especially in the high-throughput screening of active components from the natural medicinal plants. Copyright © 2018 John Wiley & Sons, Ltd.

  1. Resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1990-06-01

    This report describes the US Nuclear Regulatory Commission's (NRC's) Generic Safety Issue 29, ''Bolting Degradation or Failure in Nuclear Power Plants,'' including the bases for establishing the issue and its historical highlights. The report also describes the activities of the Atomic Industrial Forum (AIF) relevant to this issue, including its cooperation with the Materials Properties Council (MPC) to organize a task group to help resolve the issue. The Electric Power Research Institute, supported by the AIF/MPC task group, prepared and issued a two-volume document that provides, in part, the technical basis for resolving Generic Safety Issue 29. This report presents the NRC's review and evaluation of the two-volume document and NRC's conclusion that this document, in conjunction with other information from both industry and NRC, provides the bases for resolving this issue

  2. 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

  3. Relative planting times on the production components in sesame/cowpea bean intercropping in organic system

    Directory of Open Access Journals (Sweden)

    Afrânio César de Araújo

    2013-12-01

    Full Text Available Aiming at better land use, small farmers usually plant sesame and cowpea bean intercropped with other crops. The aim of this work was to analyze and quantify the influence of four relative planting times of the cowpea bean in intercropping with sesame from the standpoint of their production components, plant productivity and the index of land equivalent ratio (LER. The field experiment was conducted in a randomized blocks with four treatments and four replicates. The treatments were the sesame and the cowpea bean in intercropping with the cowpea bean planted at the same time, 7, 14 and 21days after than the sesame. A greater part of the production components of both the sesame as well the cowpea bean was affected by the intercropping and significant differences were noted among the treatments in a larger part of the parameters. As the planting of the cowpea bean became more distant from that of the sesame, the yield of the Pedaliaceae increased and the yield of the Fabaceae decreased. The results for LER findings on the other hand suggest that in the sesame/cowpea bean intercropping, when the Fabaceae is planted seven days after the sesame, there is better use of the land and a largest possibility to the producer earning a profit.

  4. Affinity and selectivity of plant proteins for red wine components relevant to color and aroma traits.

    Science.gov (United States)

    Granato, Tiziana Mariarita; Ferranti, Pasquale; Iametti, Stefania; Bonomi, Francesco

    2018-08-01

    The effects of fining with various plant proteins were assessed on Aglianico red wine, using both the young wine and wine aged for twelve and twenty-four months, and including wine unfined or fined with gelatin as controls. Color traits and fining efficiency were considered, along with the content of various types of phenolics and of aroma-related compounds of either varietal or fermentative origin. All agents had comparable fining efficiency, although with distinct kinetics, and had similar effects on wine color. Individual plant proteins and enzymatic hydrolyzates differed in their ability to interact with some anthocyanins, with specific proanthocyanidins complexes, and with some aroma components of fermentative origin. Changes in varietal aroma components upon fining were very limited or absent. Effects of all the fining agents tested in this study on the anthocyanidin components were most noticeable in young red wine, and decreased markedly with increasing wine ageing. Copyright © 2018 Elsevier Ltd. All rights reserved.

  5. 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.

  6. Quality assurance in the planning and construction of components for nuclear power plants and large chemical plants

    International Nuclear Information System (INIS)

    Doerling

    1975-01-01

    High safety technical requirements must be demanded of the components of these plants to avoid economical hazards and to protect life and health. These requirements necessitate that each phase of the task completion, i.e. in planning, construction, fabrication and assembly, be carried out systematically and totally in order to produce a component with optimum quality. Quality assurance cannot then merely be a quality control in a conventional sense carried out during fabrication. It is much more an aimed procedure which is oriented to the functional requirements of the components - or rather to the function carrier. The concept presented on the quality assurance gives me the right as a constructor to treat this subject. (orig./LH) [de

  7. 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

  8. Conceptual benefits of passive nuclear power plants and their effect on component design

    International Nuclear Information System (INIS)

    DeVine, J.C. Jr.

    1996-01-01

    Today, nearly ten years after the advanced light water reactor (ALWR) Program was conceived by US utility leaders, and a decade and a half since a new nuclear power plant was ordered in the US, the ALWR passive plant is coming into its own. This design concept, a midsized simplified light water reactor, features extremely reliable passive systems for accident prevention and mitigation and combines proven experience with state-of-the-art engineering and human factors. It is now emerging as the front runner to become the next generation reactor in the US and perhaps around the world. Although simple and straightforward in concept, the passive plant is in many respects a significant departure from previous trends in reactor engineering. Successful implementation of this concept presents numerous challenges to the designers of passive plant systems and components. This paper provides a brief history of the ALWR program, it outlines the ALWR passive plant design objectives and principles, and it summarizes with examples their implications on component design. (orig.)

  9. Critical components of odors in evaluating the performance of food waste composting plants

    International Nuclear Information System (INIS)

    Mao, I-F.; Tsai, C.-J.; Shen, S.-H.; Lin, T.-F.; Chen, W.-K.; Chen, M.-L.

    2006-01-01

    The current Taiwan government policy toward food waste management encourages composting for resource recovery. This study used olfactometry, gas chromatography-mass spectrometry (GC-MS) and gas detector tubes to evaluate the ambient air at three of the largest food waste composting plants in Taiwan. Ambient air inside the plants, at exhaust outlets and plant boundaries was examined to determine the comprehensive odor performance, critical components, and odor elimination efficiencies of various odor control engineering. Analytical results identified 29 compounds, including ammonia, amines, acetic acid, and multiple volatile organic compounds (VOCs) (hydrocarbons, ketones, esters, terpenes and S-compounds) in the odor from food waste composting plants. Concentrations of six components - ammonia, amines, dimethyl sulfide, acetic acid, ethyl benzene and p-Cymene - exceeded human olfactory thresholds. Ammonia, amines, dimethyl sulfide and acetic acid accounted for most odors compared to numerous VOCs. The results also show that the biotrickling filter was better at eliminating the concentrations of odor, NH 3 , amines, S-compounds and VOCs than the chemical scrubber and biofilters. All levels measured by olfactometry at the boundaries of food waste composting plants (range, 74-115 Odor Concentration (OC)) exceeded Taiwan's EPA standard of 50 OC. This study indicated that the malodor problem continued to be a significant problem for food waste recovery

  10. Critical components of odors in evaluating the performance of food waste composting plants

    Energy Technology Data Exchange (ETDEWEB)

    Mao, I-F. [Institute of Environmental Health Sciences, National Yang-Ming University, No. 155, Sec.2, Li-Nong St., Beitou, Taipei, Taiwan (China)]. E-mail: ifmao@ym.edu.tw; Tsai, C.-J. [Institute of Environmental Health Sciences, National Yang-Ming University, No. 155, Sec.2, Li-Nong St., Beitou, Taipei, Taiwan (China); Shen, S.-H. [Department of Environment Management, Jin Wen Institute of Technology, No. 99, An-Chung Rd., Hsin-Tien City, Taipei, Taiwan (China); Lin, T.-F. [Institute of Environmental Engineering, National Cheng Kung University, No. 1, Ta-Hsueh Rd., Tainan, Taiwan (China); Chen, W.-K. [Department of Environment Management, Jin Wen Institute of Technology, No. 99, An-Chung Rd., Hsin-Tien City, Taipei, Taiwan (China); Chen, M.-L. [Institute of Environmental Health Sciences, National Yang-Ming University, No. 155, Sec.2, Li-Nong St., Beitou, Taipei, Taiwan (China)]. E-mail: mlchen@ym.edu.tw

    2006-11-01

    The current Taiwan government policy toward food waste management encourages composting for resource recovery. This study used olfactometry, gas chromatography-mass spectrometry (GC-MS) and gas detector tubes to evaluate the ambient air at three of the largest food waste composting plants in Taiwan. Ambient air inside the plants, at exhaust outlets and plant boundaries was examined to determine the comprehensive odor performance, critical components, and odor elimination efficiencies of various odor control engineering. Analytical results identified 29 compounds, including ammonia, amines, acetic acid, and multiple volatile organic compounds (VOCs) (hydrocarbons, ketones, esters, terpenes and S-compounds) in the odor from food waste composting plants. Concentrations of six components - ammonia, amines, dimethyl sulfide, acetic acid, ethyl benzene and p-Cymene - exceeded human olfactory thresholds. Ammonia, amines, dimethyl sulfide and acetic acid accounted for most odors compared to numerous VOCs. The results also show that the biotrickling filter was better at eliminating the concentrations of odor, NH{sub 3}, amines, S-compounds and VOCs than the chemical scrubber and biofilters. All levels measured by olfactometry at the boundaries of food waste composting plants (range, 74-115 Odor Concentration (OC)) exceeded Taiwan's EPA standard of 50 OC. This study indicated that the malodor problem continued to be a significant problem for food waste recovery.

  11. 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

  12. Effect of Light Spectral Quality on Essential Oil Components in Ocimum Basilicum and Salvia Officinalis Plants

    Directory of Open Access Journals (Sweden)

    A. S. IVANITSKIKH

    2014-07-01

    Full Text Available In plants grown with artificial lighting, variations in light spectral composition can be used for the directed biosynthesis of the target substances including essential oils, e.g. in plant factories. We studied the effect of light spectral quality on the essential oil composition in Ocimum basilicum and Salvia officinalis plants grown in controlled environment. The variable-spectrum light modules were designed using three types of high-power light-emitting diodes (LEDs with emission peaked in red, blue and red light, white LEDs, and high-pressure sodium lamps as reference. Qualitative and quantitative essential oil determinations were conducted using gas chromatography with mass selective detection and internal standard method.Sweet basil plant leaves contain essential oils (са. 1 % including linalool, pinene, eugenol, camphor, cineole, and other components. And within the genetic diversity of the species, several cultivar groups can be identified according to the flavor (aroma perceived by humans: eugenol, clove, camphor, vanilla basil. Essential oil components produce particular flavor of the basil leaves. In our studies, we are using two sweet basil varieties differing in the essential oil qualitative composition – “Johnsons Dwarf” (camphor as a major component of essential oils and “Johnsons Lemon Flavor” (contains large amount of citral defining its lemon flavor.In sage, essential oil composition is also very variable. As for the plant responses to the light environment, the highest amount of the essential oils was observed at the regimes with white and red + blue LED light. And it was three times less with red light LEDs alone. In the first two environments, thujone accumulation was higher in comparison with camphor, while red LED light and sodium lamp light favored camphor biosynthesis (three times more than thujone. The highest amount of eucalyptol was determined in plants grown with red LEDs.

  13. Pilot studies of management of ageing of nuclear power plant instrumentation and control components

    International Nuclear Information System (INIS)

    Burnay, S.G.; Simola, K.; Kossilov, A.; Pachner, J.

    1993-01-01

    This paper describes pilot studies which have been implemented to study the aging behavior of safety related component parts of nuclear power plants. In 1989 the IAEA initiated work on pilot studies related to the aging of such components. Four components were identified for study. They are the primary nozzle of a reactor vessel; a motor operated isolating valve; the concrete containment building; and instrumentation and control cables within the containment facility. The study was begun with phase 1 efforts directed toward understanding the aging process, and methods for monitoring and minimizing the effects of aging. Phase 2 efforts are directed toward aging studies, documentation of the ideas put forward, and research to answer questions identified in phase 1. This paper describes progress made on two of these components, namely the motor operated isolation valves, and in-containment I ampersand C cables

  14. [Epidemiology of contact hypersensitivity to rubber components in manufacturers of automobile tires at the Stomil plant].

    Science.gov (United States)

    Rubisz-Brzezińska, J; Bogdanowski, T; Brzezińska-Wcisło, L; Mozdzanowska, K; Bajcar, S

    1990-01-01

    Dermatological examination and patch tests with 34 rubber components were carried out in 114 tire manufacturers, 78 women and 36 men aged 29 years on average, with a mean duration of work in the plant 7 years. For correct interpretation of the obtained results patch tests with the same components were done in two control groups that is in 120 healthy subjects and 120 patients with contact dermatitis. Patch tests with proper concentrations of the studied components were evaluated after 48, 72 and 96 hours. Positive patch tests were found most frequently with antioxidants--16.6% (including IPPD--8.6%), followed by vulcanization accelerators--10.6%, and other rubber components--11.4% in all. During about 3 years of follow-up in 4 manufacturers contact allergic eczema was noted and polyvalent hypersensitivity to antioxidants and vulcanization accelerators without clinical manifestations of this hypersensitivity was diagnosed in 3 other subjects.

  15. Application of PHADEC method for the decontamination of radioactive steam piping components of Caorso plant

    International Nuclear Information System (INIS)

    Lo Frano, R.; Aquaro, D.; Fontani, E.; Pilo, F.

    2014-01-01

    Highlights: • Application of PHADEC chemical off-line methodology. • Decontamination of radioactive steam piping components of Caorso turbine building. • Experimental characterization of metallic components, e.g., by SEM analysis. • Measure of the efficiency of treatment by means of the reduction of activity and vs. the treatment time. • Minimization of secondary waste produced during decontamination activity of Caorso BWR plant. - Abstract: The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e., taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the

  16. Ageing study of protection automation components of Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Simola, K.; Haenninen, S.

    1993-07-01

    A study on ageing of reactor protection system of the Olkiluoto nuclear power plant is described. The objective of the study was to present an ageing analysis approach and apply in to the automation chains of reactor protection system of the Olkiluoto nuclear power plant. The study includes the measuring instrumentation, the protection logics, and the control electronics of some pumps and valves. The analysis is based on the information collected on the structure of the system, environmental conditions and maintenance practices of components, and operating experience. Based on this information, the possible ageing effects of equipment and their safety significance are evaluated. (orig.). (15 refs., 16 figs., 12 tabs.)

  17. Fatigue evaluation including environmental effects for primary circuit components in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Seichter, Johannes [Siempelkamp Pruef- und Gutachter-Gesellschaft mbH, Dresden (Germany); Reese, Sven H.; Klucke, Dietmar [E.ON Kernkraft GmbH, Hannover (Germany). Component Technology

    2013-06-01

    The influence of LWR coolant environment to the lifetime of materials in nuclear power plants is being discussed internationally. Environmental phenomena had been investigated in laboratory tests and published in recent years. The discussion is mainly focused both on the transition from laboratory to real plant components and on numerical calculation procedures. Since publishing of the NUREG/CR-6909 report in 2007, formulae for calculating the Fen factors have been modified several times. Various calculation procedures are discussed and recommendations are made how to avoid extremely conservative results. (orig.)

  18. Simulation of an industrial wastewater treatment plant using artificial neural networks and principal components analysis

    Directory of Open Access Journals (Sweden)

    Oliveira-Esquerre K.P.

    2002-01-01

    Full Text Available This work presents a way to predict the biochemical oxygen demand (BOD of the output stream of the biological wastewater treatment plant at RIPASA S/A Celulose e Papel, one of the major pulp and paper plants in Brazil. The best prediction performance is achieved when the data are preprocessed using principal components analysis (PCA before they are fed to a backpropagated neural network. The influence of input variables is analyzed and satisfactory prediction results are obtained for an optimized situation.

  19. Identification of seismically risk-sensitive systems and components in nuclear power plants: feasibility study

    International Nuclear Information System (INIS)

    Azarm, M.; Boccio, J.; Farahzad, P.

    1983-06-01

    An approach for the identification of risk-sensitive components in a nuclear power plant during and after a seismic event is described. Application of the methodology to two hypothetical power plants - a Boiling Water Reactor and a Pressurized Water Reactor - are presented and the results are given in tabular and graphical form. Conclusions drawn and lessons learned through the course of this study, based on the relative importance of various accident scenarios and sensitivity analyses, are discussed. In addition, the areas that may need further investigation are identified

  20. Precision Diagnosis, Monitoring and Control of Structural Component Degradation in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Han, J. H.; Choi, M. S.; Lee, D. H.; Hur, D. H.; Na, J. W.; Kim, K. M.; Hong, J. H.; Kim, H. S.

    2007-06-01

    The occurrence of structural material degradations in NPPs and their progress during operation are directly related to the safety and the integrity of NPPs. The various kinds of material degradation are usually examined by methods of material integrity evaluation and non-destructive evaluation(NDE). Material integrity evaluation is well known as classical method to interpret cause and mechanism of degradation and failure, however, this method has a limitation of detection and diagnosis for actual condition of flaws and defects occurring during plant operation, particularly for their formation in the early stage. NDE used widely for detection of defects formed on structural materials provides many information for safety regulation, plant management, repairing, however, this technique has a generic problem in its reliability due to low detectability and ability of signal analysis, etc. The objective of this research project is to develop the advanced technologies ensuring a precision diagnosis on the various kind of defects in structural materials of NPP and a high performance in material degradation evaluation. Many of the advanced technologies were developed in the 1st phase of this project. They contributed to interpret more precisely the root causes of degradation, failure and to establish the proper measures for the safety and integrity of NPPs. The accomplishment of comprehensive technology developed as planned will be practically applied to the nuclear industries and contributed to improve the safety and integrity of NPPs

  1. Maintenance Management Support Systems for component aging estimation at nuclear power plants

    International Nuclear Information System (INIS)

    Shimizu, Shunichi; Ando, Yasumasa; Morioka, Toshihiko; Okuzumi, Naoaki

    1991-01-01

    Maintenance Management Support Systems (MMSSs) for nuclear power plants have been developed using component aging estimation methods and decision tree analysis for maintenance planning. The former evaluates actual component reliability through statistical analysis on field maintenance data. The latter provides preventive maintenance (PM) planning guidance using heuristic expert knowledge and estimated reliability parameters. The following aspects have been investigated: (1) A systematic and effective method of managing components/parts design information and field maintenance data (2) A method for estimating component aging based on a statistical analysis of field maintenance data (3) A method for providing PM planning guidance using estimated component reliability/performance parameters and decision tree analysis. Based on these investigations, two MMSSs were developed. One deals with 'general maintenance data', which are common to all component types and are amenable to common data handling. The other system deals with 'specific maintenance data', which are specific to an individual component type. Both systems provide PM planning guidance for PM cycles propriety and the PM work priority. The function of these systems were verified using simulated maintenance data. (author)

  2. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  3. A study on the crack inspection signal characteristics for power plant components by phased array UT

    International Nuclear Information System (INIS)

    Cho, Yong Sang; Lim, Sang Gyu; Kil, Du Song

    2001-01-01

    Phased array ultrasonic testing system has become available for practical application in complicated geometry such as turbine blade root, tenon, disc in power industry. This research describes the characteristics of phased array UT signal for various type of blade roots in thermal Power Plant turbines. This application of Phased array ultrasonic testing system has been promoted mainly to save inspection time and labor cost of turbine inspection. The characteristic of phase array UT signal for power plant component is very simple to understand but to difficult for perform the inspection. Since our sophisticated inspection technique and systems are essential for the inspection of steam turbine blade roots that require high reliability, we intend to develop new technology and improve phased array technique based on the wide and much experience for the inspection of turbine components.

  4. Pilot studies on management of ageing of nuclear power plant components: Results of Phase 1

    International Nuclear Information System (INIS)

    1992-10-01

    To facilitate cooperation between the IAEA Member States and thus to enhance the safety and reliability of operating nuclear plants the IAEA has initiated pilot studies on the management of ageing of four representative plant components: the primary nozzle of the reactor pressure vessel, a motor operated valve, the concrete containment building and instrumentation and control cables. Phase 1 of the studies has been completed and its results are presented in this report. The report documents current understanding of ageing and methods for monitoring and mitigation of this ageing for the above components, identifies existing knowledge and technology gaps and defines follow-up work to deal with these gaps. Refs, figs and tabs

  5. Aging of concrete components and its significance relative to life extension of nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.

    1987-01-01

    Nuclear power currently supplies about 16% of the US electricity requirements, with the percentage expected to rise to 20% by 1990. Despite the increasing role of nuclear power in energy production, cessation of orders for new nuclear plants in combination with expiration of operating licenses for several plants in the next 15 to 20 years results in a potential loss of electrical generating capacity of 50 to 60 gigawatts during the time period 2005 to 2020. A potential timely and cost-effective solution to the problem of meeting future energy demand is available through extension of the service life of existing nuclear plants. Any consideration of plant life extension, however, must consider the concrete components in these plants, since they play a vital safety role. Under the USNRC Nuclear Plant Aging Research (NPAR) Program, a study was conducted to review operating experience and to provide background that will lead to subsequent development of a methodology for assessing and predicting the effects of aging on the performance of concrete-based structures. The approach followed was in conformance with the NPAR strategy

  6. Friction and wear studies of nuclear power plant components in pressurized high temperature water environments

    International Nuclear Information System (INIS)

    Ko, P.L.; Zbinden, M.; Taponat, M.C.; Robertson, M.F.

    1997-01-01

    The present paper is part of a series of papers aiming to present the friction and wear results of a collaborative study on nuclear power plant components tested in pressurized high temperature water. The high temperature test facilities and the methodology in presenting the kinetics and wear results are described in detail. The results of the same material combinations obtained from two very different high temperature test facilities (NRCC and EDF) are presented and discussed. (K.A.)

  7. Operating experience in cleaning sodium-wetted components at the KNK nuclear power plant

    International Nuclear Information System (INIS)

    Stade, K.Ch.

    1978-01-01

    Since 1969, components of the KNK facility, the first sodium cooled nuclear power plant in the Federal Republic of Germany, have been cleaned both by the alcohol and the wet gas techniques. This paper outlines the experience accumulated In the application of these methods, especially in cleaning steam generators and fuel elements. Some preliminary results are indicated of the attempt to clean a cold trap from the primary circuit of the KNK facility. (author)

  8. Experience with nonuniform damping in the seismic analysis of nuclear plant components

    International Nuclear Information System (INIS)

    Winkel, B.V.; Julyk, L.J.

    1983-01-01

    Individual components of nuclear power plants may exhibit pronounced differences in damping magnitude. Various methods for accounting for nonuniform damping in a structural model are reviewed and evaluated. The methods are compared by solving a beam/pipe model subjected to a typical seismic ground motion. A two-degree-of-freedom variable damping parameter study is also presented. Based upon the experience of evaluating and applying the available methods, application guidelines are presented

  9. GERB viscous dampers in applications for pipelines and other components in Czechoslovak nuclear power plants

    International Nuclear Information System (INIS)

    Masopust, R.; Podrouzek, J.

    1992-01-01

    VISCODAMPERS from GERB, Germany, are now widely used as reliable shock restraints against earthquake and other shock effects for the most important safety-related pipelines and components in several Czechoslovak nuclear power plants. Having many technical advantages they are, at the same time, relatively inexpensive in comparison with conventional snubbers. Their properties are briefly described and several practical applications are explained. (author) 3 tabs., 9 figs., 8 refs

  10. GERB viscous dampers in application for pipelines and other components in nuclear power plants

    International Nuclear Information System (INIS)

    Masopust, R.; Podrouzek, J.; Zach, J.

    1993-01-01

    VISCODAMPERS from GERB, Germany, are now widely used as reliable shock restraints against earthquake and other shock effects for the most important safety-related pipelines and components in several Czech and Slovak nuclear power plants. Having many technical advantages they are, at the same time, relatively inexpensive in comparison to conventionally used snubbers. Their properties are briefly described and several practical applications are explained in this paper. (author)

  11. Effects of Plant Density on Sweet and Baby Corn (Hybrid KSC 403 Yield and Yield Components

    Directory of Open Access Journals (Sweden)

    H Bavi

    2016-07-01

    GlM procedure. Means of all treatments were comprised using least significant difference (LSD at 5 % probability level. Results and Discussion The effects of plant density on yield components of baby corn was significant. Increasing the plant densities increased the ear number and percentage of non-standard ears. The Highest yield of ear without husk, standard and non-standard were obtained (2649.5, 766.97, and 3043.9 kg.ha-1, respectively with 13 plants.m-2. In sweet corn, increasing plant density from 7 to 13 plants.m-2, decreased row per ear, grain per row and thousand grain weight. Highest grain yield (1232.5 kg ha-1 and green ear (12607.2 kg ha-1 of sweet corn were obtained with plant density of 9.m-2. Conclusions Analysis of correlation showed that in both baby and sweet corn, there were positive and significant correlations between yield and its components. There was the high number of non-standard ears in all experimental treatments. In sweet corn, the standard ear without husk yield has positive and significant correlation with all traits except the percentage of standard ear and sheathed ear weight. In addition, unsuitable climate conditions during silking stage reduced the yield of sweet corn through the high number of aborted florets. Yield of sweet corn yield showed negative and significant correlation with grain row per ear and grain per row. However, increasing the ear number.m-2 increased yield in higher plant densities up to 9 plant.m-2 density. Generally, the baby corn had high yield with good quality in this region, but, standard ear percent of the baby corn of the hybrid KSC 403 was very low. On the other hand, sweet corn grain yield was low due to high air temperatures during pollination and maturity stages.

  12. Lifetime assessment and lifetime management for key components of nuclear power plants

    International Nuclear Information System (INIS)

    Dou Yikang; Sun Hanhong; Qu Jiadi

    2000-01-01

    On the bases of investigation on recent development of plant lifetime management in the world, the author gives some points of view on how to establish plant lifetime assessment (PLA) and management (PLM) systems for Chinese nuclear power plants. The main points lie in: 1) safety regulatory organizations, utilities and R and D institutes work cooperatively for PLA and PLM; 2) PLA and PLM make a interdependent cycle, which means that a good PLM system ensures authentic input for PLA, while veritable PLA provides valuable feedback for PLM improvement; 3) PLA and PLM should be initiated for some key components. The author also analyzes some important problems to be tackled in PLA and PLM from the view angle of a R and D institute

  13. Termitarium-Inhabiting Bacillus spp. Enhanced Plant Growth and Bioactive Component in Turmeric (Curcuma longa L.).

    Science.gov (United States)

    Chauhan, Ankit Kumar; Maheshwari, Dinesh Kumar; Dheeman, Shrivardhan; Bajpai, Vivek K

    2017-02-01

    Curcumin (diferuloyl methane) is the main bioactive component of turmeric (Curcuma longa L.) having remarkable multipotent medicinal and therapeutic applications. Two Bacilli isolated from termitarium soil and identified as Bacillus endophyticus TSH42 and Bacillus cereus TSH77 were used for bacterization of rhizome for raising C. longa ver. suguna for growth and enhancement. Both the strains showed remarkable PGP activities and also chemotactic in nature with high chemotactic index. Turmeric plants bacterized with strains B. endophyticus TSH42 and B. cereus TSH77 individually and in combination increased plant growth and turmeric production up to 18% in field trial in comparison to non-bacterized plants. High-performance liquid chromatography analysis was performed to determine the content of curcumin, which showed concentration of curcumin in un-inoculated turmeric as 3.66 g which increased by 13.6% (4.16 g) when combination of TSH42 and TSH77 was used.

  14. Potential of plant growth regulator and chlormequat chloride on alfalfa seed components

    International Nuclear Information System (INIS)

    Chen, J. S.; Lin, H.; Han, W.

    2016-01-01

    The use of plant growth regulators (PGRs) has opened new prospects for increased seed production in grasses and legumes, but little information is available on the effects of PGRs combination with chlormequat chloride (CCC) on alfalfa (Medicago sativa L.) seed yield components. This study was conducted to evaluate the effects of applying chlormequat chloride in combination with three PGRs (Naphthylacetic acid (NAA), gibberellic acid 3 (GA), and brassinolide (BR)) on seed yield, aboveground biomass, plant height, lodging, yield components. CCC was applied annually at the stooling stage while three PGRs were applied twice each year at the stages of flower bud formation and peak flowering. Results provides evidence that: (i) each PGR consistently increased seed yields, and the numbers of seeds per stem compared to untreated plants; (ii) CCC treatment reduced plant height and lodging, but also significantly decreased seed yield and did not affect aboveground biomass. (iii) effectiveness of CCC application depends on climatic conditions, especially in North-east China. (iiii) the optimum combination of CCC with a PGR to increase alfalfa seed production was failed to identify. (iiiii) no interactions between PGRs and CCC on seed yield were observed and neither the PGRs nor the CCC. But alfalfa seed yield could be improved by combining a PGR such as NAA. Our Results suggest that these PGRs could be used in alfalfa breeding to increase seed yield while maintaining high seed quality. (author)

  15. 10 CFR Appendix B to Part 110 - Illustrative List of Gas Centrifuge Enrichment Plant Components Under NRC's Export Licensing...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Gas Centrifuge Enrichment Plant... 110—Illustrative List of Gas Centrifuge Enrichment Plant Components Under NRC's Export Licensing Authority 1. Assemblies and components especially designed or prepared for use in gas centrifuges. Note: The...

  16. Post-translational regulation of miRNA pathway components, AGO1 and HYL1, in plants

    DEFF Research Database (Denmark)

    Cho, Seok Keun; Ryu, Moon Young; Shah, Pratik

    2016-01-01

    , the complexity of the proteome increases, and this then influences most biological processes. Although small RNAs are crucial regulatory elements for gene expression in most eukaryotes, PTMs of small RNA microprocessor and RNA silencing components have not been extensively investigated in plants. To date...... findings on the PTMs of microprocessor and RNA silencing components in plants....

  17. Concrete component aging and its significance relative to life extension of nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.

    1986-09-01

    The objectives of this study are to (1) expand upon the work that was initiated in the first two Electric Power Research Institute studies relative to longevity and life extension considerations of safety-related concrete components in light-water reactor (LWR) facilities and (2) provide background that will logically lead to subsequent development of a methodology for assessing and predicting the effects of aging on the performance of concrete-based materials and components. These objectives are consistent with Nuclear Plant Aging Research (NPAR) Program goals: (1) to identify and characterize aging and service wear effects that, if unchecked, could cause degradation of structures, components, and systems and, thereby, impair plant safety; (2) to identify methods of inspection, surveillance, and monitoring or of evaluating residual life of structures, components, and systems that will ensure timely detection of significant aging effects before 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

  18. Inservice inspection of heavy water plants - a tool in assessing damage to components and life extension

    International Nuclear Information System (INIS)

    Subramanian, C.V.; Thavasimuthu, M.; Bhattacharys, D.K.; Baldev Raj

    1994-01-01

    Any system and its components are expected to give trouble free service over a certain period of time known as life time. The life time is estimated during the design stage. To achieve the design life, certain level of quality are to be defined and this quality has to be worked into the components by proper fabrication processes and their compliance with quality are to be checked. In addition, one has to guard against initiation or propagation of defects which may occur due to normal and abnormal service conditions. Non-destructive test (NDT) techniques are widely used for finding the health of the component. The role of NDT extends from the production stage to the entire life period of the system. This paper highlights the periodic in-service inspection (ISI) carried out on various components of the Heavy Water Plants (HWP) in India in assessing the integrity of the components and predicting the life of the components. (author). 3 refs., 4 figs

  19. Risk assessment of component failure modes and human errors using a new FMECA approach: application in the safety analysis of HDR brachytherapy

    International Nuclear Information System (INIS)

    Giardina, M; Castiglia, F; Tomarchio, E

    2014-01-01

    Failure mode, effects and criticality analysis (FMECA) is a safety technique extensively used in many different industrial fields to identify and prevent potential failures. In the application of traditional FMECA, the risk priority number (RPN) is determined to rank the failure modes; however, the method has been criticised for having several weaknesses. Moreover, it is unable to adequately deal with human errors or negligence. In this paper, a new versatile fuzzy rule-based assessment model is proposed to evaluate the RPN index to rank both component failure and human error. The proposed methodology is applied to potential radiological over-exposure of patients during high-dose-rate brachytherapy treatments. The critical analysis of the results can provide recommendations and suggestions regarding safety provisions for the equipment and procedures required to reduce the occurrence of accidental events. (paper)

  20. Evaluation of the Waste Isolation Pilot Plant classification of systems, structures and components

    International Nuclear Information System (INIS)

    1985-07-01

    A review of the classification system for systems, structures, and components at the Waste Isolation Pilot Plant (WIPP) was performed using the WIPP Safety Analysis Report (SAR) and Bechtel document D-76-D-03 as primary source documents. The regulations of the US Nuclear Regulatory Commission (NRC) covering ''Disposal of High level Radioactive Wastes in Geologic Repositories,'' 10 CFR 60, and the regulations relevant to nuclear power plant siting and construction (10 CFR 50, 51, 100) were used as standards to evaluate the WIPP design classification system, although it is recognized that the US Department of Energy (DOE) is not required to comply with these NRC regulations in the design and construction of WIPP. The DOE General Design Criteria Manual (DOE Order 6430.1) and the Safety Analysis and Review System for AL Operation document (AL 54f81.1A) were reviewed in part. This report includes a discussion of the historical basis for nuclear power plant requirements, a review of WIPP and nuclear power plant classification bases, and a comparison of the codes and standards applicable to each quality level. Observations made during the review of the WIPP SAR are noted in the text of this reoport. The conclusions reached by this review are: WIPP classification methodology is comparable to corresponding nuclear power procedures. The classification levels assigned to WIPP systems are qualitatively the same as those assigned to nuclear power plant systems

  1. Controversy Associated With the Common Component of Most Transgenic Plants – Kanamycin Resistance Marker Gene

    Directory of Open Access Journals (Sweden)

    Srećko Jelenić

    2003-01-01

    Full Text Available Plant genetic engineering is a powerful tool for producing crops resistant to pests, diseases and abiotic stress or crops with improved nutritional value or better quality products. Currently over 70 genetically modified (GM crops have been approved for use in different countries. These cover a wide range of plant species with significant number of different modified traits. However, beside the technology used for their improvement, the common component of most GM crops is the neomycin phosphotransferase II gene (nptII, which confers resistance to the antibiotics kanamycin and neomycin. The nptII gene is present in GM crops as a marker gene to select transformed plant cells during the first steps of the transformation process. The use of antibiotic-resistance genes is subject to controversy and intense debate, because of the likelihood that clinical therapy could be compromised due to inactivation of the oral dose of the antibiotic from consumption of food derived from the transgenic plant, and because of the risk of gene transfer from plants to gut and soil microorganisms or to consumer’s cells. The present article discusses these possibilities in the light of current scientific knowledge.

  2. AVISE, ageing anticipation methodology using expert judgement and stimulation. Application to a nuclear power plant component: the pressurizer

    International Nuclear Information System (INIS)

    Bouzaiene-Marle, L.

    2005-04-01

    This thesis deals with components ageing anticipation in the context of life cycle management. The proposed approach, called AVISE, allows the identification of potentials problems related to ageing, to measure the risks in terms of degradation probability and degradation consequences and gives the adequate solutions to stop or to postpone ageing. This research was undertaken in a particular industrial context, the nuclear industry. Equipments used in this context are specific and particularly reliable. These characteristics result in limited feedback (low number of failures). To compensate for this limited information, two solutions are proposed in this approach. The first solution that we can consider as a classical one consists in using expert judgement. The second one, more original, consists in using the operation feedback of 'similar' components. In order to apply these solutions and to obtain the anticipation results, a set of methodological tools was developed and tested in a real industrial application on a nuclear power plant component: the pressurizer. The first tool is a generic process for expert judgement, identified thanks to a comparison between eleven existing methods using expert judgement. Two methods based on expert stimulation and called STIMEX-IMDP and STIMEX-IPP were elaborated. A reference list of degradation mechanisms and a reference list of ageing effects were constructed and used in the method STIMEX-IMDP in order to help expert stimulation. Then, the developed approach proposes the use of belief networks to model and quantify the risks related to the potential degradations. Finally, the construction of a conceptual data model and specifications are given for the creation of an ageing database. The data to capitalize was identified on the basis of the research undertaken in this thesis. (author)

  3. Effect of Plant Density and Weed Interference on Yield and Yied Components of Grain Sorghum

    Directory of Open Access Journals (Sweden)

    S. Sarani

    2018-01-01

    Full Text Available Introduction Weed control is an essential part of all crop production systems. Weeds reduce yields by competing with crops for water, nutrients, and sunlight. Weeds also directly reduce profits by hindering harvest operations, lowering crop quality, and producing chemicals which are harmful to crop plants. Plant density is an efficient management tool for maximizing grain yield by increasing the capture of solar radiation within the canopy, which can significantly affect development of crop-weed association. The response of yield and yield components to weed competition varies by crop and weeds species and weeds interference duration. The objective of the present study was to evaluate the effect of weed interference periods and plant density on the yield and yield components of sorghum. Materials and Methods In order to study the effect of plant density and weeds interference on weeds traits, yield and yield components of sorghum (Var. Saravan, an experiment was conducted as in factorial based on a randomized complete block design with three replications at the research field of Islamic Azad University, Birjand Branch in South Khorasan province during year of 2013. Experimental treatments consisted of three plant density (10, 20 and 30 plants m-2 and four weeds interference (weed free until end of growth season, interference until 6-8 leaf stage, interference until stage of panicle emergence, interference until end of growth season. Measuring traits included the panicle length, number of panicle per plant, number of panicle per m2, number of seed per panicle, 1000-seed weight, seed yield, biological yield, number and weight of weeds per m2. Weed sampling in each plot have done manually from a square meter and different weed species counted and oven dried at 72 °C for 48 hours. MSTAT-C statistical software used for data analysis and means compared with Duncan multiple range test at 5% probability level. Results and Discussion Results showed that

  4. Influence of shock waves as a result of assumed vessel failure on parts of the plant relevant to safety

    International Nuclear Information System (INIS)

    Danisch, R.; Graubner, U.

    1981-01-01

    The shock wave induced rupture is of subordinate importance for the laying out of the parts of the plant relevant to safety. It is covered by the precautions for maximum potential earthquakes, aircraft crashes and chemical explosions. The failure of vessels in the power house (WAZUe, SPWB) as the result of a maximum potential earthquake is extremely improbable. If a combination of the stresses resulting from maximum potential earthquakes with the hypothetical stresses resulting from vessel failure is undertaken, it can be seen that the total stresses are only increased by a minimal amount, due to the quadratic averaging of less than 3%. (orig./DG) [de

  5. Aging management and PLEX in Swiss nuclear power plants and prioritization of safety class 2 and 3 components

    International Nuclear Information System (INIS)

    Fuchs, R.; Stejskal, J.

    2000-01-01

    In this presentation ageing management of systems and components important to safety of the Swiss nuclear power plants are presented. Status of electrical components, status of mechanical components as well as status of civil structures are reviewed. The scheme of the high pressure core spray system is included

  6. Combined approach based on principal component analysis and canonical discriminant analysis for investigating hyperspectral plant response

    Directory of Open Access Journals (Sweden)

    Anna Maria Stellacci

    2012-07-01

    Full Text Available Hyperspectral (HS data represents an extremely powerful means for rapidly detecting crop stress and then aiding in the rational management of natural resources in agriculture. However, large volume of data poses a challenge for data processing and extracting crucial information. Multivariate statistical techniques can play a key role in the analysis of HS data, as they may allow to both eliminate redundant information and identify synthetic indices which maximize differences among levels of stress. In this paper we propose an integrated approach, based on the combined use of Principal Component Analysis (PCA and Canonical Discriminant Analysis (CDA, to investigate HS plant response and discriminate plant status. The approach was preliminary evaluated on a data set collected on durum wheat plants grown under different nitrogen (N stress levels. Hyperspectral measurements were performed at anthesis through a high resolution field spectroradiometer, ASD FieldSpec HandHeld, covering the 325-1075 nm region. Reflectance data were first restricted to the interval 510-1000 nm and then divided into five bands of the electromagnetic spectrum [green: 510-580 nm; yellow: 581-630 nm; red: 631-690 nm; red-edge: 705-770 nm; near-infrared (NIR: 771-1000 nm]. PCA was applied to each spectral interval. CDA was performed on the extracted components to identify the factors maximizing the differences among plants fertilised with increasing N rates. Within the intervals of green, yellow and red only the first principal component (PC had an eigenvalue greater than 1 and explained more than 95% of total variance; within the ranges of red-edge and NIR, the first two PCs had an eigenvalue higher than 1. Two canonical variables explained cumulatively more than 81% of total variance and the first was able to discriminate wheat plants differently fertilised, as confirmed also by the significant correlation with aboveground biomass and grain yield parameters. The combined

  7. Piping failures in United States nuclear power plants 1961-1995

    International Nuclear Information System (INIS)

    Bush, S.H.; Do, M.J.; Slavich, A.L.; Chockie, A.D.

    1996-01-01

    Over 1500 reported piping failures were identified and summarized based on an extensive review of tens of thousands of event reports that have been submitted to the US regulatory agencies over the last 35 years. The data base contains only piping failures; failures in vessels, pumps, valves and steam generators or any cracks that were not through-wall are not included. It was observed that there has been a marked decrease in the number of failures after 1983 for almost all sizes of pipes. This is likely due to the changes in the reporting requirements at that time and the corrective actions taken by utilities to minimize fatigue failures of small lines and IGSCC in BWRs. One failure mechanism that continues to occur is erosion-corrosion, which accounts for most of the ruptures reported and probably is responsible for the absence of downward trends in ruptures. Fatigue-vibration is also a significant contributor to piping failures. However, most of such events occur in lines approx. one inch or less in diameter. Together, erosion-corrosion and fatigue-vibration account for over 43 per cent of the failures. The overwhelming majority of failures have been leaks, over half the failures occurred in pipes with a diameter of one inch or less. Included in the report is a listing of the number of welds in various systems in LWRs

  8. 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

  9. Some experience from seismic check-ups of components of Mochovce nuclear power plant

    International Nuclear Information System (INIS)

    Masopust, R.

    1987-01-01

    The first Czechoslovak nuclear power plant with the so-called partial anti-seismic design will be built in Mochovce. The evaluation of seismic resistance is prescribed only for equipment and systems which secure the safe reactor shutdown, the withdrawal of residual heat and prevent uncontrolled release of radioactivity into the environment. The following variants were compared in the calculation analysis of the primary loop of the WWER-440 reactor for the Mochovce nuclear power plant: the seismically unsecured loop of a usual design for WWER-440 nuclear power plants, the loop provided with mechanical or hydraulic dampers and the loop provided with viscose shock absorbers. The tests showed that technically most suitable is the use of viscose shock absorbers which do not completely block the movement of the system during the earthquake but absorb it intensively. The viscose shock absorbers are also much cheaper than the dampers. Briefly described is experience with the experimental evaluation of the seismic resistance of components of the Mochovce nuclear power plant. Great difficulty was encountered by the non-existence in Czechoslovakia of a seismic table allowing simultaneous excitation in the vertical and horizontal directions. (Z.M.). 18 refs

  10. Phytophthora effector targets a novel component of small RNA pathway in plants to promote infection.

    Science.gov (United States)

    Qiao, Yongli; Shi, Jinxia; Zhai, Yi; Hou, Yingnan; Ma, Wenbo

    2015-05-05

    A broad range of parasites rely on the functions of effector proteins to subvert host immune response and facilitate disease development. The notorious Phytophthora pathogens evolved effectors with RNA silencing suppression activity to promote infection in plant hosts. Here we report that the Phytophthora Suppressor of RNA Silencing 1 (PSR1) can bind to an evolutionarily conserved nuclear protein containing the aspartate-glutamate-alanine-histidine-box RNA helicase domain in plants. This protein, designated PSR1-Interacting Protein 1 (PINP1), regulates the accumulation of both microRNAs and endogenous small interfering RNAs in Arabidopsis. A null mutation of PINP1 causes embryonic lethality, and silencing of PINP1 leads to developmental defects and hypersusceptibility to Phytophthora infection. These phenotypes are reminiscent of transgenic plants expressing PSR1, supporting PINP1 as a direct virulence target of PSR1. We further demonstrate that the localization of the Dicer-like 1 protein complex is impaired in the nucleus of PINP1-silenced or PSR1-expressing cells, indicating that PINP1 may facilitate small RNA processing by affecting the assembly of dicing complexes. A similar function of PINP1 homologous genes in development and immunity was also observed in Nicotiana benthamiana. These findings highlight PINP1 as a previously unidentified component of RNA silencing that regulates distinct classes of small RNAs in plants. Importantly, Phytophthora has evolved effectors to target PINP1 in order to promote infection.

  11. Computer-aided stress analysis system for nuclear plant primary components

    International Nuclear Information System (INIS)

    Murai, Tsutomu; Tokumaru, Yoshio; Yamazaki, Junko.

    1980-06-01

    Generally it needs a vast quantity of calculation to make the stress analysis reports of nuclear plant primary components. In Japan, especially, stress analysis reports are under obligation to make for each plant. In Mitsubishi Heavy Industries, Ltd., We have been making great efforts to rationalize the process of analysis for about these ten years. As the result of rationalization up to now, a computer-aided stress analysis system using graphic display, graphic tablet, data file, etc. was accomplished and it needs us only the least hand work. In addition we developed a fracture safety analysis system. And we are going to develop the input generator system for 3-dimensional FEM analysis by graphics terminals in the near future. We expect that when the above-mentioned input generator system is accomplished, it will be possible for us to solve instantly any case of problem. (author)

  12. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  13. Identification of nonlinear dynamics in power plant components using neural networks

    International Nuclear Information System (INIS)

    Parlos, A.G.; Fernandez, B.; Tsai, W.K.

    1990-01-01

    Advances in digital computer technology have enabled widespread implementation of closed-loop digital control systems in a variety of industries. In some instances, however, the complexity of the plant and the uncertainty associated with the parameters involved in the mathematical modeling narrow the range of applicability of most systematic control system design methodologies. A multiyear project has been initiated to assess the feasibility of the artificial neural networks (ANNs) technology for computerized enhanced diagnostics and control of nuclear power plant components. At this stage of the project, a new methodology, based on backpropagation learning, has been developed for identifying the nonlinear dynamic systems from a set of input-output data known as the training set

  14. Identification of Methylosome Components as Negative Regulators of Plant Immunity Using Chemical Genetics.

    Science.gov (United States)

    Huang, Shuai; Balgi, Aruna; Pan, Yaping; Li, Meng; Zhang, Xiaoran; Du, Lilin; Zhou, Ming; Roberge, Michel; Li, Xin

    2016-12-05

    Nucleotide-binding leucine-rich repeat (NLR) proteins serve as immune receptors in both plants and animals. To identify components required for NLR-mediated immunity, we designed and carried out a chemical genetics screen to search for small molecules that can alter immune responses in Arabidopsis thaliana. From 13 600 compounds, we identified Ro 8-4304 that was able to specifically suppress the severe autoimmune phenotypes of chs3-2D (chilling sensitive 3, 2D), including the arrested growth morphology and heightened PR (Pathogenesis Related) gene expression. Further, six Ro 8-4304 insensitive mutants were uncovered from the Ro 8-4304-insensitive mutant (rim) screen using a mutagenized chs3-2D population. Positional cloning revealed that rim1 encodes an allele of AtICln (I, currents; Cl, chloride; n, nucleotide). Genetic and biochemical analysis demonstrated that AtICln is in the same protein complex with the methylosome components small nuclear ribonucleoprotein D3b (SmD3b) and protein arginine methyltransferase 5 (PRMT5), which are required for the biogenesis of small nuclear ribonucleoproteins (snRNPs) involved in mRNA splicing. Double mutant analysis revealed that SmD3b is also involved in the sensitivity to Ro 8-4304, and the prmt5-1 chs3-2D double mutant is lethal. Loss of AtICln, SmD3b, or PRMT5 function results in enhanced disease resistance against the virulent oomycete pathogen Hyaloperonospora arabidopsidis Noco2, suggesting that mRNA splicing plays a previously unknown negative role in plant immunity. The successful implementation of a high-throughput chemical genetic screen and the identification of a small-molecule compound affecting plant immunity indicate that chemical genetics is a powerful tool to study whole-organism plant defense pathways. Copyright © 2016 The Author. Published by Elsevier Inc. All rights reserved.

  15. Failure investigation of a secondary super heater tube in a 140 MW thermal power plant

    Directory of Open Access Journals (Sweden)

    Atanu Saha

    2017-04-01

    Full Text Available This article describes the findings of a detailed investigation into the failure of a secondary super heater tube in a 140 MW thermal power plant. Preliminary macroscopic examinations along with visual examination, dimensional measurement and chemical analysis were carried out to deduce the probable cause of failure. In addition optical microscopy was a necessary supplement to understand the cause of failure. It was concluded that the tube had failed due to severe creep damage caused by high metal temperature during service. The probable causes of high metal temperature may be in sufficient flow of steam due to partial blockage, presence of thick oxide scale on ID surface, high flue gas temperature etc. rupture.

  16. Development of a mobile unit for 'in loco' sistematic decontamination in nuclear power plants components

    International Nuclear Information System (INIS)

    Camargo, G.A.M.

    1986-01-01

    A mobile decontamination unit was developed to perform 'in situ' decontamination of tanks and pressure vessels belonging to the reactor auxiliary and ancillary systems. The whole system, including a control desk, is assembled in 6 trolleys which can be moved inside the plant, thus enabling component decontamination by injecting demineralized water at a pressure of approx. 50 bar and temperatures up to 90 0 , with or without chemical additives. Considering the versatility and easy handling demonstrated after extensive testing, this new system shall be used in Angra 2 and 3. (Author) [pt

  17. Detection and sizing of defects in structural components of a nuclear power plant by ECT

    International Nuclear Information System (INIS)

    Chen, Z.; Miya, K.

    2004-01-01

    In this paper, progress of ECT (eddy current testing) technique for inspection of stress corrosion cracks in a structural component of a nuclear power plant is reported. Access and scanning vehicle (robot), advanced probes for steam generator tube inspection, development and evaluation of new probes for welding joint, and ECT based crack sizing technique are described respectively. Based on these new techniques, it is clarified that ECT can play as a supplement of ultrasonic techniques for the welding zone inspection. It is also proved in this work that new ECT sensors are efficient even for a stainless plate as thick as 15 mm. (authors)

  18. Detection and Sizing of Defects in Structural Components of a Nuclear Power Plant by ECT

    International Nuclear Information System (INIS)

    Chen Zhenmao; Miya, Kenzo

    2005-01-01

    In this paper, progress of ECT technique for inspection of stress corrosion cracks in a structural component of a nuclear power plant is reported. Access and scanning vehicle (robot), advanced probes for SG tube inspection, development and evaluation of new probes for welding joint, and ECT based crack sizing technique are described respectively. Based on these new techniques, it is clarified that ECT can play as a supplement of UT for the welding zone inspection. It is also proved in this work that new ECT sensors are efficient even for a stainless plate as thick as 15mm

  19. Towards standardized testing methodologies for optical properties of components in concentrating solar thermal power plants

    Science.gov (United States)

    Sallaberry, Fabienne; Fernández-García, Aránzazu; Lüpfert, Eckhard; Morales, Angel; Vicente, Gema San; Sutter, Florian

    2017-06-01

    Precise knowledge of the optical properties of the components used in the solar field of concentrating solar thermal power plants is primordial to ensure their optimum power production. Those properties are measured and evaluated by different techniques and equipment, in laboratory conditions and/or in the field. Standards for such measurements and international consensus for the appropriate techniques are in preparation. The reference materials used as a standard for the calibration of the equipment are under discussion. This paper summarizes current testing methodologies and guidelines for the characterization of optical properties of solar mirrors and absorbers.

  20. Survey of artificial intelligence methods for detection and identification of component faults in nuclear power plants

    International Nuclear Information System (INIS)

    Reifman, J.

    1997-01-01

    A comprehensive survey of computer-based systems that apply artificial intelligence methods to detect and identify component faults in nuclear power plants is presented. Classification criteria are established that categorize artificial intelligence diagnostic systems according to the types of computing approaches used (e.g., computing tools, computer languages, and shell and simulation programs), the types of methodologies employed (e.g., types of knowledge, reasoning and inference mechanisms, and diagnostic approach), and the scope of the system. The major issues of process diagnostics and computer-based diagnostic systems are identified and cross-correlated with the various categories used for classification. Ninety-five publications are reviewed

  1. Ageing studies on materials, components and process instruments used in nuclear power plants

    International Nuclear Information System (INIS)

    Bora, J.S.

    1997-04-01

    This report is a compilation of test results of thermal and radiation ageing tests carried out in the laboratory over a period of 25 years on diverse engineering materials, components and instruments used in nuclear power plants. Test items covered are different types of electrical cables, elastomers, surface coatings, electrical and electronics components and process instruments. Effects of thermal and radiation ageing on performance parameters are shown in tabular forms. Apart from finding the characteristics, capabilities and limitations of test items, ageing research has helped in pin-pointing sub-standard and critical parts and necessary corrective action has been taken. This report is expected to be quite useful to the manufacturers users and researchers for reference and guidance. (author)

  2. Lifetime management for mechanical systems, structures and components in nuclear power plants

    International Nuclear Information System (INIS)

    Roos, E.; Herter, K.-H.; Schuler, X.

    2006-01-01

    Guidelines, codes and standards contain regulations and requirements with respect to the quality of mechanical systems, structures and components (SSC) of nuclear power plants. These concern safe operation during the total lifetime (lifetime management), safety against ageing phenomena (ageing management) as well as proof of integrity (e.g. break exclusion or avoidance of fracture). Within this field the ageing management is a key element. Depending on the safety-relevance of the SSC under observation including preventive maintenance various tasks are required in particular to clarify the mechanisms which contribute system-specifically to the damage of the components and systems and to define their controlling parameters which have to be monitored and checked. Appropriate continuous or discontinuous measures are to be considered in this connection. The approach to ensure a high standard of quality in operation and the management of the technical and organisational aspects are demonstrated and explained

  3. Component and System Sensitivity Considerations for Design of a Lunar ISRU Oxygen Production Plant

    Science.gov (United States)

    Linne, Diane L.; Gokoglu, Suleyman; Hegde, Uday G.; Balasubramaniam, Ramaswamy; Santiago-Maldonado, Edgardo

    2009-01-01

    Component and system sensitivities of some design parameters of ISRU system components are analyzed. The differences between terrestrial and lunar excavation are discussed, and a qualitative comparison of large and small excavators is started. The effect of excavator size on the size of the ISRU plant's regolith hoppers is presented. Optimum operating conditions of both hydrogen and carbothermal reduction reactors are explored using recently developed analytical models. Design parameters such as batch size, conversion fraction, and maximum particle size are considered for a hydrogen reduction reactor while batch size, conversion fraction, number of melt zones, and methane flow rate are considered for a carbothermal reduction reactor. For both reactor types the effect of reactor operation on system energy and regolith delivery requirements is presented.

  4. Risk-based management of remaining life of power plant components

    International Nuclear Information System (INIS)

    Roos, E.; Jovanovic, A.S.; Maile, K.; Auerkari, P.

    1999-01-01

    The paper describes application of different modules of the MPA-System ALIAS in risk-based management of remaining life of power plant components. The system allows comprehensive coverage of all aspects of the remaining life management, including also the risk analysis and risk management. In addition, thanks to the modular character of the system it is also possible to implement new methods: In the case described here, a new (probabilistic) method for determination of the next inspection time for the components exposed to creep loading has been developed and implemented in the system. Practical application of the method has shown (a) that the mean values obtained by the method fall into the range of results obtained by other methods (based on expert knowledge), and (b) that it is possible to quantify the probability of aberration from the mean values. This in turn allows quantifying the additional risks linked to e.g. prolonging of inspection intervals. (orig.) [de

  5. Design and development of major balance of plant components in solid oxide fuel cell system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Wen-Tang; Huang, Cheng-Nan; Tan, Hsueh-I; Chao, Yu [Institute of Nuclear Energy Research Atomic Energy Council, Taoyuan County 32546 (Taiwan, Province of China); Yen, Tzu-Hsiang [Green Technology Research Institute, CPC Corporation, Chia-Yi City 60036 (Taiwan, Province of China)

    2013-07-01

    The balance of plant (BOP) of a Solid Oxide Fuel Cell (SOFC) system with a 2 kW stack and an electric efficiency of 40% is optimized using commercial GCTool software. The simulation results provide a detailed understanding of the optimal operating temperature, pressure and mass flow rate in all of the major BOP components, i.e., the gas distributor, the afterburner, the reformer and the heat exchanger. A series of experimental trials are performed to validate the simulation results. Overall, the results presented in this study not only indicate an appropriate set of operating conditions for the SOFC power system, but also suggest potential design improvements for several of the BOP components.

  6. The use of the acoustic emission for the components of the primary circuit of the nuclear power plants

    International Nuclear Information System (INIS)

    Svoboda, V.

    1992-01-01

    Full text: The Modrany Engineering Works (Modranske strojirny) is a producer and a final supplier of the main connecting piping circuit systems and valves for the nuclear power plants (type VVER 440 and VVER 1000) built in Czechoslovakia. Besides the delivery and assembly of valves and components methods there were developed for a monitoring of the stated equipment ability of a service in the Material and Diagnostic Laboratory, which is a part of the company. An important object of this work is to obtain a sufficient set of data and to work out suitable methods, on the basis of which it would be possible to perform a serious estimation of residual service life of the main piping components after certain service operation of the nuclear power plant. During the operation of a nuclear power station a failure of the main piping circuit could happen in either of two possible modes: 1.) A sudden break - by an unstable defect propagation leading to a. final fracture of the piping; 2) A fatigue failure - which is characterised by a gradual subcritical growth of defect in relation to the loading parameters. This process is frequently accelerated by further processes, e.g. corrosion. It is therefore suitable to use such physical and mechanical quantities, which characterize the material damage. Acoustic emission signals belongs to these quantities. A knowledge of the response of these signals in relation to the damage of the material gives us the possibility to evaluate the residual life of the piping containing defects. The importance of this is increasing mainly after a long period of service. She paper deals in details with experience gained in application of acoustic emission, during pressure tests of primary circuit components (elbow, welds, T- junction etc) in laboratory conditions which imitate those in service. There are shown some results of cyclic fatigue tests by internal pressure on prototypes models and specimen. Acoustic emission method represents the

  7. 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

  8. A multivariate statistical methodology for detection of degradation and failure trends using nuclear power plant operational data

    International Nuclear Information System (INIS)

    Samanta, P.K.; Teichmann, T.

    1990-01-01

    In this paper, a multivariate statistical method is presented and demonstrated as a means for analyzing nuclear power plant transients (or events) and safety system performance for detection of malfunctions and degradations within the course of the event based on operational data. The study provides the methodology and illustrative examples based on data gathered from simulation of nuclear power plant transients (due to lack of easily accessible operational data). Such an approach, once fully developed, can be used to detect failure trends and patterns and so can lead to prevention of conditions with serious safety implications

  9. Structural analysis and incipient failure detection of primary circuit components based on correlation-analysis and finite-element models

    International Nuclear Information System (INIS)

    Olma, B.J.

    1977-01-01

    A method is presented to compute vibrational power spectral densities (VPSD's) of primary circuit components based on a finite-element representation of the primary circuit. First this method has been applied to the sodium cooled reactor KNK, Karlsruhe. Now a further application is being developed for a BWR-nuclear power plant. The experimentally determined VPSD's can be considered as the output of a multiple input-output system. They have to be explained as the frequency response of a multidimensional mechanical system, which is excited by stochastic and deterministic mechanical driving forces. The stochastic mechanical forces are generated by the dynamic pressure fluctuations of the fluid. The deterministic mechanical forces are caused by the pressure fluctuations, which are induced by the main coolant pumps or by standing waves. The excitation matrix can be obtained from measured pressure fluctuations. The vibration transfer function matrix can be computed from the mass matrix, damping matrix and stiffness matrix of a theoretical finite-element model or mass-spring model. Based on this theory the computer code 'STAMPO' has been established. This program has been applied to the KNK reactor. The excitation matrix was created from measured jet-noise pressure fluctuations. The mass-, stiffness- and damping matrix has been extracted from a SAP-IV-model of the primary system. Sequentially for each frequency point the complete VPSD matrix has been computed. The diagonal elements of this matrix represent the vibrational auto-power spectral densities, the off-diagonal elements represent the vibrational cross-power spectral densities. The calculations give good agreement with measured VPSD's. The comparison shows that the measured jet-noise pressure fluctuations act nearly uncorrelated on the structure, whereas the output VPSD's are well correlated

  10. Small punch creep test: A promising methodology for high temperature plant components life evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Tettamanti, S. [CISE SpA, Milan (Italy); Crudeli, R. [ENEL SpA, Milan (Italy)

    1998-12-31

    CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

  11. Small punch creep test: A promising methodology for high temperature plant components life evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Tettamanti, S [CISE SpA, Milan (Italy); Crudeli, R [ENEL SpA, Milan (Italy)

    1999-12-31

    CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

  12. Seismic proving tests on the reliability for large components and equipment of nuclear power plants

    International Nuclear Information System (INIS)

    Ohno, Tokue; Tanaka, Nagatoshi

    1988-01-01

    Since Japan has destructive earthquakes frequently, the structural reliability for large components and equipment of nuclear power plants are rigorously required. They are designed using sophisticated seismic analyses and have not yet encountered a destructive earthquake. When nuclear power plants are planned, it is very important that the general public understand the structural reliability during and after an earthquake. Seismic Proving Tests have been planned by Ministry of International Trade and Industry (Miti) to comply with public requirement in Japan. A large-scale high-performance vibration table was constructed at Tasted Engineering Laboratory of Nuclear Power Engineering Test Center (NU PEC), in order to prove the structural reliability by vibrating the test model (of full scale or close to the actual size) in the condition of a destructive earthquake. As for the test models, the following four items were selected out of large components and equipment important to the safety: Reactor Containment Vessel; Primary Coolant Loop or Primary Loop Recirculation System; Reactor Pressure Vessel; and Reactor Core Internals. Here is described a brief of the vibration table, the test method and the results of the tests on PWR Reactor Containment Vessel and BWR Primary Loop Recirculation System (author)

  13. 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.)

  14. Development of an integrated database management system to evaluate integrity of flawed components of nuclear power plant

    International Nuclear Information System (INIS)

    Mun, H. L.; Choi, S. N.; Jang, K. S.; Hong, S. Y.; Choi, J. B.; Kim, Y. J.

    2001-01-01

    The object of this paper is to develop an NPP-IDBMS(Integrated DataBase Management System for Nuclear Power Plants) for evaluating the integrity of components of nuclear power plant using relational data model. This paper describes the relational data model, structure and development strategy for the proposed NPP-IDBMS. The NPP-IDBMS consists of database, database management system and interface part. The database part consists of plant, shape, operating condition, material properties and stress database, which are required for the integrity evaluation of each component in nuclear power plants. For the development of stress database, an extensive finite element analysis was performed for various components considering operational transients. The developed NPP-IDBMS will provide efficient and accurate way to evaluate the integrity of flawed components

  15. Diagnosing component faults in a generic nuclear power plant using counterfactual and temporal reasoning

    International Nuclear Information System (INIS)

    Oehrstroem, P.; Nielsen, F.R.; Pedersen, S.A.

    1992-01-01

    The subject of main interest is the logical and epistemological aspects of diagnostic reasoning. The aim was to understand the role of conditionals and causality in this respect. A model of causal and temporal reasoning was developed and evaluated in a controlled but complex setting. The generic nuclear power plant was used as a test ground. The coherence and scope of a logical theory of diagnostic reasoning was studied in order to discover whether the theory constitutes an adequate tool for making correct diagnoses of component faults in a generic nuclear power plant. A diagnosing system based on the CIMP system was run on a computer model of a nuclear power plant, various errors were then introduced. The aim of the diagnosis is mainly explanation and only partly repair. The causal field defines a conceptual framework within which the diagnostic purpose is given and within which various diagnostic possibilities and causal relationships are given, here with regard to error detection in a control room. The causal field is tacitly given and related to the operator's training and experience. The logical aspects of the problem of the diagnosis is described. The computer model is described and the symptom language is introduced. The process of reasoning about the possible diagnosis is presented. The utilization of ideas similiar to the heuristic classification is discussed. A data base command language for manipulating lists of symptoms is described and the design of a CIMP user interface for symptom language visualization is outlined. (AB)

  16. The plant i-AAA protease controls the turnover of an essential mitochondrial protein import component.

    Science.gov (United States)

    Opalińska, Magdalena; Parys, Katarzyna; Murcha, Monika W; Jańska, Hanna

    2018-01-29

    Mitochondria are multifunctional organelles that play a central role in energy metabolism. Owing to the life-essential functions of these organelles, mitochondrial content, quality and dynamics are tightly controlled. Across the species, highly conserved ATP-dependent proteases prevent malfunction of mitochondria through versatile activities. This study focuses on a molecular function of the plant mitochondrial inner membrane-embedded AAA protease (denoted i -AAA) FTSH4, providing its first bona fide substrate. Here, we report that the abundance of the Tim17-2 protein, an essential component of the TIM17:23 translocase (Tim17-2 together with Tim50 and Tim23), is directly controlled by the proteolytic activity of FTSH4. Plants that are lacking functional FTSH4 protease are characterized by significantly enhanced capacity of preprotein import through the TIM17:23-dependent pathway. Taken together, with the observation that FTSH4 prevents accumulation of Tim17-2, our data point towards the role of this i -AAA protease in the regulation of mitochondrial biogenesis in plants. © 2018. Published by The Company of Biologists Ltd.

  17. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  18. The safety related aspects of pressure components in nuclear power plants

    International Nuclear Information System (INIS)

    Lindackers, K.H.

    1979-01-01

    Over the last two years the safety philosophy for nuclear power plants in the Federal Republic of Germany has changed considerably, as everyone working in the field perceives. The original and appropriate philosophy of risk minimalisation through graduated safety barriers has been more and more replaced by the utopian goal of total prevention of any damage. The reasons for this development are discussed briefly especially regarding pressure components. The very numerous pressure components of a nuclear power station are not all of equal importance with respect to safety. Although considerable efforts have been made, it has not been possible, to date, to achieve an agreement between operators, manufacturers, licensing authorities, independent experts, and other specialists about the safety related classification of the manifold pressure bearing parts in nuclear power stations. The background of this extremely regrettable situation is explained. In the last part of the paper the author suggests a simple and clear safety philosophy for pressure components in nuclear power stations. This philosophy is orientated both on Safety Regulations of the Radiation Protection Decree ('Strahlenschutzverordnung') of the 13th October 1976 and on the Safety Criteria for Nuclear Power Stations from 21st October 1977. Only a simple, clear framework can make a contribution to the further improvement of the already exceptional safety of nuclear facilities and to the removal of obstacles in the licensing procedure which, taken as a whole, tie up skilled personnel to a senseless degree, involve considerable financial expenditure, and have no relevance for the safety of nuclear power plants. (orig.) [de

  19. A Study on the Organizational Components Affecting the Communication-Related Events in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lee, Seung Min; Jang, In Seok; Seong, Poong Hyun

    2009-01-01

    It is important to communicate clearly and effectively in order to achieve and improve team performance, also in the view point of safety, in nuclear power plant (NPP). Researchers have studied on lots of accidents and incidents related to communication and analyzed the elements affecting communication fail in the side of sender-receiver communication process so that they have found which process was failed to communicate each other. But we cannot disregard on human cognition, level of understanding, and individual or team characteristic on the communication process, so we need to analyze the elements of communication-related events in the side of human and team components that we will find why operators could not avoid failing their communication. In this paper we enumerate key organizational components, collect events related to communication in NPP and count the total number of components affecting communication fail. Finally we perform the pairwise-comparison using those values and understand major factors affecting communication-related events

  20. Three technical issues in fatigue damage assessment of nuclear power plant components

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1991-01-01

    This paper addresses three technical issues that affect the fatigue damage assessment of nuclear power plant components: the effect of the environment on the fatigue life, the importance of the loading sequence in calculating the fatigue crack-initiation damage, and the adequacy of current inservice inspection requirements and methods to characterize fatigue cracks. The environmental parameters that affect the fatigue life of carbon and low alloy steel components are the sulphur content in the steel, the temperature, the amount of dissolved oxygen in the coolant, and the presence of oxidizing agents such as copper oxide. The occurrence of large-amplitude stress cycles early in a component's life followed by low-amplitude stress cycles may cause crack initiation at a cumulative usage factor less than 1.0. The current inservice inspection requirements include volumetric inspections of welds but not of some susceptible sites in the base metal. In addition, the conventional ultrasonic testing techniques need to be improved for reliable detection and accurate sizing of fatigue cracks. 28 refs., 4 figs., 1 tab

  1. Studies on the muscle-paralyzing components of the juice of the banana plant.

    Science.gov (United States)

    Singh, Y N; Inman, W D; Johnson, A; Linnell, E J

    1993-01-01

    The stem juice of the banana plant (Musa species) has been used as an arrow poison by African tribesmen. Lyophilized, partially purified extracts of the juice augment and then block both directly and indirectly evoked contractions of the mouse diaphragm. We have isolated, purified and determined the chemical composition of the active ingredients, and characterized their pharmacological activity. The lyophilized sample was extracted with a methanol-water (MeOH-H2O) (50/50) mixture and vacuum filtered. The filtrate was rotary evaporated and crystallized in a MeOH-H2O mixture to yield potassium nitrate crystals (melting point 332-334 degrees C). The filtrate was concentrated and chromatographed over Sephadex LH-20 gel using MeOH-H2O (40/60) as the eluent. The active component was found to be magnesium nitrate crystals (melting point 87-89 degrees C). In the mouse isolated phrenic nerve-hemidiaphragm preparation, the pharmacological profile of the first component was similar to that for authentic potassium nitrate which augments in low concentrations, and in higher concentrations augments, and then blocks both directly evoked muscle contraction the neuromuscular transmission. The second component had a profile of activity similar to that for authentic magnesium nitrate which only blocks neuromuscular transmission. It can be concluded that the two major active principles in the banana stem juice are potassium nitrate and magnesium nitrate.

  2. Metallurgical considerations in the design of creep exposed, high temperature components for advanced power plants

    International Nuclear Information System (INIS)

    Schubert, F.

    1990-08-01

    Metallic components in advanced power generating plants are subjected to temperatures at which the material properties are significantly time-dependent, so that the creep properties become dominant for the design. In this investigation, methods by which such components are to be designed are given, taking into account metallurgical principles. Experimental structure mechanics testing of component related specimens carried out for representative loading conditions has confirmed the proposed methods. The determination of time-dependent design values is based on a scatterband evaluation of long-term testing data obtained for a number of different heats of a given alloy. The application of computer-based databank systems is recommendable. The description of the technically important secondary creep rate based on physical metallurgy principles can be obtained using the exponential relationship originally formulated by Norton, ε min = k.σ n . The deformation of tubes observed under internal pressure with a superimposed static or cyclic tensile stress and a torsion loading can be adequately described with the derived, three-dimensional creep equation (Norton). This is also true for the description of creep ratcheting and creep buckling phenomena. By superimposing a cyclic stress, the average creep rate is increased in one of the principal deformation axes. This is also true for the creep crack growth rate. The Norton equation can be used to derive this type of deformation behaviour. (orig.) [de

  3. The Effect of Drought Stress on Morphological Characteristics and Yield Components of Medicinal Plant Fenugreek

    Directory of Open Access Journals (Sweden)

    N. Bazzazi

    2013-06-01

    Full Text Available Fenugreek (Trigonella foenum-graecum L. is one of the oldest medicinal plants. In order to study water-stress effects on some morphological characteristics of fenugreek, an experiment was carried out in a strip plots based on randomized complete blocks design with three replicates, at Research Farm of Shahrekord University, Shahrekord, Iran, in 2010. The first factor was allocated to four water stress levels (irrigation after depletion of 20 (as control, 40, 60 and 80% of available soil moisture and the second factor was six fenugreek landraces (Shiraz, Ardestan, Tirancheh, Yazd, Jahrom and Hindi. The results of ANOVA and comparison of means indicated that the effect of water stress was significant for all traits and variation was observed among landraces for all the studied characteristics. Mean comparison showed that drought stress reduced days to flowering, days to maturity, plant height and yield components (number of pods per plant, number of seeds per pod and 1000-kernel weight. It was also revealed that water stress caused reduction in biological yield (43% and grain yield (42.3% of all genotypes. Comparison between landraces indicated that maximum biological and grain yield belonged to Ardestan landrace. Assessment of cluster analysis showed that it was possible to classify Ardestan, Shiraz and Tirancheh as a single group having tolerance to water stress. In general, based on obtained results, the Ardestan landrace, with 22.37 g/plant, had the highest biological yield and Hindi landrace, with 73.83 days to maturity, was the most early-maturing one.

  4. Status on the Component Models Developed in the Modelica Framework: High-Temperature Steam Electrolysis Plant & Gas Turbine Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Suk Kim, Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); McKellar, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Boardman, Richard D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-10-01

    This report has been prepared as part of an effort to design and build a modeling and simulation (M&S) framework to assess the economic viability of a nuclear-renewable hybrid energy system (N-R HES). In order to facilitate dynamic M&S of such an integrated system, research groups in multiple national laboratories have been developing various subsystems as dynamic physics-based components using the Modelica programming language. In fiscal year (FY) 2015, Idaho National Laboratory (INL) performed a dynamic analysis of two region-specific N-R HES configurations, including the gas-to-liquid (natural gas to Fischer-Tropsch synthetic fuel) and brackish water reverse osmosis desalination plants as industrial processes. In FY 2016, INL has developed two additional subsystems in the Modelica framework: a high-temperature steam electrolysis (HTSE) plant and a gas turbine power plant (GTPP). HTSE has been proposed as a high priority industrial process to be integrated with a light water reactor (LWR) in an N-R HES. This integrated energy system would be capable of dynamically apportioning thermal and electrical energy (1) to provide responsive generation to the power grid and (2) to produce alternative industrial products (i.e., hydrogen and oxygen) without generating any greenhouse gases. A dynamic performance analysis of the LWR/HTSE integration case was carried out to evaluate the technical feasibility (load-following capability) and safety of such a system operating under highly variable conditions requiring flexible output. To support the dynamic analysis, the detailed dynamic model and control design of the HTSE process, which employs solid oxide electrolysis cells, have been developed to predict the process behavior over a large range of operating conditions. As first-generation N-R HES technology will be based on LWRs, which provide thermal energy at a relatively low temperature, complementary temperature-boosting technology was suggested for integration with the

  5. Health Effects of Bioactive Components in Plant Foods; Results and Opinion of the EU-COST926 Action

    NARCIS (Netherlands)

    Verkerk, R.; Piskula, M.; Bovy, Arnaud; Dekker, M.

    2014-01-01

    This paper reviews the main results of EU-action: “COST 926: Impact of new technologies on the health benefits and safety of bioactive plant compounds”. The bioavailability and the effects on gene expression of various bioactive components in plant foods are described in relation with their

  6. Component behaviour in the 700 C power plant. Numerical and experimental investigations

    International Nuclear Information System (INIS)

    Schmidt, Kay H.

    2013-01-01

    Currently martensitic steels are used in fossil fired power plants with maximum working temperatures up to 625 C. These steels do not show the required creep rupture strength at the target temperature of 700 C. For these high temperatures, new materials like the nickel base alloys have to be qualified for power plants services. Originating from the weld of turbine materials, nickel base alloys show outstanding creep rupture strength. An alloy with good prospects out of the material class of the nickel base alloys is Alloy 617 mod. However, this material is expensive due to its high nickel content. Furthermore, the complex machinability of this material leads to an additional increase in expenses. A complete fabrication of the boiler area using Alloy 617 mod is not economically feasible, which means that the usage of this material has to be limited to the temperature weld of 625 C to 700 C. For the boiler area with temperatures below 625 C the well proven 9 % to 12 % Cr-steels, like T/P92 and VM12/VM12-SHC may be used. In the weld of low temperatures up to 525 C the usage of the 2.5 % Cr-steel T/P24 offers numerous advantages, in particular in the fabrication of membrane walls. This material shows good creep properties up to temperatures of 525 C and, for thin walled components, T24 can be welded without post weld heat treatment by using suitable techniques. For a successful design and fabrication of a 700 C fossil fired power plant, appropriate materials have to be qualified. Here, a special focus is set on the creep properties of these materials. The presented work is a significant contribution to the qualification of these materials. First, the materials Alloy 617 mod, T/P92, VM12/VM12-SHC and T24 are briefly introduced and characterized. After this, the materials are investigated in a detailed creep testing program. This program includes investigations on base material, extracted from tubes, pipes and inductive bends of pipes. In addition, crossweld specimens

  7. Failure diagnostics of rotational machines and machine groups in the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Nagy, I.; Kiss, G.

    1997-01-01

    Machine failure diagnostic system based on vibration analysis is described. The design of the measurements, the measurement procedure, and the evaluation of the results are presented. The common diagnostic technology for groups of rotation machinery has several advantages. The rules used for the evaluation of three-directional vibration measurements are shown, and the identification criteria of some specific failures are demonstrated. The steps of expert decision making and the diagnosis procedure are discussed through practical examples. (R.P.)

  8. Advanced numerical description of the behavior of 700 C steam power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Maile, K. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Schmidt, K.; Roos, E.; Klenk, A.; Speicher, M.

    2009-07-01

    To make full use of the strength potential of new boiler materials like the new 9-11% Cr steels and nickel based alloys, taking into account their specific stress-strain relaxation behavior, new design methods are required in the design of today's power plants. Highly loaded components are approaching more and more the classical design limits with regard to critical wall thicknesses and the related tolerable thermal gradients, due to planed increases of steam parameters like steam pressure and steam temperature. ''Design by analysis'' can be realized by modern state of the art Numerical Finite Element (FE) simulation codes and in some cases by the use of user defined advanced inelastic material laws. These material laws have to be adjusted to specific material behavior of new boiler materials. To model the strain and stress situation in components under high temperature loading, a constitutive equation based on a Graham-Walles approach is used in this paper. Furthermore essential steps and recommendations to implement experimental data in the user defined subroutines and the subsequent integration of the subroutines in modern FE codes like ABAQUS trademark and ANSYS trademark are given. As an example, the results of FE simulations of components like hollow cylinders and waterwall like components made of Alloy 617 or 9-11% Cr steels are discussed and verified with experimental results. In a last step, the successful application of the developed creep equation will be demonstrated by calculating the creep strains and stress relaxation of a P92 steam header under constant loading. (orig.)

  9. Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals

    International Nuclear Information System (INIS)

    1999-10-01

    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 be therefore 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 at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant

  10. 10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

    Science.gov (United States)

    2010-01-01

    ... Equipment and Components Under NRC Export Licensing Authority H Appendix H to Part 110 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Pt. 110, App. H Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under...

  11. Quantifying the Components of Evapotranspiration from Plant Communities, Soil Evaporation and Plant Transpiration, with Oxygen-18 Isotopes and Micrometeorology

    Energy Technology Data Exchange (ETDEWEB)

    Denmead, Tom [CSIRO Centre for Environmental Mechanics, GPO Box 821, Canberra, ACT 2601 (Australia); Heng, Lee; Nguyen, Long [Soil and Water Management and Crop Nutrition Section, IAEA (Austria); Zeeman, Matthias [Karlsruhe Institute of Technology, Garmisch-Partenkirchen (Germany); Mayr, Leo; Arrillaga, Jose Luis [Soil and Water Management and Crop Nutrition Laboratory, IAEA (Austria); Cepuder, Peter [Department of Water-Atmosphere-Environment, Institute for Hydraulics and Rural Water Management (BOKU), Vienna (Austria)

    2013-01-15

    The Keeling plot (Keeling, 1961) approach has been shown to provide an estimate of the relative proportions of water vapour emanating from evaporation (E) from soil, and transpiration (T) from the plant canopy (Moreira et. al., 1997; Williams et al., 2004). This estimate can be used in conjunction with measurements of the net water vapour flux and evapotranspiration (ET), to quantify the E and T components using an Inverse Lagrangian (IL) approach based on canopy turbulence (Raupach, 1989), which allows the identification of water vapour in the different canopy layers (Denmead et al., 2005). A study was carried out on a wheat crop over a 3-day period in April (daily temperatures ranged from 14-23''oC) at the BOKU experimental field outside Vienna to provide an independent check of the relative proportions of soil evaporation (E) and plant transpiration (T) estimated by the Keeling plot {sigma}{sup 18}O isotope analysis and by the application of the IL model of water vapour transport in plant canopies. The eddy covariance instrumentation to measure ET was provided by the Karlsruhe Institute of Technology at Garmisch-Partenkirchen, Germany. Transpiration rates, estimated by the {sigma}{sup 18}O isotopic technique were similar to those derived from Inverse Lagrangian analyses. indicating that the IL and isotopic analyses gave essentially the same partitioning of evapotranspiration into E and T. The use of the IL analysis to determine water vapour in different segments of the canopy is illustrated. In these observations the soil was dry (9-12 %) and soil evaporation was small. The eddy covariance approach confirmed the correctness of the IL analysis for the total water loss from the canopy (to within 6%) (data not shown). The IL and the isotopic analyses gave essentially the same partitioning of ET into E and T for 3 days on a dry soil. The isotopic analysis using {sigma}{sup 18}O gave E/ET {approx} 4% and T/ET {approx} 96%, while IL analysis gave corresponding figures

  12. The effects of aging on electrical and I ampersand C components: Results of US Nuclear Plant Aging Research

    International Nuclear Information System (INIS)

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

    1993-01-01

    The US NRC's hardware oriented engineering research program for plant aging and degradation monitoring has achieved results in the area of electrical, control, and instrumentation (ECI) components used in nuclear power plants (NPPs). The principal goals of the program, known as the Nuclear Power Plant Aging Research (NPAR) Program, are to understand the effects of age-related degradation in NPPs and how to manage and mitigate them effectively. This paper describes how these goals have been achieved for key ECI components used in the safety systems of NPPs. The status of relevant on-going and planned research projects is also provided

  13. The effects of aging on electrical and I ampersand C components: Results of US nuclear plant aging research

    International Nuclear Information System (INIS)

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

    1991-01-01

    The US NRC's hardware oriented engineering research program for plant aging and degradation monitoring has achieved results in the area of electrical, control, and instrumentation (ECI) components used in nuclear power plants (NPPs). The principal goals of the program, known as the Nuclear Power Plant Aging Research (NPAR) Program, are to understand the effects of age-related degradation in NPPs and how to manage and mitigate them effectively. This paper describes how these goals have been achieved for key ECI components used in the safety systems of NPPs. The status of relevant on-going and planned research projects is also provided

  14. Application of Ultra High Pressure Cavitation Peening to Prevent PWSCC on Primary Plant Components

    Energy Technology Data Exchange (ETDEWEB)

    Poling, G.R.

    2015-07-01

    Primary Water Stress Corrosion Cracking (PWSCC) on Alloy 600/82/182 susceptible materials can lead to increased costs for maintenance and repair/replacement activities on nuclear power plant primary components. A process called Ultra High Pressure (UHP) cavitation peening can be safely and cost effectively applied to the susceptible materials to generate compressive stresses on the surface and prevent PWSCC initiation. AREVA has developed the tooling systems to apply the UHP cavitation peening process on reactor vessel head penetration nozzles, bottom mounted nozzles and primary nozzles. Applying the UHP cavitation peening process before PWSCC initiation will prevent future repairs/replacements, reduce maintenance costs, and provide more effective on-time for the reactor. (Author)

  15. Security Hardened Cyber Components for Nuclear Power Plants: Phase I SBIR Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Franusich, Michael D. [SpiralGen, Inc., Pittsburgh, PA (United States)

    2016-03-18

    SpiralGen, Inc. built a proof-of-concept toolkit for enhancing the cyber security of nuclear power plants and other critical infrastructure with high-assurance instrumentation and control code. The toolkit is based on technology from the DARPA High-Assurance Cyber Military Systems (HACMS) program, which has focused on applying the science of formal methods to the formidable set of problems involved in securing cyber physical systems. The primary challenges beyond HACMS in developing this toolkit were to make the new technology usable by control system engineers and compatible with the regulatory and commercial constraints of the nuclear power industry. The toolkit, packaged as a Simulink add-on, allows a system designer to assemble a high-assurance component from formally specified and proven blocks and generate provably correct control and monitor code for that subsystem.

  16. Materials, manufacture and testing of pressurized components of high-power steam power plants

    International Nuclear Information System (INIS)

    Blind, D.; Foehl, J.; Issler, L.; Schellhammer, W.; Sturm, D.; Kussmaul, K.; Heinrich, D.; Meyer, H.J.; Prestel, W.

    1981-01-01

    This is the first German review of materials, production and testing of pressure components of high-capacity steam power plants. The authors have been working in this field for years; their special subject has been the availability and reliability of pressure vessels, in particular in nuclear engineering. Fundamentals are presented as well as the findings obtained at the state Materials Testing Institute in Stuttgart. The material is presented in a well-structured classification; the most recent international findings, especially of the USA, are presented. This is possible due to the close cooperation between the Stuttgart institute and a number of US research institutes. The new subject of fracture mechanics is treated in some detail; its fundamentals are discussed from the American point of view while German considerations - in particular of the Reactor Safety Commission - are taken into account in the field of applications. (orig.) [de

  17. Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

    International Nuclear Information System (INIS)

    Charalambus, B.

    1992-01-01

    Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative. (orig.)

  18. Myrtaceae Plant Essential Oils and their β-Triketone Components as Insecticides against Drosophila suzukii

    Directory of Open Access Journals (Sweden)

    Chung Gyoo Park

    2017-06-01

    Full Text Available Spotted wing drosophila (SWD, Drosophila suzukii (Matsumura, Diptera: Drosophilidae is recognized as an economically important pest in North America and Europe as well as in Asia. Assessments were made for fumigant and contact toxicities of six Myrtaceae plant essential oils (EOs and their components to find new alternative types of insecticides active against SWD. Among the EOs tested, Leptospermum citratum EO, consisting mainly of geranial and neral, exhibited effective fumigant activity. Median lethal dose (LD50; mg/L values of L. citratum were 2.39 and 3.24 for males and females, respectively. All tested EOs except Kunzea ambigua EO exhibited effective contact toxicity. LD50 (µg/fly values for contact toxicity of manuka and kanuka were 0.60 and 0.71, respectively, for males and 1.10 and 1.23, respectively, for females. The LD50 values of the other 3 EOs-L. citratum, allspice and clove bud were 2.11–3.31 and 3.53–5.22 for males and females, respectively. The non-polar fraction of manuka and kanuka did not show significant contact toxicity, whereas the polar and triketone fractions, composed of flavesone, isoleptospermone and leptospermone, exhibited efficient activity with the LD50 values of 0.13–0.37 and 0.22–0.57 µg/fly for males and females, respectively. Our results indicate that Myrtaceae plant EOs and their triketone components can be used as alternatives to conventional insecticides.

  19. Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

    International Nuclear Information System (INIS)

    De Almeida, Valmor F.

    2011-01-01

    Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for

  20. Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

    Energy Technology Data Exchange (ETDEWEB)

    de Almeida, Valmor F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2011-08-15

    Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for

  1. Interaction Effects of Planting Date and Weed Competition on Yield and Yield Components of Three white Bean Cultivars in Semirom

    Directory of Open Access Journals (Sweden)

    A. Yadavi

    2012-06-01

    Full Text Available Unsuitable planting and weed competition are the most important factors that greatly reduce the yield of bean. In order to study the effect of planting date on yield and yield components of three white bean cultivars in weed infest and weed free condition a factorial experiment with randomized complete block design and three replications was carried out at Semirom in 2009. The treatments were planting date (May10, May 25 and June 9 and white bean cultivars (Shekofa, Pak and Daneshkade and two levels of weed infestation (weedy and weed free. Results showed that planting date, weed competition and cultivars had significant effects on yield and yield components of white bean. The 30-day delay in planting date reduced the number of pods per plant, seeds per pod, 100 seed weight and biological yield of white bean cultivars, 22.5, 18, 20.1 and 22.5 percent respectively. Also weed competition, reduced the number of seeds per pod, 100 seed weight and biological yield respectively by 13.5, 5.7 and 27.1 percent. Result of planting date and weed competition interaction effects indicated that the weed competition decreased grain yield (53% in third planting date more than others and delay in planting date was companion with increasing weed density and dry weight in flowering stage of bean. Also Shekofa cultivar had highest grain yield (3379 kg/ha at the first planting date and weed free condition.

  2. PROBABILITY OF FAILURE OF THE TRUDOCK CRANE SYSTEM AT THE WASTE ISOLATION PILOT PLANT (WIPP)

    International Nuclear Information System (INIS)

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

    2000-01-01

    This probabilistic analysis of WIPP TRUDOCK crane failure is based on two sources of failure data. The source for operator errors is the report by Swain and Guttman, NUREG/CR-1278-F, August 1983. The source for crane cable hook breaks was initially made by WIPP/WID-96- 2196, Rev. O by using relatively old (1970s) U.S. Navy data (NUREG-0612). However, a helpful analysis by R.K. Deremer of PLG guided the authors to values that were more realistic and more conservative, with the recommendation that the crane cable/hook failure rate should be 2.5 x 10-6 per demand. This value was adopted and used. Based on these choices a mean failure rate of 9.70 x 10-3(1/yr) was calculated. However, a mean rate by itself does not reveal the level of confidence to be associated with this number. Guidance to making confidence calculations came from the report by Swain and Guttman, who stated that failure data could be described by lognormal distributions. This is in agreement with the widely use d reports (by DOE and others) NPRD-95 and NPRD-91, on failure data. The calculations of confidence levels showed that the mean failure rate of 9.70x 10-3(1/yr) corresponded to a percentile value of approximately 71; i.e. there is a 71% likelihood that the failure rate is less than 9.70x 10-3(1/yr). One also calculated that there is a 95% likelihood that the failure rate is less than 29.6x 10-3(1/yr). Or, as stated previously, there is a 71% likelihood that not more than one dropped load will occur in 103 years. Also, there is a 95% likelihood that not more than one dropped load will occur in approximately 34 years. It is the responsibility of DOE to select the confidence level at which it desires to operate

  3. Fatigue evaluation including environmental effects for primary circuit components in nuclear power plants

    International Nuclear Information System (INIS)

    Seichter, Johannes; Reese, Sven H.; Klucke, Dietmar

    2013-01-01

    The influence of LWR coolant environment to the lifetime of materials in Nuclear Power Plants is in discussion internationally. Environmental phenomena were investigated in laboratory tests and published in recent years. The discussion is mainly focused both on the transition from laboratory to real plant components and on numerical calculation procedures. Since publishing of the NUREG/CR-6909 report in 2007, formulae for calculating the Fen factors have been modified several times. Various calculation procedures like the so called 'Strain-integrated Method' and 'Simplified Approach' have been published while each approach yields to different results. The recent revision of the calculation procedure, proposed by ANL in 2012, is presented and discussed with regard to possible variations in the results depending on the assumptions made. In German KTA Rules the effect of environmentally assisted fatigue (EAF) is taken into account by means of so called attention thresholds. If the threshold value is exceeded, further measures like NDT, in-service inspections including fracture mechanical evaluations or detailed assessment procedures have to be performed. One way to handle those measures is to apply sophisticated procedures and to show that the calculated CUF is below the defined attention thresholds. On the basis of a practical example, methods and approaches will be discussed and recommendations in terms of avoiding over-conservatism and misinterpretation will be presented.

  4. Fatigue evaluation including environmental effects for primary circuit components in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Seichter, Johannes [Siempelkamp Pruef- und Gutachter-Gesellschaft mbH, Dresden (Germany); Reese, Sven H.; Klucke, Dietmar [Component Technology Global Unit Generation, E.ON Kernkraft GmbH, Hannover (Germany)

    2013-05-15

    The influence of LWR coolant environment to the lifetime of materials in Nuclear Power Plants is in discussion internationally. Environmental phenomena were investigated in laboratory tests and published in recent years. The discussion is mainly focused both on the transition from laboratory to real plant components and on numerical calculation procedures. Since publishing of the NUREG/CR-6909 report in 2007, formulae for calculating the Fen factors have been modified several times. Various calculation procedures like the so called 'Strain-integrated Method' and 'Simplified Approach' have been published while each approach yields to different results. The recent revision of the calculation procedure, proposed by ANL in 2012, is presented and discussed with regard to possible variations in the results depending on the assumptions made. In German KTA Rules the effect of environmentally assisted fatigue (EAF) is taken into account by means of so called attention thresholds. If the threshold value is exceeded, further measures like NDT, in-service inspections including fracture mechanical evaluations or detailed assessment procedures have to be performed. One way to handle those measures is to apply sophisticated procedures and to show that the calculated CUF is below the defined attention thresholds. On the basis of a practical example, methods and approaches will be discussed and recommendations in terms of avoiding over-conservatism and misinterpretation will be presented.

  5. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  6. Investigation of waste heat recovery of binary geothermal plants using single component refrigerants

    Science.gov (United States)

    Unverdi, M.

    2017-08-01

    In this study, the availability of waste heat in a power generating capacity of 47.4 MW in Germencik Geothermal Power Plant has been investigated via binary geothermal power plant. Refrigerant fluids of 7 different single components such as R-134a, R-152a, R-227ea, R-236fa, R-600, R-143m and R-161 have been selected. The binary cycle has been modeled using the waste heat equaling to mass flow rate of 100 kg/s geothermal fluid. While the inlet temperature of the geothermal fluid into the counter flow heat exchanger has been accepted as 110°C, the outlet temperature has been accepted as 70°C. The inlet conditions have been determined for the refrigerants to be used in the binary cycle. Finally, the mass flow rate of refrigerant fluid and of cooling water and pump power consumption and power generated in the turbine have been calculated for each inlet condition of the refrigerant. Additionally, in the binary cycle, energy and exergy efficiencies have been calculated for 7 refrigerants in the availability of waste heat. In the binary geothermal cycle, it has been found out that the highest exergy destruction for all refrigerants occurs in the heat exchanger. And the highest and lowest first and second law efficiencies has been obtained for R-600 and R-161 refrigerants, respectively.

  7. Study on applicability of highly corrosion-resistant amorphous coating techniques to components of reprocessing plant

    International Nuclear Information System (INIS)

    Ebata, Makoto; Okuyama, Gen; Chiba, Shigeru; Matsunaga, Tsunebumi

    1991-01-01

    In view of the growing need for prolongation of lives of reprocessing plant installations, we recently investigated the applicability of highly corrosion-resistant amorphous coating techniques to such plant components as to be subjected to a badly corrosive environment created by high temperatures, boiling nitric acid (HNO 3 ), etc. As the result, giving a preference to the Ta-based amorphous alloys exhibiting high corrosion-resistance in HNO 3 solutions, we made specimens of stainless steel plates coated with the above amorphous alloys through the sputtering process thereof. To our satisfaction, these specimens successfully passed various HNO 3 corrosion tests as described later on. Ta-based amorphous films give cathodic protection to 310 Nb stainless steel plates, and that with extremely low corrosion rates of themselves as protecting agents. For these reasons, we are confident that there will be no practical problems at all, in case we adopt stainless steel plates partially coated with such amorphous alloys for use in a nitric-acid environment. In this paper, we explain the comparative tests for various amorphous alloys with different compositions, referring also to the thus-selected Ta-based amorphous alloy along with several kinds of corrosion tests specially arranged for the same alloy. (author)

  8. Consideration of a design optimization method for advanced nuclear power plant thermal-hydraulic components

    International Nuclear Information System (INIS)

    Ridluan, Artit; Tokuhiro, Akira; Manic, Milos; Patterson, Michael; Danchus, William

    2009-01-01

    In order to meet the global energy demand and also mitigate climate change, we anticipate a significant resurgence of nuclear power in the next 50 years. Globally, Generation III plants (ABWR) have been built; Gen' III+ plants (EPR, AP1000 others) are anticipated in the near term. The U.S. DOE and Japan are respectively pursuing the NGNP and MSFR. There is renewed interest in closing the fuel cycle and gradually introducing the fast reactor into the LWR-dominated global fleet. In order to meet Generation IV criteria, i.e. thermal efficiency, inherent safety, proliferation resistance and economic competitiveness, plant and energy conversion system engineering design have to increasingly meet strict design criteria with reduced margin for reliable safety and uncertainties. Here, we considered a design optimization approach using an anticipated NGNP thermal system component as a Case Study. A systematic, efficient methodology is needed to reduce time consuming trial-and-error and computationally-intensive analyses. We thus developed a design optimization method linking three elements; that is, benchmarked CFD used as a 'design tool', artificial neural networks (ANN) to accommodate non-linear system behavior and enhancement of the 'design space', and finally, response surface methodology (RSM) to optimize the design solution with targeted constraints. The paper presents the methodology including guiding principles, an integration of CFD into design theory and practice, consideration of system non-linearities (such as fluctuating operating conditions) and systematic enhancement of the design space via application of ANN, and a stochastic optimization approach (RSM) with targeted constraints. Results from a Case Study optimizing the printed circuit heat exchanger for the NGNP energy conversion system will be presented. (author)

  9. Application range affected by software failures in safety relevant instrumentation and control systems of nuclear power plants

    International Nuclear Information System (INIS)

    Jopen, Manuela; Mbonjo, Herve; Sommer, Dagmar; Ulrich, Birte

    2017-03-01

    This report presents results that have been developed within a BMUB-funded research project (Promotion Code 3614R01304). The overall objective of this project was to broaden the knowledge base of GRS regarding software failures and their impact in software-based instrumentation and control (I and C) systems. To this end, relevant definitions and terms in standards and publications (DIN, IEEE standards, IAEA standards, NUREG publications) as well as in the German safety requirements for nuclear power plants were analyzed first. In particular, it was found that the term ''software fault'' is defined differently and partly contradictory in the considered literature sources. For this reason, a definition of software fault was developed on the basis of the software life cycle of software-based I and C systems within the framework of this project, which takes into account the various aspects relevant to software faults and their related effects. It turns out that software failures result from latent faults in a software-based control system, which can lead to a non-compliant behavior of a software-based I and C system. Hereby a distinction should be made between programming faults and specification faults. In a further step, operational experience with software failures in software-based I and C systems in nuclear facilities and in nonnuclear sector was investigated. The identified events were analyzed with regard to their cause and impacts and the analysis results were summarized. Based on the developed definition of software failure and on the COMPSIS-classification scheme for events related to software based I and C systems, the COCS-classification scheme was developed to classify events from operating experience with software failures, in which the events are classified according to the criteria ''cause'', ''affected system'', ''impact'' and ''CCF potential''. This classification scheme was applied to evaluate the events identified in the framework of this project

  10. Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

    1996-01-01

    A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

  11. In Vitro Control of Post-Harvest Fruit Rot Fungi by Some Plant Essential Oil Components

    Directory of Open Access Journals (Sweden)

    Gian Luigi Rana

    2012-02-01

    Full Text Available Eight substances that are main components of the essential oils from three Mediterranean aromatic plants (Verbena officinalis, Thymus vulgaris and Origanum vulgare, previously found active against some phytopathogenic Fungi and Stramenopila, have been tested in vitro against five etiological agents of post-harvest fruit decay, Botrytis cinerea, Penicillium italicum, P. expansum, Phytophthora citrophthora and Rhizopus stolonifer. The tested compounds were β-fellandrene, β-pinene, camphene, carvacrol, citral, o-cymene, γ-terpinene and thymol. Citral exhibited a fungicidal action against P. citrophthora; carvacrol and thymol showed a fungistatic activity against P. citrophthora and R. stolonifer. Citral and carvacrol at 250 ppm, and thymol at 150 and 250 ppm stopped the growth of B. cinerea. Moreover, thymol showed fungistatic and fungicidal action against P. italicum. Finally, the mycelium growth of P. expansum was inhibited in the presence of 250 ppm of thymol and carvacrol. These results represent an important step toward the goal to use some essential oils or their components as natural preservatives for fruits and foodstuffs, due to their safety for consumer healthy and positive effect on shelf life extension of agricultural fresh products.

  12. Plant immune and growth receptors share common signalling components but localise to distinct plasma membrane nanodomains.

    Science.gov (United States)

    Bücherl, Christoph A; Jarsch, Iris K; Schudoma, Christian; Segonzac, Cécile; Mbengue, Malick; Robatzek, Silke; MacLean, Daniel; Ott, Thomas; Zipfel, Cyril

    2017-03-06

    Cell surface receptors govern a multitude of signalling pathways in multicellular organisms. In plants, prominent examples are the receptor kinases FLS2 and BRI1, which activate immunity and steroid-mediated growth, respectively. Intriguingly, despite inducing distinct signalling outputs, both receptors employ common downstream signalling components, which exist in plasma membrane (PM)-localised protein complexes. An important question is thus how these receptor complexes maintain signalling specificity. Live-cell imaging revealed that FLS2 and BRI1 form PM nanoclusters. Using single-particle tracking we could discriminate both cluster populations and we observed spatiotemporal separation between immune and growth signalling platforms. This finding was confirmed by visualising FLS2 and BRI1 within distinct PM nanodomains marked by specific remorin proteins and differential co-localisation with the cytoskeleton. Our results thus suggest that signalling specificity between these pathways may be explained by the spatial separation of FLS2 and BRI1 with their associated signalling components within dedicated PM nanodomains.

  13. Impact of valve failures on the safety and reliability of light water nuclear power plants

    International Nuclear Information System (INIS)

    Riddington, J.W.; Reyer, R.J.

    1980-01-01

    A study of the causes of, and solutions for, recurrent valve failures has been performed. The frequency and root causes of valve problems were identified from licensee event reports and meetings with utility, NSSS, and valve manufacturer personnel. Three generic problems (stem leakage, seat leakage, and inadequate specification) and four valve specific problems were identified. The four valve specific problems and their principal causes are: (1) BWR pilot operated safety relief valves (pilot valve leakage); (2) spring loaded safety relief valves (water solid and two-phase flow behavior); (3) PWR feedwater regulating valves (trim degradation and packing failures); and (4) air operated solenoid valves (jamming due to foreign matter in service air). The first two valve specific problems are the subject of current industry programs. Programs intended to address stem leakage, seat leakage, timely exchange of valve failure information, testing of valves, and adequate specification, selection, and maintenance of valves will be outlined

  14. The estimate of ecological risk for ground ecosystems in case of nuclear power plant failures

    International Nuclear Information System (INIS)

    Kremlenkov, D.Y.; Kremlenkov, M.Y.

    2003-01-01

    Full text: The stochastic nature of radiation damage generates a need of forecasting information about possible consequences for environment and people. In this article it is given the estimate of probable damage to forest-and agricultural ecosystems from radionuclide emergency pollution in case of nuclear plant failures (for early emergency period). This estimate is based on radio-ecological risk conception which provide with the application of radioactive substances distribution models in atmosphere, as were calculation of absorbent radiation dose in critical ecosystem groups-calculation of probable area of lost ecosystems has been done by using the program written in Pascal. The quantitative estimate of environmental loss has been conducted for diverse classes of atmospheric stability. The value of ecological dose range (ELD) to coniferous forest is 30 Gy, deciduous forest - 300 Gy, agricultural crop - 60 Gy. The value of minimum ecological dose range (MELD) for all ecosystems is 10 Gy. In dose spread from MELD to ELD the ecological damage is proportional to absorbed dose. The ecological damage to ground ecosystems caused by cesium-137 and strontium-90 emergency pollution is primarily depended on the scale of radionuclide emergency pollution as well as weather conditions and radio-stability of critical vegetal ecosystem groups. On the assumption of a dose spread from MELD to ELD, ecological risk defined in probable ecosystem's destruction area is estimated: for cesium-137 pollution about 2 % of coniferous forest and from 4 to 9 % of deciduous forest; for strontium-90 pollution from 2 to 4 % of agricultural crop. As the scale of cesium-137 emergency pollution rise from 10 4 to 10 5 Cu the probable damage determined in ecosystem's destruction area increase 12-19 times to coniferous forest ecosystem and 15-36 times to deciduous forest according to weather conditions. The probable damage to coniferous and deciduous forest rise 11-17 times in proportion as the scale

  15. Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-09-15

    direction of research, development and demonstration in this area. The technologies discussed in this project are intended to establish the state of the art in surveillance, diagnostics and prognostics (SDP) technologies for equipment and process health monitoring in nuclear facilities. It is also intended to identify technology gaps and research needs of the nuclear industry in the area of SDP. The report draws on the conventional SDP technologies, as well as the latest tools, algorithms and techniques that have emerged over the last few years, especially in enabling technologies including fast data acquisition, data storage, data qualification and data analysis algorithms, such as empirical and physical modelling techniques. These new tools have made it possible to identify problems earlier and with better resolution. The significance of the material presented in this report is that it contributes not only to the current needs of the nuclear industry but also to the design improvements of the next generation of reactors. For example, the nuclear industry is currently striving to operate the plants for up to 80 years or more, as the value of nuclear assets has risen in recent years, resulting partly from environmental concerns with fossil energy production, as well as increased future demand for base load electricity. This long term operation (LTO) or life extension goal of the nuclear industry has stimulated renewed interest in more frequent monitoring of equipment to guard against ageing effects, not to mention the economic benefits that SDP implementation can produce, and contributions to radiation exposure that is as low as reasonably achievable, reduction of human errors, and optimized maintenance. Together with capabilities that enhance situational awareness, the technologies described in this report will enable more holistic management of plant structures, systems and components (SSCs), maintain high capacity factor in LTO and enable higher levels of safe operation

  16. Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2013-01-01

    direction of research, development and demonstration in this area. The technologies discussed in this project are intended to establish the state of the art in surveillance, diagnostics and prognostics (SDP) technologies for equipment and process health monitoring in nuclear facilities. It is also intended to identify technology gaps and research needs of the nuclear industry in the area of SDP. The report draws on the conventional SDP technologies, as well as the latest tools, algorithms and techniques that have emerged over the last few years, especially in enabling technologies including fast data acquisition, data storage, data qualification and data analysis algorithms, such as empirical and physical modelling techniques. These new tools have made it possible to identify problems earlier and with better resolution. The significance of the material presented in this report is that it contributes not only to the current needs of the nuclear industry but also to the design improvements of the next generation of reactors. For example, the nuclear industry is currently striving to operate the plants for up to 80 years or more, as the value of nuclear assets has risen in recent years, resulting partly from environmental concerns with fossil energy production, as well as increased future demand for base load electricity. This long term operation (LTO) or life extension goal of the nuclear industry has stimulated renewed interest in more frequent monitoring of equipment to guard against ageing effects, not to mention the economic benefits that SDP implementation can produce, and contributions to radiation exposure that is as low as reasonably achievable, reduction of human errors, and optimized maintenance. Together with capabilities that enhance situational awareness, the technologies described in this report will enable more holistic management of plant structures, systems and components (SSCs), maintain high capacity factor in LTO and enable higher levels of safe operation

  17. Comparison between Japan and the United States in the frequency of events in equipment and components at nuclear power plants

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2007-01-01

    The Institute of Nuclear Safety System, Incorporated (INSS) conducted trend analyses until 2005 to compare the frequency of events in certain electrical components and instrumentation components at nuclear power plants between Japan and the United States. The results revealed that events have occurred approximately an order of magnitude less often in Japan than in the United States. This paper compared Japan and the United States in more detail in terms of how often events - events reported under the reporting standards of the Nuclear Information Archive (NUCIA) or the Institute of Nuclear Power Operations (INPO) - occurred in electrical components, instrumentation components and mechanical components at nuclear power plants. The results were as follows: (1) In regard to electrical components and instrumentation components, events have occurred one-eighth less frequently in Japan than in the United States, suggesting that the previous results were correct. (2) Events have occurred more often in mechanical components than electrical components and instrumentation components in both Japan and the United States, and there was a smaller difference in the frequency of events in mechanical components between the two countries. (3) Regarding mechanical components, it was found that events in the pipes for critical systems and equipment, such as reactor coolant systems, emergency core cooling systems, instrument and control systems, ventilating and air-conditioning systems, and turbine equipment, have occurred more often in Japan than in the United States. (4) The above observations suggest that there is little scope for reducing the frequency of events in electrical components and instrumentation components, but that mechanical components such as pipes for main systems like emergency core cooling systems and turbine equipment in the case of PWRs, could be improved by re-examining inspection methods and intervals. (author)

  18. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

    International Nuclear Information System (INIS)

    2001-02-01

    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, design 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 wearout 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) including the Soviet designed water moderated and 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 also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

  19. Functional components of the bacterial CzcCBA efflux system reduce cadmium uptake and accumulation in transgenic tobacco plants.

    Science.gov (United States)

    Nesler, Andrea; DalCorso, Giovanni; Fasani, Elisa; Manara, Anna; Di Sansebastiano, Gian Pietro; Argese, Emanuele; Furini, Antonella

    2017-03-25

    Cadmium (Cd) is a toxic trace element released into the environment by industrial and agricultural practices, threatening the health of plants and contaminating the food/feed chain. Biotechnology can be used to develop plant varieties with a higher capacity for Cd accumulation (for use in phytoremediation programs) or a lower capacity for Cd accumulation (to reduce Cd levels in food and feed). Here we generated transgenic tobacco plants expressing components of the Pseudomonas putida CzcCBA efflux system. Plants were transformed with combinations of the CzcC, CzcB and CzcA genes, and the impact on Cd mobilization was analysed. Plants expressing PpCzcC showed no differences in Cd accumulation, whereas those expressing PpCzcB or PpCzcA accumulated less Cd in the shoots, but more Cd in the roots. Plants expressing both PpCzcB and PpCzcA accumulated less Cd in the shoots and roots compared to controls, whereas plants expressing all three genes showed a significant reduction in Cd levels only in shoots. These results show that components of the CzcCBA system can be expressed in plants and may be useful for developing plants with a reduced capacity to accumulate Cd in the shoots, potentially reducing the toxicity of food/feed crops cultivated in Cd-contaminated soils. Copyright © 2016 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

    Conceicao Junior, Osmar

    2009-01-01

    This study consists on the application of the Failure Modes and EffectsAnalysis (FMEA), a hazard identification and a risk assessment technique, tothe Emergency Cooling System (ECS) of an experimental nuclear power plant,which is responsible for mitigating the consequences of an eventual loss ofcoolant accident on the Pressurized Water Reactor (PWR). Such analysisintends to identify possible weaknesses on the design of the system andpropose some improvements in order to maximize its reliability. To achievethis goal a detailed study of the system was carried on (through itstechnical documentation), the correspondent reliability block diagram wasobtained, the FMEA analysis was executed and, finally, some suggestions werepresented. (author)