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Sample records for assessing reactor reliability

  1. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual

    International Nuclear Information System (INIS)

    Gilbert, B.G.; Reece, W.J.; Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.

    1990-12-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal computer. NUCLARR can furnish the end user with data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented is a five-volume series of reports. Volume V: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software. This document is organized in three sections. Part 1 is the summary description, which presents an overview of the NUCLARR system and data processing procedures. Part 2 contains all data and information relevant to the human error probability (HEP) data side of NUCLARR. Data and information for the hardware component failure data (HCFD) side are presented in Part 3. 7 refs

  2. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)

    International Nuclear Information System (INIS)

    Gilbert, B.G.; Reece, W.J.; Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.

    1990-12-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal computer. NUCLARR can furnish the end user with data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume 5: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software. This document is organized in three sections. Part 1 is the summary description, which presents an overview of the NUCLARR system and data processing procedures. Part 2 contains all data and information relevant to the human error probability (HEP) data side of NUCLARR. Data and information for the hardware component failure data (HCFD) side are presented in Part 3. 7 refs

  3. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual

    International Nuclear Information System (INIS)

    Gilbert, B.G.; Reece, W.J.; Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.

    1990-12-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal computer. NUCLARR can furnish the end user with data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume V: Data Manual provides a hard-copy representation of all data and related information available within the NUCLARR system software. This document is organized in three sections. Part 1 is the summary description, which presents an overview of the NUCLARR system and data processing procedures. Part 2 contains all data and information relevant to the human error probability (HEP) side of NUCLARR. Data and information for the hardware component failure data (HCFD) side are presented in Part 3. 7 refs., 1 fig

  4. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Summary description

    International Nuclear Information System (INIS)

    Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.; Groh, M.R.; Gentillon, C.D.; Gilbert, B.G.; Reece, W.J.

    1990-05-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automate data base management system for storing and processing human error probability and hardware component failure rate data. The NUCLARR system software resides on an IBM (or compatible) personal microcomputer. NUCLARR can be accessed by the end user to furnish data suitable for input in human and/or hardware reliability analysis to support a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. This document Volume 1, of this series is the Summary Description, which presents an overview of the data management system, including a description of data collection, data qualification, data structure, and taxonomies. Programming activities, procedures for processing data, a user's guide, and hard copy data manual are presented in Volumes 2 through 5, NUREG/CR-4639

  5. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  6. Reliability assessment of the fueling machine of the CANDU reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.

    1985-01-01

    Fueling of CANDU-reactors is carried out by two fueling machines, each serving one end of the reactor. The fueling machine becomes a part of the primary heat transport system during the refueling operations, and hence, some refueling machine malfunctions could result in a small scale-loss-of-coolant accident. Fueling machine failures and the failure sequences are discussed. The unavailability of the fueling machine is estimated by using fault tree analysis. The probability of mechanical failure of the fueling machine interface is estimated as 1.08 x 10 -5 . (orig.) [de

  7. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh, E-mail: mukeshd@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarty, Aranyak [School of Nuclear Studies and Application, Jadavpur University, Kolkata 700032 (India); Nayak, A.K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Prasad, Hari; Gopika, V. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-10-15

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

  8. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Chakravarty, Aranyak; Nayak, A.K.; Prasad, Hari; Gopika, V.

    2014-01-01

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components

  9. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Werner, James Elmer [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; McKellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Kennedy, John Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Biersdorf, John Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division

    2017-04-01

    The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

  10. Nuclear computerized library for assessing reactor reliability (NUCLARR): User's guide: Part 3, NUCLARR system description

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gentillon, C.D.; Gertman, D.I.; Beers, G.H.; Galyean, W.J.; Gilbert, B.G.

    1988-06-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. NUCLARR can be used by the end user to furnish data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume IV of this series is the User's Guide for operating the NUCLARR software and is presented in three parts. This document, Part 3: NUCLARR System Description, provides an in-depth discussion of the design characteristics and special features of the NUCLARR software. Part 3 also presents the organization of the data base structures and techniques used to manipulate the data

  11. Nuclear computerized library for assessing reactor reliability (NUCLARR): Part 2, Guide to operations: User's guide

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gentillon, C.D.; Gertman, D.I.; Beers, G.H.; Galyean, W.J.; Gilbert, B.G.

    1988-06-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. NUCLARR can be used by the end user to furnish data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume IV of this series is the User's Guide for operating the NUCLARR software and is presented in three parts. This volume, Part 2: Guide to Operations, contains the instructions and basic procedures for using the NUCLARR software. Part 2 provides guidance and information for getting started, performing the desired functions, and making the most efficient use of the system's features

  12. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Volume 1, Summary description

    International Nuclear Information System (INIS)

    Gertman, D.I.; Gilmore, W.E.; Galyean, W.J.; Groh, M.R.; Gentillon, C.D.; Gilbert, B.G.

    1988-02-01

    A data management system has been implemented which supports a variety of risk-related analyses and provides a repository of hardware component failure and human error probability data to the risk analyst. The Nuclear Computerized Library for Assessing Reactor Reliability, NUCLARR, is an interactive, graphically oriented system which resides on a personal computer (PC) or PC-compatible environment. An overview of the data management system, including a description of data collection, specification, data structure, and taxonomies, is presented in Volume I of this report. Programming activities, procedures for processing data, user's guide, and hard copy data manual are presented in Volumes II through V

  13. Reliability assessment of emergency exhaust system in a pool-type research reactor

    International Nuclear Information System (INIS)

    Khan, S.A.

    1991-01-01

    The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations. The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performend on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundance in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g. simultaneous compressed air and electric power supply failure. Based upon the results obtained, some recommendations are made. (orig.)

  14. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    International Nuclear Information System (INIS)

    Vierow, Karen; Aldemir, Tunc

    2009-01-01

    The project entitled, 'Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors', was conducted as a DOE NERI project collaboration between Texas A and M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  15. Nuclear computerized library for assessing reactor reliability (NUCLARR): Data manual: Part 1, Summary description

    International Nuclear Information System (INIS)

    Gertman, D.I.; Gilbert, B.G.; Gilmore, W.E.; Galyean, W.J.

    1988-06-01

    This volume of a five-volume series summarizes those data currently resident in the first release of the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) data base. The raw human error probability (HEP) and hardware component failure data (HCFD) contained herein are accompanied by a glossary of terms and the HEP and hardware taxonomies used to structure the data. Instructions are presented on how the user may navigate through the NUCLARR data management system to find anchor values to assist in solving risk-related problems. Volume V: Data Manual will be updated on a periodic basis so that risk analysis without access to a computer may have access to the largest NUCLARR data. This document Part 1 of Volume 5 introduces aspects of the NUCLARR data base management system and prepares the reader for reviewing data in other Parts of Volume 5

  16. Reliability of reactor materials

    International Nuclear Information System (INIS)

    Toerroenen, K.; Aho-Mantila, I.

    1986-05-01

    This report is the final technical report of the fracture mechanics part of the Reliability of Reactor Materials Programme, which was carried out at the Technical Research Centre of Finland (VTT) through the years 1981 to 1983. Research and development work was carried out in five major areas, viz. statistical treatment and modelling of cleavage fracture, crack arrest, ductile fracture, instrumented impact testing as well as comparison of numerical and experimental elastic-plastic fracture mechanics. In the area of cleavage fracture the critical variables affecting the fracture of steels are considered in the frames of a statistical model, so called WST-model. Comparison of fracture toughness values predicted by the model and corresponding experimental values shows excellent agreement for a variety of microstructures. different posibilities for using the model are discussed. The development work in the area of crack arrest testing was concentrated in the crack starter properties, test arrangement and computer control. A computerized elastic-plastic fracture testing method with a variety of test specimen geometries in a large temperature range was developed for a routine stage. Ductile fracture characteristics of reactor pressure vessel steel A533B and comparable weld material are given. The features of a new, patented instrumented impact tester are described. Experimental and theoretical comparisons between the new and conventional testers indicated clearly the improvements achieved with the new tester. A comparison of numerical and experimental elastic-plastic fracture mechanics capabilities at VTT was carried out. The comparison consisted of two-dimensional linear elastic as well as elastic-plastic finite element analysis of four specimen geometries and equivalent experimental tests. (author)

  17. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Aldemir, Tunc

    2009-09-10

    The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  18. Probabilistic safety assessment of Tehran Research Reactor using systems analysis programs for hands-on integrated reliability evaluations

    International Nuclear Information System (INIS)

    Hosseini, M.H.; Nematollahi, M.R.; Sepanloo, K.

    2004-01-01

    Probabilistic safety assessment application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this document the application of the probabilistic safety assessment to the Tehran Research Reactor is presented. The level 1 practicabilities safety assessment application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantifications, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using systems analysis programs for hands-on integrated reliability evaluations software

  19. Assessment of FBR MONJU accident management reliability in causing reactor trips

    International Nuclear Information System (INIS)

    Sotsu, Masutake; Kurisaka, Kenichi

    2010-01-01

    This paper describes a method and application of quantitatively evaluating Accident Management (AM) reliability upon a reactor trip failure for the MONJU fast breeder reactor using a PSA technique. The present method comprises an allowable time estimation that is based on plant transient response analysis using the Super-COPD code that was developed for use in best estimates of the plant dynamics of MONJU and in estimating failure probability of operator's actions in AMs within the allowable time based on time records obtained from simulator training. Application of this method to MONJU resulted in the estimation that the allowable time for an unprotected loss-of-heat sink event would be more than the longest observed time of 326 s. The corresponding operation failure probability would be less than 0.1 even after taking the uncertainty into consideration. Combining this with a level 1 PSA revealed that the total frequency of core damage accompanying a reactor trip failure at MONJU could be decreased by at least 50 percent due to the reactor trip AM. (author)

  20. 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

  1. Reliability Assessment Method of Reactor Protection System Software by Using V and Vbased Bayesian Nets

    International Nuclear Information System (INIS)

    Eom, H. S.; Park, G. Y.; Kang, H. G.; Son, H. S.

    2010-07-01

    Developed a methodology which can be practically used in quantitative reliability assessment of a safety c ritical software for a protection system of nuclear power plants. The base of the proposed methodology is V and V being used in the nuclear industry, which means that it is not affected with specific software development environments or parameters that are necessary for the reliability calculation. Modular and formal sub-BNs in the proposed methodology is useful tool to constitute the whole BN model for reliability assessment of a target software. The proposed V and V based BN model estimates the defects in the software according to the performance of V and V results and then calculate reliability of the software. A case study was carried out to validate the proposed methodology. The target software is the RPS SW which was developed by KNICS project

  2. Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

    International Nuclear Information System (INIS)

    Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

    1988-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

  3. Reliability studies in research reactors

    International Nuclear Information System (INIS)

    Albuquerque, Tob Rodrigues de

    2013-01-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  4. Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

    Science.gov (United States)

    Ha, Taesung

    A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential

  5. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  6. Reliability assessment on decay heat removal system of a fast reactor

    International Nuclear Information System (INIS)

    Hioki, Kazumasa

    1991-01-01

    The reliability of a decay heat removal system (DHRS) is influenced by the success criteria, the components which constitute the system, the support systems configuration, and the mission time. Assessments were performed to investigate quantitatively the effects of these items. Failure probabilities of DHRS under forced or natural circulation modes were calculated and then components and systems of large importance for each mode were identified. (author)

  7. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Part 1, Overview of NUCLARR data retrieval: User's guide

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gentillon, C.D.; Gertman, D.I.; Beers, G.H.; Galyean, W.J.; Gilbert, B.G.

    1988-06-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. NUCLARR can be used by the end user to furnish data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume IV of this series is the User's Guide for operating the NUCLARR software and is presented in three parts. This document, Part 1: Overview of NUCLARR Data Retrieval provides an introductory overview to the system's capabilities and procedures for data retrieval. The methods and criteria for selection of data sources and entering them into the NUCLARR system are also described in this document

  8. Reliability assurance for regulation of advanced reactors

    International Nuclear Information System (INIS)

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-01-01

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

  9. Reliability assurance for regulation of advanced reactors

    International Nuclear Information System (INIS)

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1992-01-01

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. this paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

  10. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    International Nuclear Information System (INIS)

    Touati, Said; Chennai, Salim; Souli, Aissa

    2015-01-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  11. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    Energy Technology Data Exchange (ETDEWEB)

    Touati, Said; Chennai, Salim; Souli, Aissa [Nuclear Research Centre of Birine, Ain Oussera, Djelfa Province (Algeria)

    2015-07-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  12. FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

    Full Text Available ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN

  13. Safety and reliability assessment

    International Nuclear Information System (INIS)

    1979-01-01

    This report contains the papers delivered at the course on safety and reliability assessment held at the CSIR Conference Centre, Scientia, Pretoria. The following topics were discussed: safety standards; licensing; biological effects of radiation; what is a PWR; safety principles in the design of a nuclear reactor; radio-release analysis; quality assurance; the staffing, organisation and training for a nuclear power plant project; event trees, fault trees and probability; Automatic Protective Systems; sources of failure-rate data; interpretation of failure data; synthesis and reliability; quantification of human error in man-machine systems; dispersion of noxious substances through the atmosphere; criticality aspects of enrichment and recovery plants; and risk and hazard analysis. Extensive examples are given as well as case studies

  14. Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

    International Nuclear Information System (INIS)

    2014-09-01

    Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

  15. Dynamic reliability and risk assessment of the accident localization system of the Ignalina NPP RBMK-1500 reactor

    International Nuclear Information System (INIS)

    Kopustinskas, V.; Augutis, J.; Rimkevicius, S.

    2005-01-01

    The paper presents reliability and risk analysis of the RBMK-1500 reactor accident localization system (ALS) (confinement), which prevents radioactive releases to the environment. Reliability of the system was estimated and compared by two methods: the conventional fault tree method and an innovative dynamic reliability model, based on stochastic differential equations. Frequency of radioactive release through ALS was also estimated. The results of the study indicate that conventional fault tree modeling techniques in this case apply high degree of conservatism in the system reliability estimates. One of the purposes of the ALS reliability study was to demonstrate advantages of the dynamic reliability analysis against the conventional fault/event tree methods. The Markovian framework to deal with dynamic aspects of system behavior is presented. Although not analyzed in detail, the framework is also capable of accounting for non-constant component failure rates. Computational methods are proposed to solve stochastic differential equations, including analytical solution, which is possible only for relatively small and simple systems. Other numerical methods, like Monte Carlo and numerical schemes of differential equations are analyzed and compared. The study is finalized with concluding remarks regarding both the studied system reliability and computational methods used

  16. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 2: Human error probability (HEP) data; Volume 5, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Reece, W.J.; Gilbert, B.G.; Richards, R.E. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    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.

  17. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 2: Human error probability (HEP) 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

  18. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual, Part 2: Human Error Probability (HEP) 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

  19. Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

    International Nuclear Information System (INIS)

    Smith, M.S.; Ragland, W.A.

    1989-01-01

    The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

  20. Reliability test for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Uchiyama, Junichi

    1998-01-01

    41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

  1. Mechanical development for reliable reactor components

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Metcalfe, R.

    1983-09-01

    The CANDU reactor has achieved worldwide distinction because of its reliable performance. To achieve this, special attention was given to the reliability and maintainability of components in the heavy water circuits. Development programs were initiated early in the history of the CANDU reactor to improve the effectiveness of pump seals, valves, and static seals because of unacceptable performance of the commercial equipment then available. As a result, pump seals with a five year life now appear achievable, and valves and static seals are no longer a significant concern in CANDU reactors. Increasing effort is being given remotely operated tools and fabrication systems for radioactive environments

  2. Human Reliability Analysis for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; David I. Gertman

    2012-06-01

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  3. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  4. Approach to reliability assessment

    International Nuclear Information System (INIS)

    Green, A.E.; Bourne, A.J.

    1975-01-01

    Experience has shown that reliability assessments can play an important role in the early design and subsequent operation of technological systems where reliability is at a premium. The approaches to and techniques for such assessments, which have been outlined in the paper, have been successfully applied in variety of applications ranging from individual equipments to large and complex systems. The general approach involves the logical and systematic establishment of the purpose, performance requirements and reliability criteria of systems. This is followed by an appraisal of likely system achievment based on the understanding of different types of variational behavior. A fundamental reliability model emerges from the correlation between the appropriate Q and H functions for performance requirement and achievement. This model may cover the complete spectrum of performance behavior in all the system dimensions

  5. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    Schatz, R.A.; Duetsch, K.L.

    1974-01-01

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  6. power system reliability in supplying nuclear reactors

    International Nuclear Information System (INIS)

    Gad, M.M.M.

    2007-01-01

    this thesis presents a simple technique for deducing minimal cut set (MCS) from the defined minimal path set (MPS) of generic distribution system and this technique have been used to evaluate the basic reliability indices of Egypt's second research reactor (ETRR-2) electrical distribution network. the alternative system configurations are then studied to evaluate their impact on service reliability. the proposed MCS approach considers both sustained and temporary outage. the temporary outage constitutes an important parameter in characterizing the system reliability indices for critical load point in distribution system. it is also consider the power quality impact on the reliability indices

  7. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  8. Use of RMPS to assess the reliability of Passive Safety Systems in CAREM-like reactor, past and present experiences. Second progress report

    International Nuclear Information System (INIS)

    Giménez, M; Mezio, F.; Zanocco, P.; Lorenzo, G.

    2011-01-01

    Conclusions: • RMPS is being used successfully to assess the fulfillment of design criteria from a probabilistic point of view, in case of LOHS and LOCA, considering uncertainties in the reactor, in the passive safety systems and in the models as well. • Allows to quantify the probability of Event Tree headers related to some systems whose demand depends on the accidental sequence evolution (i.e. probability to demand a safety valve in case of a LOHS with success of the PRHRS, but working under deteriorated conditions). • Functional reliability quantification not already used in CAREM PSA, (Fault Trees or in Event Trees?)

  9. Reliability analysis of the reactor protection system with fault diagnosis

    International Nuclear Information System (INIS)

    Lee, D.Y.; Han, J.B.; Lyou, J.

    2004-01-01

    The main function of a reactor protection system (RPS) is to maintain the reactor core integrity and reactor coolant system pressure boundary. The RPS consists of the 2-out-of-m redundant architecture to assure a reliable operation. The system reliability of the RPS is a very important factor for the probability safety assessment (PSA) evaluation in the nuclear field. To evaluate the system failure rate of the k-out-of-m redundant system is not so easy with the deterministic method. In this paper, the reliability analysis method using the binomial process is suggested to calculate the failure rate of the RPS system with a fault diagnosis function. The suggested method is compared with the result of the Markov process to verify the validation of the suggested method, and applied to the several kinds of RPS architectures for a comparative evaluation of the reliability. (orig.)

  10. Uncertainties and reliability theories for reactor safety

    International Nuclear Information System (INIS)

    Veneziano, D.

    1975-01-01

    What makes the safety problem of nuclear reactors particularly challenging is the demand for high levels of reliability and the limitation of statistical information. The latter is an unfortunate circumstance, which forces deductive theories of reliability to use models and parameter values with weak factual support. The uncertainty about probabilistic models and parameters which are inferred from limited statistical evidence can be quantified and incorporated rationally into inductive theories of reliability. In such theories, the starting point is the information actually available, as opposed to an estimated probabilistic model. But, while the necessity of introducing inductive uncertainty into reliability theories has been recognized by many authors, no satisfactory inductive theory is presently available. The paper presents: a classification of uncertainties and of reliability models for reactor safety; a general methodology to include these uncertainties into reliability analysis; a discussion about the relative advantages and the limitations of various reliability theories (specifically, of inductive and deductive, parametric and nonparametric, second-moment and full-distribution theories). For example, it is shown that second-moment theories, which were originally suggested to cope with the scarcity of data, and which have been proposed recently for the safety analysis of secondary containment vessels, are the least capable of incorporating statistical uncertainty. The focus is on reliability models for external threats (seismic accelerations and tornadoes). As an application example, the effect of statistical uncertainty on seismic risk is studied using parametric full-distribution models

  11. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

    2000-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  12. Reliability tests for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

    1998-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  13. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Guide to data processing and revision: Part 2, Human error probability data entry and revision procedures

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gertman, D.I.; Gilbert, B.G.; Reece, W.J.

    1988-11-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability (HEP) and hardware component failure data (HCFD). The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. Users can perform data base searches to furnish HEP estimates and HCFD rates. In this manner, the NUCLARR system can be used to support a variety of risk assessment activities. This volume, Volume 3 of a 5-volume series, presents the procedures used to process HEP and HCFD for entry in NUCLARR and describes how to modify the existing NUCLARR taxonomy in order to add either equipment types or action verbs. Volume 3 also specifies the various roles of the administrative staff on assignment to the NUCLARR Clearinghouse who are tasked with maintaining the data base, dealing with user requests, and processing NUCLARR data. 5 refs., 34 figs., 3 tabs

  14. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Guide to data processing and revision: Part 3, Hardware component failure data entry and revision procedures

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Gertman, D.I.; Gilbert, B.G.; Reece, W.J.

    1988-11-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability (HEP) and hardware component failure data (HCFD). The NUCLARR system software resides on an IBM (or compatible) personal micro-computer. Users can perform data base searches to furnish HEP estimates and HCFD rates. In this manner, the NUCLARR system can be used to support a variety of risk assessment activities. This volume, Volume 3 of a 5-volume series, presents the procedures used to process HEP and HCFD for entry in NUCLARR and describes how to modify the existing NUCLARR taxonomy in order to add equipment types of action verbs. Volume 3 also specifies the various roles of the administrative staff on assignment to the NUCLARR Clearinghouse who are tasked with maintaining the data base, dealing with user requests, and processing NUCLARR data

  15. CRL research reactor diesel generator reliability study 1960 - 1992

    International Nuclear Information System (INIS)

    Winfield, D.J.; McCauley, G.M.

    1994-07-01

    A data base has been provided for the Chalk River Laboratories (CRL) research reactor diesel generator reliability, for use in risk assessment studies of CRL research reactors. Data from 1960 to end of 1992 have been collected, representing 358 diesel generator years of experience. The data is used to provide failure-to-start probabilities and failure-to-run rates. Data is also classified according to subsystem failures, multiple failures and common cause failures. Comparisons with other recent studies of nuclear power plant diesel generator reliability have been made. This revision updates the 1989 September report. (author). 14 refs., 13 tabs., 10 figs

  16. CRNL research reactor diesel generator reliability study 1960-1985

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1989-09-01

    A data base has been provided for the CRNL research reactor diesel generator reliability, for use in risk assessment studies of CRNL research reactors. Data from 1960 to the present have been collected, representing 281 diesel generator years of experience. The data is used to provide failure-to-start probabilities and failure-to-run rates. Data is also classified according to subsystem failures, multiple failures and common cause failures. Comparisons with other recent studies of nuclear power plant diesel generator reliability have been made

  17. Reliability analysis of reactor protection systems

    International Nuclear Information System (INIS)

    Alsan, S.

    1976-07-01

    A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)

  18. Software reliability assessment

    International Nuclear Information System (INIS)

    Barnes, M.; Bradley, P.A.; Brewer, M.A.

    1994-01-01

    The increased usage and sophistication of computers applied to real time safety-related systems in the United Kingdom has spurred on the desire to provide a standard framework within which to assess dependable computing systems. Recent accidents and ensuing legislation have acted as a catalyst in this area. One particular aspect of dependable computing systems is that of software, which is usually designed to reduce risk at the system level, but which can increase risk if it is unreliable. Various organizations have recognized the problem of assessing the risk imposed to the system by unreliable software, and have taken initial steps to develop and use such assessment frameworks. This paper relates the approach of Consultancy Services of AEA Technology in developing a framework to assess the risk imposed by unreliable software. In addition, the paper discusses the experiences gained by Consultancy Services in applying the assessment framework to commercial and research projects. The framework is applicable to software used in safety applications, including proprietary software. Although the paper is written with Nuclear Reactor Safety applications in mind, the principles discussed can be applied to safety applications in all industries

  19. Operator reliability assessment system (OPERAS)

    International Nuclear Information System (INIS)

    Singh, A.; Spurgin, A.J.; Martin, T.; Welsch, J.; Hallam, J.W.

    1991-01-01

    OPERAS is a personal-computer (PC) based software to collect and process simulator data on control-room operators responses during requalification training scenarios. The data collection scheme is based upon approach developed earlier during the EPRI Operator Reliability Experiments project. The software allows automated data collection from simulator, thus minimizing simulator staff time and resources to collect, maintain and process data which can be useful in monitoring, assessing and enhancing the progress of crew reliability and effectiveness. The system is designed to provide the data and output information in the form of user-friendly charts, tables and figures for use by plant staff. OPERAS prototype software has been implemented at the Diablo Canyon (PWR) and Millstone (BWR) plants and is currently being used to collect operator response data. Data collected from similator include plant-state variables such as reactor pressure and temperature, malfunction, times at which annunciators are activated, operator actions and observations of crew behavior by training staff. The data and systematic analytical results provided by the OPERAS system can contribute to increase objectivity by the utility probabilistic risk analysis (PRA) and training staff in monitoring and assessing reliability of their crews

  20. Swimming pool reactor reliability and safety analysis

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1997-01-01

    A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

  1. Manual on reliability data collection for research reactor PSAs

    Energy Technology Data Exchange (ETDEWEB)

    1992-01-01

    The IAEA has been actively promoting performance of probabilistic safety assessment (PSA) studies for research reactors. From 1986 to 1988 the IAEA undertook a Coordinated Research Programme (CRP) on PSA for research reactors which helped promote use of PSA and foster a broad exchange of information. Although the basic methodological approach in performing a research reactor PSA is understood, some unresolved issues, data availability being among them, still exist. To address the issue on the international level, the IAEA initiated a new CRP on ``Data Acquisition for Research Reactors PSA Studies``. The aim of the CRP is to develop a data collection system and generate research reactor specific reliability data for use in PSAs. The achieve this aim a set of precise definitions should be adopted. A set of definitions developed specifically for research reactors and covering classification of equipment and failure terms, reliability parameters, failure modes and other terms necessary for data collection and processing is presented in this document which is based on discussions during the first meeting of the CRP held in Vienna in October 1989 and during the second meeting held in Beijing, China, in October 1990. Refs and figs.

  2. Manual on reliability data collection for research reactor PSAs

    International Nuclear Information System (INIS)

    1992-01-01

    The IAEA has been actively promoting performance of probabilistic safety assessment (PSA) studies for research reactors. From 1986 to 1988 the IAEA undertook a Coordinated Research Programme (CRP) on PSA for research reactors which helped promote use of PSA and foster a broad exchange of information. Although the basic methodological approach in performing a research reactor PSA is understood, some unresolved issues, data availability being among them, still exist. To address the issue on the international level, the IAEA initiated a new CRP on ''Data Acquisition for Research Reactors PSA Studies''. The aim of the CRP is to develop a data collection system and generate research reactor specific reliability data for use in PSAs. The achieve this aim a set of precise definitions should be adopted. A set of definitions developed specifically for research reactors and covering classification of equipment and failure terms, reliability parameters, failure modes and other terms necessary for data collection and processing is presented in this document which is based on discussions during the first meeting of the CRP held in Vienna in October 1989 and during the second meeting held in Beijing, China, in October 1990. Refs and figs

  3. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  4. Reactor safety assessment system

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  5. OSS reliability measurement and assessment

    CERN Document Server

    Yamada, Shigeru

    2016-01-01

    This book analyses quantitative open source software (OSS) reliability assessment and its applications, focusing on three major topic areas: the Fundamentals of OSS Quality/Reliability Measurement and Assessment; the Practical Applications of OSS Reliability Modelling; and Recent Developments in OSS Reliability Modelling. Offering an ideal reference guide for graduate students and researchers in reliability for open source software (OSS) and modelling, the book introduces several methods of reliability assessment for OSS including component-oriented reliability analysis based on analytic hierarchy process (AHP), analytic network process (ANP), and non-homogeneous Poisson process (NHPP) models, the stochastic differential equation models and hazard rate models. These measurement and management technologies are essential to producing and maintaining quality/reliable systems using OSS.

  6. Reliability Assessment Of Wind Turbines

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard

    2014-01-01

    Reduction of cost of energy for wind turbines are very important in order to make wind energy competitive compared to other energy sources. Therefore the turbine components should be designed to have sufficient reliability but also not be too costly (and safe). This paper presents models...... for uncertainty modeling and reliability assessment of especially the structural components such as tower, blades, substructure and foundation. But since the function of a wind turbine is highly dependent on many electrical and mechanical components as well as a control system also reliability aspects...... of these components are discussed and it is described how there reliability influences the reliability of the structural components. Two illustrative examples are presented considering uncertainty modeling, reliability assessment and calibration of partial safety factors for structural wind turbine components exposed...

  7. New Approaches to Reliability Assessment

    DEFF Research Database (Denmark)

    Ma, Ke; Wang, Huai; Blaabjerg, Frede

    2016-01-01

    of energy. New approaches for reliability assessment are being taken in the design phase of power electronics systems based on the physics-of-failure in components. In this approach, many new methods, such as multidisciplinary simulation tools, strength testing of components, translation of mission profiles......, and statistical analysis, are involved to enable better prediction and design of reliability for products. This article gives an overview of the new design flow in the reliability engineering of power electronics from the system-level point of view and discusses some of the emerging needs for the technology...

  8. Reliability Assessment of Concrete Bridges

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle; Middleton, C. R.

    This paper is partly based on research performed for the Highways Agency, London, UK under the project DPU/9/44 "Revision of Bridge Assessment Rules Based on Whole Life Performance: concrete bridges". It contains the details of a methodology which can be used to generate Whole Life (WL) reliability...... profiles. These WL reliability profiles may be used to establish revised rules for concrete bridges. This paper is to some extend based on Thoft-Christensen et. al. [1996], Thoft-Christensen [1996] et. al. and Thoft-Christensen [1996]....

  9. Reliability assessment of Wind turbines

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard

    2015-01-01

    Wind turbines can be considered as structures that are in between civil engineering structures and machines since they consist of structural components and many electrical and machine components together with a control system. Further, a wind turbine is not a one-of-a-kind structure...... but manufactured in series production based on many component tests, some prototype tests and zeroseries wind turbines. These characteristics influence the reliability assessment where focus in this paper is on the structural components. Levelized Cost Of Energy is very important for wind energy, especially when...... comparing to other energy sources. Therefore much focus is on cost reductions and improved reliability both for offshore and onshore wind turbines. The wind turbine components should be designed to have sufficient reliability level with respect to both extreme and fatigue loads but also not be too costly...

  10. Operational human performance reliability assessment (OHPRA)

    International Nuclear Information System (INIS)

    Haas, P.M.; Swanson, P.J.; Connelly, E.M.

    1993-01-01

    Operational Human Performance Reliability Assessment (OHPRA) is an approach for assessing human performance that is being developed in response to demands from modern process industries for practical and effective tools to assess and improve human performance, and therefore overall system performance and safety. The single most distinguishing feature of the approach is that is defines human performance in open-quotes operationalclose quotes terms. OHPRA is focused not on generation of human error probabilities, but on practical analysis of human performance to aid management in (1) identifying open-quotes fixableclose quotes problems and (2) providing input on the importance and nature of potential improvements. Development of the model in progress uses a unique approach for eliciting expert strategies for assessing performance. A PC-based model incorporating this expertise is planned. A preliminary version of the approach has already been used successfully to identify practical human performance problems in reactor and chemical process plant operations

  11. Reliability of reactor plant water cleanup pumps

    International Nuclear Information System (INIS)

    Pearson, J.L.

    1979-01-01

    Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

  12. Creation of reactor's reliable system of emergency energy supply

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Brovkin, A.Yu.; Petukhov, V.K.; Chekushin, A.I.; Chernyaev, V.P.; Yagotinets, N.A.

    1998-01-01

    System of reliable power supply of the WWR-K reactor complex is described, which completely provides safety operation of reactor equipment in the case of total voltage loss from external power transmission lines as well as under destruction of accumulation batteries by earthquake more than 6 balls. Switching on in operation of diesel-generators and system of constant current supply from accumulator batteries is occurred automatically under cessation of voltage supply from centralized power system. Reliable reactor dampening in case it work on capacity has been ensured. Reactor cooling under its emergency shutdown during both the partial or the total loss of coolant in first counter has been carried out. Under full coolant loss the system of emergency reactor cooling has been switched on in operation

  13. Review of the treat upgrade reactor scram system reliability analysis

    International Nuclear Information System (INIS)

    Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

    1984-10-01

    In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

  14. Reliability assessment of nuclear structural systems

    International Nuclear Information System (INIS)

    Reich, M.; Hwang, H.

    1983-01-01

    Reliability assessment of nuclear structural systems has been receiving more emphasis over the last few years. This paper deals with the recent progress made by the Structural Analysis Division of Brookhaven National Laboratory (BNL), in the development of a probability-based reliability analysis methodology for safety evaluation of reactor containments and other seismic category I structures. An important feature of this methodology is the incorporation of finite element analysis and random vibration theory. By utilizing this method, it is possible to evaluate the safety of nuclear structures under various static and dynamic loads in terms of limit state probability. Progress in other related areas, such as the establishment of probabilistic characteristics for various loads and structural resistance, are also described. Results of an application of the methodology to a realistic reinforced concrete containment subjected to dead and live loads, accidental internal pressures and earthquake ground accelerations are presented

  15. Optimizing the design and operation of reactor emergency systems using reliability analysis techniques

    International Nuclear Information System (INIS)

    Snaith, E.R.

    1975-01-01

    Following a reactor trip various reactor emergency systems, e.g. essential power supplies, emergency core cooling and boiler feed water arrangements are required to operate with a high degree of reliability. These systems must therefore be critically assessed to confirm their capability of operation and determine their reliability of performance. The use of probability analysis techniques enables the potential operating reliability of the systems to be calculated and this can then be compared with the overall reliability requirements. However, a system reliability analysis does much more than calculate an overall reliability value for the system. It establishes the reliability of all parts of the system and thus identifies the most sensitive areas of unreliability. This indicates the areas where any required improvements should be made and enables the overall systems' designs and modes of operation to be optimized, to meet the system and hence the overall reactor safety criteria. This paper gives specific examples of sensitive areas of unreliability that were identified as a result of a reliability analysis that was carried out on a reactor emergency core cooling system. Details are given of modifications to design and operation that were implemented with a resulting improvement in reliability of various reactor sub-systems. The report concludes that an initial calculation of system reliability should represent only the beginning of continuing process of system assessment. Data on equipment and system performance, particularly in those areas shown to be sensitive in their effect on the overall nuclear power plant reliability, should be collected and processed to give reliability data. These data should then be applied in further probabilistic analyses and the results correlated with the original analysis. This will demonstrate whether the required and the originally predicted system reliability is likely to be achieved, in the light of the actual history to date of

  16. Software reliability and safety in nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor

  17. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  18. Human reliability assessment and probabilistic risk assessment

    International Nuclear Information System (INIS)

    Embrey, D.E.; Lucas, D.A.

    1989-01-01

    Human reliability assessment (HRA) is used within Probabilistic Risk Assessment (PRA) to identify the human errors (both omission and commission) which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. There exist a variey of HRA techniques and the selection of an appropriate one is often difficult. This paper reviews a number of available HRA techniques and discusses their strengths and weaknesses. The techniques reviewed include: decompositional methods, time-reliability curves and systematic expert judgement techniques. (orig.)

  19. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  20. Considerations concerning the reliability of reactor safety equipment

    International Nuclear Information System (INIS)

    Furet, J.; Guyot, Ch.

    1967-01-01

    A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [fr

  1. Interactive reliability assessment using an integrated reliability data bank

    International Nuclear Information System (INIS)

    Allan, R.N.; Whitehead, A.M.

    1986-01-01

    The logical structure, techniques and practical application of a computer-aided technique based on a microcomputer using floppy disc Random Access Files is described. This interactive computational technique is efficient if the reliability prediction program is coupled directly to a relevant source of data to create an integrated reliability assessment/reliability data bank system. (DG)

  2. A human reliability analysis of the University of New Mexico's AGN- 201M nuclear research reactor

    International Nuclear Information System (INIS)

    Brumburgh, G.P.; Heger, A.S.

    1992-01-01

    During 1990--1991, a probabilistic risk assessment was conducted on the University of New Mexico's AGN-201M nuclear research reactor to address the risk and consequence of a maximum hypothetical release accident. The assessment indicated a potential for consequential human error to precipitate Chis scenario. Subsequently, a human reliability analysis was performed to evaluate the significance of human interaction on the reactor's safety systems. This paper presents the results of that investigation

  3. The evaluation of operator reliability factors on power reactor

    International Nuclear Information System (INIS)

    Karlina, Itjeu; Supriatna, Piping; W, Suharyo; Santosa, Kussigit; Darlis; S, Bambang; Y, Sasongko

    1999-01-01

    The sophisticated technology system was not assured the reliability system itself because it has contained a part of human dependence affected successfully of reactor operation either how work smoothly and safe or failure ac cured and then accident appears promptly. The evaluation of operator reliability factor on ABWR power reactor has been carried out which consist of criterion skill and workload according to NUREG/CR-2254, NUREG/CR-4016 and NUREG-0835 the reactor operation reliability emphasize to the operator are synergic between skill and workload themselves. The employee's skill will affect to the type and level of their tasks. The operator's skill depend on education and experiences, position or responsibility of tasks, physical conditions (age uninvalid of physic/mental

  4. Approach to developing reliable space reactor power systems

    International Nuclear Information System (INIS)

    Mondt, J.F.; Shinbrot, C.H.

    1991-01-01

    The Space Reactor Power System Project is in the engineering development phase of a three-phase program. During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described in this paper along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top down systems approach which includes a point design based on a detailed technical specification of a 100 kW power system

  5. Test reactor risk assessment methodology

    International Nuclear Information System (INIS)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

  6. Human reliability assessment in context

    International Nuclear Information System (INIS)

    Hollnagel, Erik

    2005-01-01

    Human Reliability Assessment (HRA) is conducted on the unspoken premise that 'human error' is a meaningful concept and that it can be associated with individual actions. The basis for this assumption it found in the origin of HRA, as a necessary extension of PSA to account for the impact of failures emanating from human actions. Although it was natural to model HRA on PSA, a large number of studies have shown that the premises are wrong, specifically that human and technological functions cannot be decomposed in the same manner. The general experience from accident studies also indicates that action failures are a function of the context, and that it is the variability of the context rather than the 'human error probability' that is the much sought for signal. Accepting this will have significant consequences for the way in which HRA, and ultimately also PSA, should be pursued

  7. MAPLE-X10 reactor safety assessment

    International Nuclear Information System (INIS)

    Cotnam, K.D.; Lounsbury, R.I.; Gillespie, G.E.

    1990-01-01

    This paper reports on the safety assessment of the 10 MW MAPLE-X10 reactor which has involved a substantial component of PSA analysis to supplement deterministic analysis. Initiating events are identified through the use of a master logic diagram. The events are then examined through event sequence diagrams, at the concept design stage, followed by a set of reliability analyses that are coordinated with the event sequence diagrams. Improvements identified through the reliability analyses are incorporated into the design to ensure that safety objectives are attained

  8. An approach for assessing ALWR passive safety system reliability

    International Nuclear Information System (INIS)

    Hake, T.M.

    1991-01-01

    Many of the advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive rather than active systems to perform safety functions. Despite the reduced redundancy of the passive systems as compared to active systems in current plants, the assertion is that the overall safety of the plant is enhanced due to the much higher expected reliability of the passive systems. In order to investigate this assertion, a study is being conducted at Sandia National Laboratories to evaluate the reliability of ALWR passive safety features in the context of probabilistic risk assessment (PRA). The purpose of this paper is to provide a brief overview of the approach to this study. The quantification of passive system reliability is not as straightforward as for active systems, due to the lack of operating experience, and to the greater uncertainty in the governing physical phenomena. Thus, the adequacy of current methods for evaluating system reliability must be assessed, and alternatives proposed if necessary. For this study, the Westinghouse Advanced Passive 600 MWe reactor (AP600) was chosen as the advanced reactor for analysis, because of the availability of AP600 design information. This study compares the reliability of AP600 emergency cooling system with that of corresponding systems in a current generation reactor

  9. Reliability assurance programme guidebook for advanced light water reactors

    International Nuclear Information System (INIS)

    2001-12-01

    To facilitate the implementation of reliability assurance programmes (RAP) within future advanced reactor programmes and to ensure that the next generation of commercial nuclear reactors achieves the very high levels of safety, reliability and economy which are expected of them, in 1996, the International Atomic Energy Agency (IAEA) established a task to develop a guidebook for reliability assurance programmes. The draft RAP guidebook was prepared by an expert consultant and was reviewed/modified at an Advisory Group meeting (7-10 April 1997) and at a consults meeting (7-10 October 1997). The programme for the RAP guidebook was reported to and guided by the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR). This guidebook will demonstrate how the designers and operators of future commercial nuclear plants can exploit the risk, reliability and availability engineering methods and techniques developed over the past two decades to augment existing design and operational nuclear plant decision-making capabilities. This guidebook is intended to provide the necessary understanding, insights and examples of RAP management systems and processes from which a future user can derive his own plant specific reliability assurance programmes. The RAP guidebook is intended to augment, not replace, specific reliability assurance requirements defined by the utility requirements documents and by individual nuclear steam supply system (NSSS) designers. This guidebook draws from utility experience gained during implementation of reliability and availability improvement and risk based management programmes to provide both written and diagrammatic 'how to' guidance which can be followed to assure conformance with the specific requirements outlined by utility requirements documents and in the development of a practical and effective plant specific RAP in any IAEA Member State

  10. Advanced reactor passive system reliability demonstration analysis for an external event

    International Nuclear Information System (INIS)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

    2017-01-01

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event

  11. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor

    2017-03-01

    Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

  12. Advanced reactor passive system reliability demonstration analysis for an external event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

    2017-03-15

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

  13. The Centralized Reliability Data Organization (CREDO); an advanced nuclear reactor reliability, availability, and maintainability data bank and data analysis center

    International Nuclear Information System (INIS)

    Knee, H.E.

    1991-01-01

    The Centralized Reliability Data Organization (CREDO) is a data bank and data analysis center, which since 1985 has been jointly sponsored by the US Department of Energy's (US DOE's) Office of Technology Support Programs and Japan's Power Reactor and Nuclear Fuel Development Corporation (PNC). It focuses on reliability, availability and maintainability (RAM) data for components (e.g. valves, pumps, etc.) operating in advanced nuclear reactor facilities. As originally intended, the purpose of the CREDO system was to provide a centralized source of accurate, up-to-date data and information for use in RAM analyses necessary for meeting DOE's data needs in the areas of advanced reactor safety assessments, design and licensing. In particular, creation of the CREDO system was considered an essential element needed to fulfill the DOE Breeder Reactor Safety Program's commitment of 'identifying and exploiting areas in which probabilistic methods can be developed and used in making reactor safety Research and Development choices and optimizing designs of safety systems'. CREDO and its operation are explained. (author)

  14. Design reliability assurance program for Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Beom-Su; Han, Jin-Kyu; Na, Jang Hwan; Yoo, Kyung Yeong

    1997-01-01

    The Korean Next Generation Reactor (KNGR) project is to develop standardized nuclear power plant design for the construction of future nuclear power plants in Korea. The main purpose of the KNGR project is to develop the advanced nuclear power plants, which enhance safety and economics significantly through the incorporation of design concepts for severe accident prevention and mitigation, supplementary passive safety concept, simplification and application of modularization and so on. For those, Probabilistic Safety Assessment (PSA) and availability study will be performed at the early stage of the design, and the Design Reliability Assurance Program (D-RAP) is applied in the development of the KNGR to ensure that the safety and availability evaluated in the PSA and availability study at the early phase of the design is maintained through the detailed design, construction, procurement and operation of the plants. This paper presents the D-RAP concept that could be applied at the stage of the basic design of the nuclear power plants, based on the models for the reference plants and/or similar plants. 4 refs., 1 fig

  15. FFTF [Fast Flux Test Facility] reactor shutdown system reliability reevaluation

    International Nuclear Information System (INIS)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations

  16. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose Luis

    1996-01-01

    Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

  17. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose L.

    2000-01-01

    Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

  18. Reliability of redundant structures of nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Vojnovic, B.

    1983-01-01

    In this paper, reliability of various redundant structures of PWR protection systems has been analysed. Structures of reactor tip systems as well as the systems for activation of safety devices have been presented. In all those systems redundancy is achieved by means of so called majority voting logic ('r out of n' structures). Different redundant devices have been compared, concerning probability of occurrence of safe as well as unsafe failures. (author)

  19. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

    There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

  20. Structural reliability assessment capability in NESSUS

    Science.gov (United States)

    Millwater, H.; Wu, Y.-T.

    1992-07-01

    The principal capabilities of NESSUS (Numerical Evaluation of Stochastic Structures Under Stress), an advanced computer code developed for probabilistic structural response analysis, are reviewed, and its structural reliability assessed. The code combines flexible structural modeling tools with advanced probabilistic algorithms in order to compute probabilistic structural response and resistance, component reliability and risk, and system reliability and risk. An illustrative numerical example is presented.

  1. Reliability assessments in qualitative health promotion research.

    Science.gov (United States)

    Cook, Kay E

    2012-03-01

    This article contributes to the debate about the use of reliability assessments in qualitative research in general, and health promotion research in particular. In this article, I examine the use of reliability assessments in qualitative health promotion research in response to health promotion researchers' commonly held misconception that reliability assessments improve the rigor of qualitative research. All qualitative articles published in the journal Health Promotion International from 2003 to 2009 employing reliability assessments were examined. In total, 31.3% (20/64) articles employed some form of reliability assessment. The use of reliability assessments increased over the study period, ranging from qualitative articles decreased. The articles were then classified into four types of reliability assessments, including the verification of thematic codes, the use of inter-rater reliability statistics, congruence in team coding and congruence in coding across sites. The merits of each type were discussed, with the subsequent discussion focusing on the deductive nature of reliable thematic coding, the limited depth of immediately verifiable data and the usefulness of such studies to health promotion and the advancement of the qualitative paradigm.

  2. Generic component reliability data for research reactor PSA

    International Nuclear Information System (INIS)

    1997-02-01

    The purpose of this document is to provide reference generic component-reliability information for a variety of research reactor types. As noted in Section 2 and Table IV, component data accumulated over many years is in the database. It is expected that the report should provide representative data which will remain valid for a number of years. The database provides component failure rates on a time and/or demand related basis according to the operational modes of the components. No update of the database is presently planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available by the contributing research reactor facilities themselves. As noted in Section 1.1, the report does not include a detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussions regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, 7 tabs

  3. Domestic manufacturing and reliability improvement of reactor water recirculation equipment

    International Nuclear Information System (INIS)

    Kobayashi, Hidekazu; Oi, Masao; Shida, Toichi; Yokomori, Takashi

    1982-01-01

    The reactor coolant recirculation system is one of the important systems to control the reactor output in BWR nuclear power plants. Its components require high reliability and maintainability as well as controllability. For many Japanese nuclear power plants, recirculation pumps, fluid couplings and others have been imported so far. Hitachi Ltd. has established a domestic manufacturing organization through the development and test of these equipment. The fundamental design conditions for these equipment are the improvement of the rate of utilization of plant facility, the capability to follow load, and output power stability. In this paper, the specifications, the investigation of moment of inertia and the design features of recirculation pumps, driving motors and variable frequency power supply systems are described. The paper also reports on the combination test implemented to evaluate the recirculation system. The combination test includes the test using water rheostat for the power source facility and the loading test for a recirculation pump. The application of those system equipment to an actual plant was analyzed and evaluated on a basis of the test data obtained. The result showed that the equipment can achieve the rate of change of reactor power of 30%/min. Those equipment have been employed for No. 2 reactor plant of the Fukushima No. 2 Nuclear Power Station, the Tokyo Electric Power Co., Inc. (Wakatsuki, Y.)

  4. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Mirvakili, S.M.

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  5. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  6. Quality and Reliability Aspects in Nuclear Power Reactor Fuel Engineering

    International Nuclear Information System (INIS)

    2015-01-01

    In order to decrease costs and increase competitiveness, nuclear utilities use more challenging operational conditions, longer fuel cycles and higher burnups, which require modifications in fuel designs and materials. Different aspects of quality assurance and control, as well as analysis of fuel performance have been considered in a number of specialized publications. The present publication provides a concise but comprehensive overview of all interconnected quality and reliability issues in fuel fabrication, design and operation. It jointly tackles technical, safety and organizational aspects, and contains examples of state of the art developments and good practices of coordinated work of fuel designers, vendors and reactor operators

  7. Proving Test on the Reliability for Reactor Containment Vessel

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.

    1988-01-01

    NUPEC (Nuclear Power Engineering Test Center) has started an eight-year project of Proving Test on the Reliability for Reactor Containment Vessel since June 1987. The objective of this project is to confirm the integrity of containment vessels under severe accident conditions. This paper shows the outline of this project. The test Items are (1) Hydrogen mixing and distribution test, (2) Hydrogen burning test, (3) Iodine trapping characteristics test, and (4) Structural behavior test. Based on the test results, computer codes are verified and as the results of analysis and evaluation by the computer codes, containment integrity is to be confirmed

  8. Operator reliability assessment system (OPERAS)

    International Nuclear Information System (INIS)

    Spurgin, A.J.; Hallam, J.W.; Spurgin, J.P.; Singh, A.

    1991-01-01

    The paper gives an overview of the OPERAS project. It discusses the background which led to the design of the PC-based data collection and analysis system connected to plant training simulators including those used for nuclear power plants. The usefulness of a system like OPERAS was perceived during an earlier EPRI project, the Operator Reliability Experiments project, by EPRI and PG and E. The data collection and analysis approaches used in OPERAS were developed during the ORE project. The paper not only discusses the design of OPERAS but discusses the functions performed and the current experiences with the two prototype systems. Also listed are potential uses of OPERAS by utility personnel in Operations, Training and PRA groups

  9. Containment concepts assessment for the SEAFP reactor

    International Nuclear Information System (INIS)

    Di Pace, L.; Natalizio, A.

    2000-01-01

    A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

  10. Results of a Demonstration Assessment of Passive System Reliability Utilizing the Reliability Method for Passive Systems (RMPS)

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia; Grelle, Austin

    2015-04-26

    Advanced small modular reactor designs include many advantageous design features such as passively driven safety systems that are arguably more reliable and cost effective relative to conventional active systems. Despite their attractiveness, a reliability assessment of passive systems can be difficult using conventional reliability methods due to the nature of passive systems. Simple deviations in boundary conditions can induce functional failures in a passive system, and intermediate or unexpected operating modes can also occur. As part of an ongoing project, Argonne National Laboratory is investigating various methodologies to address passive system reliability. The Reliability Method for Passive Systems (RMPS), a systematic approach for examining reliability, is one technique chosen for this analysis. This methodology is combined with the Risk-Informed Safety Margin Characterization (RISMC) approach to assess the reliability of a passive system and the impact of its associated uncertainties. For this demonstration problem, an integrated plant model of an advanced small modular pool-type sodium fast reactor with a passive reactor cavity cooling system is subjected to a station blackout using RELAP5-3D. This paper discusses important aspects of the reliability assessment, including deployment of the methodology, the uncertainty identification and quantification process, and identification of key risk metrics.

  11. A reliability assessment methodology for the VHTR passive safety system

    International Nuclear Information System (INIS)

    Lee, Hyungsuk; Jae, Moosung

    2014-01-01

    The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

  12. Reevaluating Interrater Reliability in Offender Risk Assessment

    NARCIS (Netherlands)

    van der Knaap, L.M.; Leenarts, L.E.W.; Born, M.P.; Oosterveld, P.

    2012-01-01

    Offender risk and needs assessment, one of the pillars of the risk-need-responsivity model of offender rehabilitation, usually depends on raters assessing offender risk and needs. The few available studies of interrater reliability in offender risk assessment are, however, limited in the

  13. Reevaluating Interrater Reliability in Offender Risk Assessment

    Science.gov (United States)

    van der Knaap, Leontien M.; Leenarts, Laura E. W.; Born, Marise Ph.; Oosterveld, Paul

    2012-01-01

    Offender risk and needs assessment, one of the pillars of the risk-need-responsivity model of offender rehabilitation, usually depends on raters assessing offender risk and needs. The few available studies of interrater reliability in offender risk assessment are, however, limited in the generalizability of their results. The present study…

  14. Reliability evaluation of the Savannah River reactor leak detection system

    International Nuclear Information System (INIS)

    Daugherty, W.L.; Sindelar, R.L.; Wallace, I.T.

    1991-01-01

    The Savannah River Reactors have been in operation since the mid-1950's. The primary degradation mode for the primary coolant loop piping is intergranular stress corrosion cracking. The leak-before-break (LBB) capability of the primary system piping has been demonstrated as part of an overall structural integrity evaluation. One element of the LBB analyses is a reliability evaluation of the leak detection system. The most sensitive element of the leak detection system is the airborne tritium monitors. The presence of small amounts of tritium in the heavy water coolant provide the basis for a very sensitive system of leak detection. The reliability of the tritium monitors to properly identify a crack leaking at a rate of either 50 or 300 lb/day (0.004 or 0.023 gpm, respectively) has been characterized. These leak rates correspond to action points for which specific operator actions are required. High reliability has been demonstrated using standard fault tree techniques. The probability of not detecting a leak within an assumed mission time of 24 hours is estimated to be approximately 5 x 10 -5 per demand. This result is obtained for both leak rates considered. The methodology and assumptions used to obtain this result are described in this paper. 3 refs., 1 fig., 1 tab

  15. Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

    1991-10-01

    This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs

  16. Emergency diesel generator reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    A program to apply some of the techniques of reliability engineering to the High Flux Isotope Reactor (HFIR) was started on August 8, 1992. Part of the program was to track the conditional probabilities of the emergency diesel generators responding to a valid demand. This was done to determine if the performance of the emergency diesel generators (which are more than 25 years old) has deteriorated. The conditional probabilities of the diesel generators were computed and trended for the period from May 1990 to December 1992. The calculations indicate that the performance of the emergency diesel generators has not deteriorated in recent years, i.e., the conditional probabilities of the emergency diesel generators have been fairly stable over the last few years. This information will be one factor than may be considered in the decision to replace the emergency diesel generators

  17. Reliability aspects of radiation damage in reactor pressure vessel mterials

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1985-01-01

    The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

  18. An approach for assessing human decision reliability

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-01-01

    This paper presents a method to study human reliability in decision situations related to nuclear power plant disturbances. Decisions often play a significant role in handling of emergency situations. The method may be applied to probabilistic safety assessments (PSAs) in cases where decision making is an important dimension of an accident sequence. Such situations are frequent e.g. in accident management. In this paper, a modelling approach for decision reliability studies is first proposed. Then, a case study with two decision situations with relatively different characteristics is presented. Qualitative and quantitative findings of the study are discussed. In very simple decision cases with time pressure, time reliability correlation proved out to be a feasible reliability modelling method. In all other decision situations, more advanced probabilistic decision models have to be used. Finally, decision probability assessment by using simulator run results and expert judgement is presented

  19. Assessing reliability in energy supply systems

    International Nuclear Information System (INIS)

    McCarthy, Ryan W.; Ogden, Joan M.; Sperling, Daniel

    2007-01-01

    Reliability has always been a concern in the energy sector, but concerns are escalating as energy demand increases and the political stability of many energy supply regions becomes more questionable. But how does one define and measure reliability? We introduce a method to assess reliability in energy supply systems in terms of adequacy and security. It derives from reliability assessment frameworks developed for the electricity sector, which are extended to include qualitative considerations and to be applicable to new energy systems by incorporating decision-making processes based on expert opinion and multi-attribute utility theory. The method presented here is flexible and can be applied to any energy system. To illustrate its use, we apply the method to two hydrogen pathways: (1) centralized steam reforming of imported liquefied natural gas with pipeline distribution of hydrogen, and (2) on-site electrolysis of water using renewable electricity produced independently from the electricity grid

  20. Reliability assessment based on subjective inferences

    International Nuclear Information System (INIS)

    Ma Zhibo; Zhu Jianshi; Xu Naixin

    2003-01-01

    The reliability information which comes from subjective analysis is often incomplete prior. This information can be generally assumed to exist in the form of either a stated prior mean of R (reliability) or a stated prior credibility interval on R. An efficient approach is developed to determine a complete beta prior distribution from the subjective information according to the principle of maximum entropy, and the the reliability of survival/failure product is assessed via Bayes theorem. Numerical examples are presented to illustrate the methods

  1. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  2. Mechanical reliability assessment of optical fibres in Radiation environments

    International Nuclear Information System (INIS)

    Van Uffelen, M.

    2006-01-01

    After more than two decades of intensive research and even some pioneering applications in space, optical fibres are now finding their way in various radiation environments, including both fission and future fusion nuclear-power plants, and high-energy physics experiments. For example, next to distributed monitoring applications of large nuclear infrastructures, fibre-optics can also be used for data communications during maintenance operations in the reactor vessel of the future ITER (International Thermonuclear Experimental Reactor), or for plasma diagnostics applications during operation of the reactor. These maintenance and diagnostics tasks require the optical fibres to withstand extremely high doses of radiation, up to MGy dose levels and temperatures above 150 degrees Celsius. The reliability assessment of fibre-optic systems for their qualification in nuclear environments often requires to meet stringent radiation tolerance levels. The majority of (usually accelerated) radiation assessments have so far focused on optical properties, such as wavelength-dependent radiation induced attenuation and radio-luminescence. The relation of these radiation effects with the fabrication methods and other environmental parameters has been the subject of years of research. Only a few results are available on the long-term evolution of mechanical properties of irradiated optical fibres. As a first step towards understanding the long-term reliability of fibre-optic composite cables in hostile radiation environments, we therefore performed dynamic fatigue tests with different commercial-grade optical fibres, both multi-mode and single-mode types

  3. Human reliability. Is probabilistic human reliability assessment possible?

    International Nuclear Information System (INIS)

    Mosneron Dupin, F.

    1996-01-01

    The possibility of carrying out Probabilistic Human Reliability Assessments (PHRA) is often doubted. Basing ourselves on the experience Electricite de France (EDF) has acquired in Probabilistic Safety Assessments for nuclear power plants, we show why the uncertainty of PHRA is very high. We then specify the limits of generic data and models for PHRA: very important factors are often poorly taken into account. To account for them, you need to have proper understanding of the actual context in which operators work. This demands surveys on the field (power plant and simulator) all of which must be carried out with behaviours science skills. The idea of estimating the probabilities of operator failure must not be abandoned, but probabilities must be given less importance, for they are only approximate indications. The qualitative aspects of PHRA should be given greater value (analysis process and qualitative insights). That is why the description (illustrated by case histories) of the main mechanisms of human behaviour, and of their manifestations in the nuclear power plant context (in terms of habits, attitudes, and informal methods and organization in particular) should be an important part of PHRA handbooks. These handbooks should also insist more on methods for gathering information on the actual context of the work of operators. Under these conditions, the PHRA should be possible and even desirable as a process for systematic analysis and assessment of human intervention. (author). 24 refs, 2 figs, 1 tab

  4. Human reliability. Is probabilistic human reliability assessment possible?

    Energy Technology Data Exchange (ETDEWEB)

    Mosneron Dupin, F

    1997-12-31

    The possibility of carrying out Probabilistic Human Reliability Assessments (PHRA) is often doubted. Basing ourselves on the experience Electricite de France (EDF) has acquired in Probabilistic Safety Assessments for nuclear power plants, we show why the uncertainty of PHRA is very high. We then specify the limits of generic data and models for PHRA: very important factors are often poorly taken into account. To account for them, you need to have proper understanding of the actual context in which operators work. This demands surveys on the field (power plant and simulator) all of which must be carried out with behaviours science skills. The idea of estimating the probabilities of operator failure must not be abandoned, but probabilities must be given less importance, for they are only approximate indications. The qualitative aspects of PHRA should be given greater value (analysis process and qualitative insights). That is why the description (illustrated by case histories) of the main mechanisms of human behaviour, and of their manifestations in the nuclear power plant context (in terms of habits, attitudes, and informal methods and organization in particular) should be an important part of PHRA handbooks. These handbooks should also insist more on methods for gathering information on the actual context of the work of operators. Under these conditions, the PHRA should be possible and even desirable as a process for systematic analysis and assessment of human intervention. (author). 24 refs, 2 figs, 1 tab.

  5. The DYLAM approach to systems safety and reliability assessment

    International Nuclear Information System (INIS)

    Amendola, A.

    1988-01-01

    A survey of the principal features and applications of DYLAM (Dynamic Logical Analytical Methodology) is presented, whose basic principles can be summarized as follows: after a particular modelling of the component states, computerized heuristical procedures generate stochastic configurations of the system, whereas the resulting physical processes are simultaneously simulated to give account of the possible interactions between physics and states and, on the other hand, to search for system dangerous configurations and related probabilities. The association of probabilistic techniques for describing the states with physical equations for describing the process results in a very powerful tool for safety and reliability assessment of systems potentially subjected to dangerous incidental transients. A comprehensive picture of DYLAM capability for manifold applications can be obtained by the review of the study cases analyzed (LMFBR core accident, systems reliability assessment, accident simulation, man-machine interaction analysis, chemical reactors safety, etc.)

  6. An approach for assessing ALWR passive safety system reliability

    International Nuclear Information System (INIS)

    Hake, T.M.

    1991-01-01

    Many advanced light water reactor designs incorporate passive rather than active safety features for front-line accident response. A method for evaluating the reliability of these passive systems in the context of probabilistic risk assessment has been developed at Sandia National Laboratories. This method 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. These processes provide the system's driving force; examples are natural circulation and gravity-induced injection. This paper describes the method, and provides some preliminary results of application of the approach to the Westinghouse AP600 design

  7. Analysis of dependent failures in risk assessment and reliability evaluation

    International Nuclear Information System (INIS)

    Fleming, K.N.; Mosleh, A.; Kelley, A.P. Jr.; Gas-Cooled Reactors Associates, La Jolla, CA)

    1983-01-01

    The ability to estimate the risk of potential reactor accidents is largely determined by the ability to analyze statistically dependent multiple failures. The importance of dependent failures has been indicated in recent probabilistic risk assessment (PRA) studies as well as in reports of reactor operating experiences. This article highlights the importance of several different types of dependent failures from the perspective of the risk and reliability analyst and provides references to the methods and data available for their analysis. In addition to describing the current state of the art, some recent advances, pitfalls, misconceptions, and limitations of some approaches to dependent failure analysis are addressed. A summary is included of the discourse on this subject, which is presented in the Institute of Electrical and Electronics Engineers/American Nuclear Society PRA Procedures Guide

  8. Automated reliability assessment for spectroscopic redshift measurements

    Science.gov (United States)

    Jamal, S.; Le Brun, V.; Le Fèvre, O.; Vibert, D.; Schmitt, A.; Surace, C.; Copin, Y.; Garilli, B.; Moresco, M.; Pozzetti, L.

    2018-03-01

    Context. Future large-scale surveys, such as the ESA Euclid mission, will produce a large set of galaxy redshifts (≥106) that will require fully automated data-processing pipelines to analyze the data, extract crucial information and ensure that all requirements are met. A fundamental element in these pipelines is to associate to each galaxy redshift measurement a quality, or reliability, estimate. Aim. In this work, we introduce a new approach to automate the spectroscopic redshift reliability assessment based on machine learning (ML) and characteristics of the redshift probability density function. Methods: We propose to rephrase the spectroscopic redshift estimation into a Bayesian framework, in order to incorporate all sources of information and uncertainties related to the redshift estimation process and produce a redshift posterior probability density function (PDF). To automate the assessment of a reliability flag, we exploit key features in the redshift posterior PDF and machine learning algorithms. Results: As a working example, public data from the VIMOS VLT Deep Survey is exploited to present and test this new methodology. We first tried to reproduce the existing reliability flags using supervised classification in order to describe different types of redshift PDFs, but due to the subjective definition of these flags (classification accuracy 58%), we soon opted for a new homogeneous partitioning of the data into distinct clusters via unsupervised classification. After assessing the accuracy of the new clusters via resubstitution and test predictions (classification accuracy 98%), we projected unlabeled data from preliminary mock simulations for the Euclid space mission into this mapping to predict their redshift reliability labels. Conclusions: Through the development of a methodology in which a system can build its own experience to assess the quality of a parameter, we are able to set a preliminary basis of an automated reliability assessment for

  9. Assessment of the Reliability of Concrete Bridges

    DEFF Research Database (Denmark)

    Middleton, C. R.; Thoft-Christensen, Palle

    Although there has been a considerable amount of research into different aspects of concrete bridge reliability, it has still not been widely adopted in professional practice other than in the development and calibration of codes. This situation appears to be changing as there has been a signific......Although there has been a considerable amount of research into different aspects of concrete bridge reliability, it has still not been widely adopted in professional practice other than in the development and calibration of codes. This situation appears to be changing as there has been...... adopted to assist in achieving this goal. Rather than review the specific research on this subject this paper examines a number of key issues related to the practical application of reliability analysis to the assessment of concrete bridges....

  10. Reliability analysis and assessment of structural systems

    International Nuclear Information System (INIS)

    Yao, J.T.P.; Anderson, C.A.

    1977-01-01

    The study of structural reliability deals with the probability of having satisfactory performance of the structure under consideration within any specific time period. To pursue this study, it is necessary to apply available knowledge and methodology in structural analysis (including dynamics) and design, behavior of materials and structures, experimental mechanics, and the theory of probability and statistics. In addition, various severe loading phenomena such as strong motion earthquakes and wind storms are important considerations. For three decades now, much work has been done on reliability analysis of structures, and during this past decade, certain so-called 'Level I' reliability-based design codes have been proposed and are in various stages of implementation. These contributions will be critically reviewed and summarized in this paper. Because of the undesirable consequences resulting from the failure of nuclear structures, it is important and desirable to consider the structural reliability in the analysis and design of these structures. Moreover, after these nuclear structures are constructed, it is desirable for engineers to be able to assess the structural reliability periodically as well as immediately following the occurrence of severe loading conditions such as a strong-motion earthquake. During this past decade, increasing use has been made of techniques of system identification in structural engineering. On the basis of non-destructive test results, various methods have been developed to obtain an adequate mathematical model (such as the equations of motion with more realistic parameters) to represent the structural system

  11. Event and fault tree model for reliability analysis of the greek research reactor

    International Nuclear Information System (INIS)

    Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes

    2013-01-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

  12. Event and fault tree model for reliability analysis of the greek research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: atalbuquerque@ien.gov.br, E-mail: btony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

  13. Reliability

    OpenAIRE

    Condon, David; Revelle, William

    2017-01-01

    Separating the signal in a test from the irrelevant noise is a challenge for all measurement. Low test reliability limits test validity, attenuates important relationships, and can lead to regression artifacts. Multiple approaches to the assessment and improvement of reliability are discussed. The advantages and disadvantages of several different approaches to reliability are considered. Practical advice on how to assess reliability using open source software is provided.

  14. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Dixon, B.W.; Bray, M.A.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

  15. Balancing human and technical reliability in the design of advanced nuclear reactors

    International Nuclear Information System (INIS)

    Papin, Bernard

    2011-01-01

    Highlights: ► Human factors exigencies are often overseen during the early design phases of NPP. ► Optimization of reactors safety is only based on technical reliability considerations. ► The search for more technical reliability often leads to more system complexity. ► System complexity is a major contributor to the operator's poor performance. ► Our method enables to assess plant complexity and it's impact on human performance. - Abstract: The strong influence of human factors (HF) on the safety of nuclear facilities is nowadays recognised and the designers are now enforced to consider HF requirements in the design of new facilities. Yet, this consideration of human factors requirements is still more or less restricted to the latest phases of the projects, essentially for the design of human-system interfaces (HSI's) and control rooms, although the design options influencing at most the human performance in operation are indeed fixed during the very early phases of the new reactors projects. The main reason of this late consideration of HF is that there exist few methods and models for anticipating the influence of fundamental design options on the future performance of operation teams. This paper describes a set of new tools permitting (i) determination of the impact of the fundamental process design options on the future activity of the operation teams and (ii) assessment of the influence of these operational constraints on teams performance. These tools are intended to guide the design of future 4th generation (GEN4) reactors, within the frame of a global risk-informed design approach, considering technical and human reliability exigencies in a balanced way.

  16. Methodology for reliability based condition assessment

    International Nuclear Information System (INIS)

    Mori, Y.; Ellingwood, B.

    1993-08-01

    Structures in nuclear power plants may be exposed to aggressive environmental effects that cause their strength to decrease over an extended period of service. A major concern in evaluating the continued service for such structures is to ensure that in their current condition they are able to withstand future extreme load events during the intended service life with a level of reliability sufficient for public safety. This report describes a methodology to facilitate quantitative assessments of current and future structural reliability and performance of structures in nuclear power plants. This methodology takes into account the nature of past and future loads, and randomness in strength and in degradation resulting from environmental factors. An adaptive Monte Carlo simulation procedure is used to evaluate time-dependent system reliability. The time-dependent reliability is sensitive to the time-varying load characteristics and to the choice of initial strength and strength degradation models but not to correlation in component strengths within a system. Inspection/maintenance strategies are identified that minimize the expected future costs of keeping the failure probability of a structure at or below an established target failure probability during its anticipated service period

  17. Effect of dc-power-system reliability on reactor-shutdown cooling

    International Nuclear Information System (INIS)

    Kolaczkowski, A.M.; Baranowsky, P.W.; Hickman, J.W.

    1981-01-01

    The DC power systems in a nuclear power plant provide control and motive power to valves, instrumentation, emergency diesel generators, and many other components and systems during all phases of plant operation including abnormal shutdowns and accident situations. A specific area of concern is the adequacy of the minimum design requirements for DC power systems, particularly with regard to multiple and common cause failures. This concern relates to the application of the single failure criterion for assuring a reliable DC power supply which may be required for the functionability of shutdown cooling systems. The results are presented of a reliability based study performed to assess the adequacy of DC power supply design requirements for currently operating light water reactors with particular attention to shutdown cooling requirements

  18. Assessment of torsatrons as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Painter, S.L.

    1992-12-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

  19. Probability safety assessment of LOOP accident to molten salt reactor

    International Nuclear Information System (INIS)

    Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

    2013-01-01

    Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

  20. An examination of reliability critical items in liquid metal reactors: An analysis by the Centralized Reliability Data Organization (CREDO)

    International Nuclear Information System (INIS)

    Humphrys, B.L.; Haire, M.J.; Koger, K.H.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.; Okubo, Y.

    1987-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 21,000 components and approximately 1300 event records. A conservative estimation is that the total component operating hours is approaching 3.5 billion hours. Because data gathering for CREDO concentrates on event (failure) information, the work reported here focuses on the reliability information contained in CREDO and the development of reliability critical items lists. That is, components are ranked in prioritized lists from worst to best performers from a reliability standpoint. For the data contained in the CREDO data base, FFTF and JOYO show reliability growth; EBR-II reveals a slight unreliability growth for those components tracked by CREDO. However, tabulations of events which cause reactor shutdowns decrease with time at each site

  1. Engineering reliability in design phase: An application to AP-600 reactor passive safety system

    International Nuclear Information System (INIS)

    Majumdr, D.; Siahpush, A.S.; Hills, S.W.

    1992-01-01

    A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

  2. Reliability assessment of passive containment isolation system using APSRA methodology

    International Nuclear Information System (INIS)

    Nayak, A.K.; Jain, Vikas; Gartia, M.R.; Srivastava, A.; Prasad, Hari; Anthony, A.; Gaikwad, A.J.; Bhatia, S.; Sinha, R.K.

    2008-01-01

    In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) has been employed for evaluation of the reliability of passive systems. The methodology has been applied to the passive containment isolation system (PCIS) of the Indian advanced heavy water reactor (AHWR). In the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criterion. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the PCIS performance. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components is evaluated through a classical PSA treatment using the generic data. The reliability of the PCIS is evaluated from the probability of availability of the components for the success of the passive containment isolation system

  3. Hanford B Reactor Building Hazard Assessment Report

    International Nuclear Information System (INIS)

    Griffin, P. W.

    1999-01-01

    The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

  4. Human reliability in probabilistic safety assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

    Nowadays a growing interest in medioambiental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processess and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects. (This relevance has been demostrated in the accidents happenned). However in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a guide to carry out a Human Reliability Analysis and c) a selected overwiev of the techniques and methodologies currently applied in this area. (Author)

  5. Human Reliability in Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

    Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs

  6. Reliability analysis of reactor systems by applying probability method; Analiza pouzdanosti reaktorskih sistema primenom metoda verovatnoce

    Energy Technology Data Exchange (ETDEWEB)

    Milivojevic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1974-12-15

    Probability method was chosen for analysing the reactor system reliability is considered realistic since it is based on verified experimental data. In fact this is a statistical method. The probability method developed takes into account the probability distribution of permitted levels of relevant parameters and their particular influence on the reliability of the system as a whole. The proposed method is rather general, and was used for problem of thermal safety analysis of reactor system. This analysis enables to analyze basic properties of the system under different operation conditions, expressed in form of probability they show the reliability of the system on the whole as well as reliability of each component.

  7. Student Precision and Reliability of the Team Sport Assessment in ...

    African Journals Online (AJOL)

    TSAP) and formative assessment of invasion sport. The specific objectives were to determine the degree of agreement among expert observers, inter-observer reliability (internal consistency), and intra observer reliability (temporal reliability).

  8. Advanced Test Reactor probabilistic risk assessment

    International Nuclear Information System (INIS)

    Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

    1993-01-01

    This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

  9. Contributions to the determination of the thermal core reliability of pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.; Melchior, H.; Prasser, H.M.

    1982-09-01

    The investigations in the field of thermohydraulics of PW reactors are aimed at a possible increase of economy and reliability of WWER-type-reactors. In detail the flow distribution at the core entrance, the modification of the power distribution as a result of an irregular temperature distribution at the core entrance, and based on the theory of hot spots the thermic core reliability are studied. In this connection qualitatively new methods are applied characterized by low expenditure. (author)

  10. Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Andersen, E.S.; Bowey, R.E.; Diaz, A.A.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1991-01-01

    This program is intended to establish the effectiveness, reliability and adequacy of inservice inspection of reactor pressure vessels and primary piping systems and the impact of ISI reliability on system integrity. The objectives of the program include: (a) determine the effectiveness and reliability of ultrasonic inservice inspection (ISI) performed on commercial, light water reactor pressure vessels and piping; (b) recommend Code changes to the inspection procedures to improve the reliability of ISI; (c) using fracture mechanics analysis, determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to assure a suitably low failure probability; (d) evaluate the degree of reliability improvement which could be achieved using improved NDE techniques; and (e) based on importance of component to safety, material properties, service conditions, and NDE uncertainties, formulate improved inservice inspection criteria (including sampling plan, frequency, and reliability of inspection) for revisions to ASME Section XI and regulatory requirements needed to assure suitably low failure probabilities

  11. Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

    2011-01-01

    The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

  12. Small Modular Reactors: Institutional Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Perkowski, Ph.D.

    2012-06-01

    ? Objectives include, among others, a description of the basic development status of “small modular reactors” (SMRs) focused primarily on domestic activity; investigation of the domestic market appeal of modular reactors from the viewpoints of both key energy sector customers and also key stakeholders in the financial community; and consideration of how to proceed further with a pro-active "core group" of stakeholders substantially interested in modular nuclear deployment in order to provide the basis to expedite design/construction activity and regulatory approval. ? Information gathering was via available resources, both published and personal communications with key individual stakeholders; published information is limited to that already in public domain (no confidentiality); viewpoints from interviews are incorporated within. Discussions at both government-hosted and private-hosted SMR meetings are reflected herein. INL itself maintains a neutral view on all issues described. Note: as per prior discussion between INL and CAP, individual and highly knowledgeable senior-level stakeholders provided the bulk of insights herein, and the results of those interviews are the main source of the observations of this report. ? Attachment A is the list of individual stakeholders consulted to date, including some who provided significant earlier assessments of SMR institutional feasibility. ? Attachments B, C, and D are included to provide substantial context on the international status of SMR development; they are not intended to be comprehensive and are individualized due to the separate nature of the source materials. Attachment E is a summary of the DOE requirements for winning teams regarding the current SMR solicitation. Attachment F deserves separate consideration due to the relative maturity of the SMART SMR program underway in Korea. Attachment G provides illustrative SMR design features and is intended for background. Attachment H is included for overview

  13. Reliability studies in research reactors; Estudo de confiabilidade em reatores de pesquisa

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Tob Rodrigues de

    2013-08-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  14. Comparative availability and reliability assessment of design options for the secondary sodium loops of the EFR

    International Nuclear Information System (INIS)

    Pamme, H.

    1989-01-01

    The EFR (European Fast Reactor) project has entered a conceptual study period where different design alternatives are compared concerning feasibility, safety and economic aspects. This paper describes a comparative probabilistic availability and reliability assessment of alternative design options for the secondary sodium loops. These loops will provide heat transfer from the reactor pool to the water-steam (power generating) side. So a high operational availability of the secondary loops during plant lifetime is essential for economic power generation. Additionally a high reliability is required to fulfill the operational decay heat removal function in case of a reactor trip. Availabilities and reliabilities of the different options were assessed using failure mode and effect analysis and the fault tree method. (orig.)

  15. Software reliability growth model for safety systems of nuclear reactor

    International Nuclear Information System (INIS)

    Thirugnana Murthy, D.; Murali, N.; Sridevi, T.; Satya Murty, S.A.V.; Velusamy, K.

    2014-01-01

    The demand for complex software systems has increased more rapidly than the ability to design, implement, test, and maintain them, and the reliability of software systems has become a major concern for our, modern society.Software failures have impaired several high visibility programs in space, telecommunications, defense and health industries. Besides the costs involved, it setback the projects. The ways of quantifying it and using it for improvement and control of the software development and maintenance process. This paper discusses need for systematic approaches for measuring and assuring software reliability which is a major share of project development resources. It covers the reliability models with the concern on 'Reliability Growth'. It includes data collection on reliability, statistical estimation and prediction, metrics and attributes of product architecture, design, software development, and the operational environment. Besides its use for operational decisions like deployment, it includes guiding software architecture, development, testing and verification and validation. (author)

  16. Assessment of physical server reliability in multi cloud computing system

    Science.gov (United States)

    Kalyani, B. J. D.; Rao, Kolasani Ramchand H.

    2018-04-01

    Business organizations nowadays functioning with more than one cloud provider. By spreading cloud deployment across multiple service providers, it creates space for competitive prices that minimize the burden on enterprises spending budget. To assess the software reliability of multi cloud application layered software reliability assessment paradigm is considered with three levels of abstractions application layer, virtualization layer, and server layer. The reliability of each layer is assessed separately and is combined to get the reliability of multi-cloud computing application. In this paper, we focused on how to assess the reliability of server layer with required algorithms and explore the steps in the assessment of server reliability.

  17. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  18. Pressurizer pump reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

  19. Reliability data collection and use in risk and availability assessment

    International Nuclear Information System (INIS)

    Colombari, V.

    1989-01-01

    For EuReDatA it is a prevailing objective to initiate and support contact between experts, companies and institutions active in reliability engineering and research. Main topics of this 6th EuReDatA Conference are: Reliability data banks; incidents data banks; common cause data; source and propagation of uncertainties; computer aided risk analysis; reliability and incidents data acquisition and processing; human reliability; probabilistic safety and availability assessment; feedback of reliability into system design; data fusion; reliability modeling and techniques; structural and mechanical reliability; consequence modeling; software and electronic reliability; reliability tests. Some conference papers are separately indexed in the database. (HP)

  20. 77 FR 10786 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Reliability...

    Science.gov (United States)

    2012-02-23

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Reliability and PRA; Notice of Meeting The ACRS Subcommittee on Reliability and PRA will hold a... Modeling Application Guide.'' The Subcommittee will hear presentations by and hold discussions with the NRC...

  1. Reliability assessment of creep rupture life for Gr. 91 steel

    International Nuclear Information System (INIS)

    Kim, Woo-Gon; Park, Jae-Young; Kim, Seon-Jin; Jang, Jinsung

    2013-01-01

    Highlights: • Statistical analysis of a number of creep rupture data based on Z parameter. • Determination of the constant C in LM parameter and long-term creep life prediction. • Generation of random variables for Z s and Z cr by Monte-Carlo simulation in a SCRI model. • Examples for design application were reasonably drawn from the viewpoints of reliability. - Abstract: This paper presents reliability assessment of the long-term creep life of Gr. 91 steel, which is a major structural material for high temperature structural components of Generation-IV reactor systems. A number of creep rupture data for Gr. 91 steel were collected through literature surveys, and the long-term creep life was predicted by Larson–Miller parameter. A “Z parameter” method was used to describe the magnitude of the deviation of the creep rupture data to a master curve. A “Service Condition-creep Rupture property Interference (SCRI) model” based on the Z parameter was used to simultaneously consider the scattering of the creep rupture data of materials and the fluctuations of service conditions in reliability assessment. A statistical analysis of the creep rupture data was conducted by the Z parameter. To carry out the SCRI model, a number of random variables for Z s describing service conditions and Z cr describing the dispersion of the creep rupture data were generated using a Monte-Carlo simulation technique. As examples for application, the creep rupture life under a certain service conditions of Gr. 91 steel was reasonably drawn from the viewpoints of reliability

  2. Method of core thermodynamic reliability determination in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, G.; Horche, W. (Ingenieurhochschule Zittau (German Democratic Republic). Sektion Kraftwerksanlagenbau und Energieumwandlung)

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied.

  3. Method of core thermodynamic reliability determination in pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied. (author)

  4. Nondestructive examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-07-01

    The Evaluation and Improvement of NDE reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties

  5. Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1988-01-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactor (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1986 through September 1987

  6. Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1988-01-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1986 through September 1987. (author)

  7. Assessment of fusion reactor development. Proceedings

    International Nuclear Information System (INIS)

    Inoue, N.; Tazima, T.

    1994-04-01

    Symposium on assessment of fusion reactor development was held to make clear critical issues, which should be resolved for the commercial fusion reactor as a major energy source in the next century. Discussing items were as follows. (1) The motive force of fusion power development from viewpoints of future energy demand, energy resources and earth environment for 'Sustainable Development'. (2) Comparison of characteristics with other alternative energy sources, i.e. fission power and solar cell power. (3) Future planning of fusion research and advanced fuel fusion (D 3 He). (4) Critical issues of fusion reactor development such as Li extraction from the sea water, structural material and safety. (author)

  8. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  9. Criteria for the assessment of reactor potential

    International Nuclear Information System (INIS)

    Carruthers, R.

    1982-01-01

    This article outlines some of the more general criteria to be used in assessing reactor potential. The interdependence of plasma and engineering parameters is considered. This demonstrated how it is the first wall power loading which is the critical parameter in assessing economic prospects. Taking some of the current conceptual designs of fusion reactors and raising the wall loading to the value needed to approach a competitive cost leads to a very challenging set of parameters. Although developed in terms of a tokamak they are figures which are applicable more generally to fusion reactors which are toroidal in form. It is not at all obvious that the tokamak could ever satisfy this criterion of economic viability, so we should not be using the parameters of existing tokamak reactor designs as the basis for assessing alternative approaches. We need to see whether there is an alternative sufficiently different as to offer a better chance of reaching these more onerous parameters. Unfortunately, so many of the alternatives differ only in magnetic geometry and their physical geometry leads to the same problems as faced by the tokamak. The traditional approach -- devising intriguing ''boxes'' for studying the confinement of plasma and then speculating on their reactor potential -- should give way to new initiatives. What we need in the fusion program is more ''reactor relevance pull'' and less ''plasma physics push'' when planning future activities

  10. Reliability analysis of reactor inspection robot(RIROB)

    International Nuclear Information System (INIS)

    Eom, H. S.; Kim, J. H.; Lee, J. C.; Choi, Y. R.; Moon, S. S.

    2002-05-01

    This report describes the method and the result of the reliability analysis of RIROB developed in Korea Atomic Energy Research Institute. There are many classic techniques and models for the reliability analysis. These techniques and models have been used widely and approved in other industries such as aviation and nuclear industry. Though these techniques and models have been approved in real fields they are still insufficient for the complicated systems such RIROB which are composed of computer, networks, electronic parts, mechanical parts, and software. Particularly the application of these analysis techniques to digital and software parts of complicated systems is immature at this time thus expert judgement plays important role in evaluating the reliability of the systems at these days. In this report we proposed a method which combines diverse evidences relevant to the reliability to evaluate the reliability of complicated systems such as RIROB. The proposed method combines diverse evidences and performs inference in formal and in quantitative way by using the benefits of Bayesian Belief Nets (BBN)

  11. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin

    2015-06-28

    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  12. Assessment of Smart Reactor Utilization for Barelang

    International Nuclear Information System (INIS)

    Sahala-M-Lumbanraja; Yuliastuti

    2007-01-01

    This paper assesses the feasibility of SMART reactor utilization in BARELANG region. BARELANG region is an industrial area located in Riau Islands Province. The need of electricity and fresh water, whether for industry growth or people, are the main problem of this region. Until now, the National Electricity Company (PLN) has not able to supply the electricity needed by industrial sector. The use of oil as a main electricity generation resource of the entire power plant has caused a tremendous generation cost. On dry seasons, the fresh water supplied by PDAM is reducing drastically. This situation occurs because water source of PDAM extremely depends on the water storage during rainy seasons. SMART reactor is a modular light reactor developed by KAERI for dual purposes, producing electricity and fresh water at the same time. The total thermal power generated by this type of reactor is about 330 M Wth with 33 % efficiency, as 90 M We connected to the electricity grid and rest is used in producing potable water with capacity 40,000 m 3 /day. Compare to the conventional reactor, SMART reactor is based on simple operation and maintenance principles, enhanced safety, easy to inspect, a relatively short construction time, small investment cost, competitive generation cost, and a flexible design to fit with the existing infrastructure. The main characteristic of SMART reactor is an integral design concept where the entire main cooling system components are located in the pressurize vessel. (author)

  13. Assessment of modern methods of human factor reliability analysis in PSA studies

    International Nuclear Information System (INIS)

    Holy, J.

    2001-12-01

    The report is structured as follows: Classical terms and objects (Probabilistic safety assessment as a framework for human reliability assessment; Human failure within the PSA model; Basic types of operator failure modelled in a PSA study and analyzed by HRA methods; Qualitative analysis of human reliability; Quantitative analysis of human reliability used; Process of analysis of nuclear reactor operator reliability in a PSA study); New terms and objects (Analysis of dependences; Errors of omission; Errors of commission; Error forcing context); and Overview and brief assessment of human reliability analysis (Basic characteristics of the methods; Assets and drawbacks of the use of each of HRA method; History and prospects of the use of the methods). (P.A.)

  14. Independent assessment for new nuclear reactor safety

    Directory of Open Access Journals (Sweden)

    D'Auria Francesco

    2017-01-01

    Full Text Available A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs. Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry. The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

  15. Independent assessment for new nuclear reactor safety

    International Nuclear Information System (INIS)

    D'Auria, F.; Glaeser, H.; Debrecin, N.

    2017-01-01

    A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty (BEPU) approach. (authors)

  16. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  17. The control-and-instrumentation system of the IEA zero power reactor and its reliability calculation

    International Nuclear Information System (INIS)

    Peluso, M.A.V.

    1978-01-01

    The control-and instrumentation system for the Instituto de Energia Atomica Zero Power Reactor is described and the design criteria are presented and discussed. The reliability analysis for the reactor protection system was performed using the fault tree method. This was done using a computer code based on the Monte Carlo simulation. That code is an adaptation of the SAFTE-I, for the IBM 360/155 IEA Computer. (Author) [pt

  18. Lifetime Reliability Assessment of Concrete Slab Bridges

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle

    A procedure for lifetime assesment of the reliability of short concrete slab bridges is presented in the paper. Corrosion of the reinforcement is the deterioration mechanism used for estimating the reliability profiles for such bridges. The importance of using sensitivity measures is stressed....... Finally the produce is illustrated on 6 existing UK bridges....

  19. Reliability of leak detection systems in light water reactors

    International Nuclear Information System (INIS)

    Kupperman, D.S.

    1987-01-01

    US Nuclear Regulatory Commission Guide 1.45 recommends the use of at least three different detection methods in reactors to detect leakage. Monitoring of both sump-flow and airborne particulate radioactivity is recommended. A third method can involve either monitoring of condensate flow rate from air coolers or monitoring of airborne gaseous radioactivity. Although the methods currently used for leak detection reflect the state of the art, other techniques may be developed and used. Since the recommendations of Regulatory Guide 1.45 are not mandatory, the technical specifications for 74 operating plants have been reviewed to determine the types of leak detection methods employed. In addition, Licensee Event Report (LER) Compilations from June 1985 to June 1986 have been reviewed to help establish actual capabilities for detecting leaks and determining their source. Work at Argonne National Laboratory has demonstrated that improvements in leak detection, location, and sizing are possible with advanced acoustic leak detection technology

  20. A probabilistic consequence assessment for a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2017-02-15

    Currently, fossil fuel is globally running out. If current trends continue, crude oil will be depleted in 20 years and natural gas in 40 years. In addition, the use of fossil resource has increased emissions of green gas such as carbon dioxide. Therefore, there has been a strong demand in recent years for producing large amounts of hydrogen as an alternative energy [1]. To generate hydrogen energy, very high temperature more than 900 C is required but this level is not easy to reach. Because a Very High Temperature Reactor (VHTR), one of next generation reactor, is able to make the temperature, it is regarded as a solution of the problem. Also, VHTR has an excellent safety in comparison with existing and other next generation reactors. Especially, a passive system, Reactor Cavity Cooling System (RCCS), is adopted to get rid of radiant heat in case of accidents. To achieve variety requirements of new designed-reactors, however, it needs to develop new methodologies and definitions different with existing method. At the same time, an application of probability safety assessment (PSA) has been proposed to ensure the safety of next generation NPPs. For this, risk-informed designs of structures have to be developed and verified. Particularly, the passive system requires to be evaluated for its reliability. The objective of this study is to improve safety of VIITR by conducting risk profile.

  1. Evaluation of research reactor fuel reliability in support of regulatory requirements

    International Nuclear Information System (INIS)

    Sokolov, Eugene N.

    2005-01-01

    This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

  2. Evaluation of research reactor fuel reliability in support of regulatory requirements

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Eugene N [Chalk River Laboratories, AECL, Chalk River, ON, K0J 1J0 (Canada)

    2005-07-01

    This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

  3. Reactor fuel reliability constraints on power shape control

    International Nuclear Information System (INIS)

    Gelhaus, F.; Hallam, J.; Sauar, T.

    1977-01-01

    The level of reliability of fuel rods operating in commercial nuclear power plants has been less than desired for a number of reasons. Several of these causes have been successfully minimized, but pellet-clad interaction failures persist. Since power and power change are dominant parameters in this failure mode, restrictions on operational maneuvers have been recommended by all U. S. fuel suppliers. Slower-than-design allowable maneuvers decrease the plant capacity factor, which can cost a utility up to $7 million per year per plant. To assist utility engineering and operations personnel in their day-to-day decisions in this regard, the Electric Power Research Institute (EPRI) is initiating a project, designated RP895, to develop a fully computerized Power Shape Monitoring System (PSMS) for core-wide fuel rod reliability prediction. This paper describes the PSMS system and details some of the hardware/software requirements as they are now perceived. Salient results from a just-completed complementary EPRI-funded study, RP509, are described; this effort employed hand data acquisition and many man-machine interfaces that will be fully integrated and automated in the PSMS. The capabilities of the PSMS will derive from the use of modern minicomputer hardware and software and from accurate computational modules that enable near-real-time predictive capability

  4. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  5. Small nuclear power reactor emergency electric power supply system reliability comparative analysis

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

    This work presents an analysis of the reliability of the emergency power supply system, of a small size nuclear power reactor. Three different configurations are investigated and their reliability analyzed. The fault tree method is used as the main tool of analysis. The work includes a bibliographic review of emergency diesel generator reliability and a discussion of the design requirements applicable to emergency electrical systems. The influence of common cause failure influences is considered using the beta factor model. The operator action is considered using human failure probabilities. A parametric analysis shows the strong dependence between the reactor safety and the loss of offsite electric power supply. It is also shown that common cause failures can be a major contributor to the system reliability. (author)

  6. Computer-aided reliability and risk assessment

    International Nuclear Information System (INIS)

    Leicht, R.; Wingender, H.J.

    1989-01-01

    Activities in the fields of reliability and risk analyses have led to the development of particular software tools which now are combined in the PC-based integrated CARARA system. The options available in this system cover a wide range of reliability-oriented tasks, like organizing raw failure data in the component/event data bank FDB, performing statistical analysis of those data with the program FDA, managing the resulting parameters in the reliability data bank RDB, and performing fault tree analysis with the fault tree code FTL or evaluating the risk of toxic or radioactive material release with the STAR code. (orig.)

  7. Assessing the reliability of skilled performance

    International Nuclear Information System (INIS)

    Dougherty, E.M.; Collins, E.P.

    1996-01-01

    In some high-hazard military tasks, precision of motor skills is crucial to avoid a risk-sensitive drop or other mishap. Sparse data supports a basic level of reliability for this activity seemingly much lower than typical, individual action, slip-mode errors would yield. A typical occurrence failure rate seems to be of the order of 1 x 10 -5 per opportunity. A speculation is tendered that this represents a difference in kind; that the reliability of skills is a different taxon than previously called out in the typical scheme of human reliability

  8. Different Reliability Assessment Approaches for Wave Energy Converters

    DEFF Research Database (Denmark)

    Ambühl, Simon; Kramer, Morten Mejlhede; Sørensen, John Dalsgaard

    2015-01-01

    Reliability assessments are of importance for wave energy converters (WECs) due to the fact that accessibility might be limited in case of failure and maintenance. These failure rates can be adapted by reliability considerations. There are two different approaches to how reliability can...

  9. Probabilistic assessment of pressure vessel and piping reliability

    International Nuclear Information System (INIS)

    Sundararajan, C.

    1986-01-01

    The paper presents a critical review of the state-of-the-art in probabilistic assessment of pressure vessel and piping reliability. First the differences in assessing the reliability directly from historical failure data and indirectly by a probabilistic analysis of the failure phenomenon are discussed and the advantages and disadvantages are pointed out. The rest of the paper deals with the latter approach of reliability assessment. Methods of probabilistic reliability assessment are described and major projects where these methods are applied for pressure vessel and piping problems are discussed. An extensive list of references is provided at the end of the paper

  10. Parts and Components Reliability Assessment: A Cost Effective Approach

    Science.gov (United States)

    Lee, Lydia

    2009-01-01

    System reliability assessment is a methodology which incorporates reliability analyses performed at parts and components level such as Reliability Prediction, Failure Modes and Effects Analysis (FMEA) and Fault Tree Analysis (FTA) to assess risks, perform design tradeoffs, and therefore, to ensure effective productivity and/or mission success. The system reliability is used to optimize the product design to accommodate today?s mandated budget, manpower, and schedule constraints. Stand ard based reliability assessment is an effective approach consisting of reliability predictions together with other reliability analyses for electronic, electrical, and electro-mechanical (EEE) complex parts and components of large systems based on failure rate estimates published by the United States (U.S.) military or commercial standards and handbooks. Many of these standards are globally accepted and recognized. The reliability assessment is especially useful during the initial stages when the system design is still in the development and hard failure data is not yet available or manufacturers are not contractually obliged by their customers to publish the reliability estimates/predictions for their parts and components. This paper presents a methodology to assess system reliability using parts and components reliability estimates to ensure effective productivity and/or mission success in an efficient manner, low cost, and tight schedule.

  11. Theoretical study of fuel element reliability in the BRIG-300 fast reactor

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Nesterenko, V.B.; Tverkovkin, B.E.

    1983-01-01

    The theoretical results on studies of the reliability of cermet symmetrically heated fuel elements under conditions of the BRIG-300 fast gas cooled reactor are presented. The investigations have been conducted at the Nuclear Power Engineering Institute of the Byelorussian Academy of Sciences. Two variants of the fuel elements are considered :the fuel element with the gas gap between fuel and can and the fuel element with tight contact between cermet fuel and can. The estimated data on can resistance, swelling of the fuel rods and cans, strains and stresses in cans, change of the gap and its thermal coductivity during the reactor operation are obtained. The results of the analysis show that cermet fuel has sufficient reliability upon oparational conditions of the reactor with dissociating gas coolant in a steady-state regime

  12. Evaluation and improvement of nondestructive evaluation reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Bates, D.J.; Deffenbaugh, J.D.; Good, M.S.; Heasler, P.G.; Mart, G.A.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Van Fleet, L.G.

    1987-01-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection (ISI) of Light Water Reactors (NDE Reliability) Program at Pacific Northwest Laboratory (PNL) was established to determine the reliability of current ISI techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this NRC program are to: determine the reliability of ultrasonic ISI performed on commercial light-water reactor (LWR) primary systems, using probabilistic fracture mechanics analysis, determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to ensure a suitably low failure probability, evaluate the degree of reliability improvement that could be achieved using improved and advanced NDE techniques, based on material properties, service conditions, and NDE uncertainties, recommend revisions to ASME Code, Section XI, and Regulatory Requirements that will ensure suitably low failure probabilities. The scope of this program is limited to ISI of primary systems; the results and recommendations may also be applicable to Class II piping systems

  13. Condition-based Human Reliability Assessment for digitalized control room

    International Nuclear Information System (INIS)

    Kang, H. G.; Jang, S. C.; Eom, H. S.; Ha, J. J.

    2005-04-01

    In safety-critical systems, the generation failure of an actuation signal is caused by the concurrent failures of the automated systems and an operator action. These two sources of safety signals are complicatedly correlated. The failures of sensors or automated systems will cause a lack of necessary information for a human operator and result in error-forcing contexts such as the loss of corresponding alarms and indications. In the conventional analysis, the Human Error Probabilities (HEP) are estimated based on the assumption of 'normal condition of indications and alarms'. In order to construct a more realistic signal-generation failure model, we have to consider more complicated conditions in a more realistic manner. In this study, we performed two kinds of investigation for addressing this issue. We performed the analytic calculations for estimating the effect of sensors failures on the system unavailability and plant risk. For the single-parameter safety signals, the analysis result reveals that the quantification of the HEP should be performed by focusing on the 'no alarm from the automatic system and corresponding indications unavailable' situation. This study also proposes a Condition-Based Human Reliability Assessment (CBHRA) method in order to address these complicated conditions in a practical way. We apply the CBHRA method to the manual actuation of the safety features such as a reactor trip and auxiliary feedwater actuation in Korean Standard Nuclear Power Plants. In the case of conventional single HEP method, it is very hard to consider the multiple HE conditions. The merit of CBHRA is clearly shown in the application to the AFAS generation where no dominating HE condition exits. In this case, even if the HE conditions are carefully investigated, the single HEP method cannot accommodate the multiple conditions in a fault tree. On the other hand, the application result of the reactor trip in SLOCA shows that if there is a dominating condition, the use

  14. Replacement energy, capacity, and reliability costs for permanent nuclear reactor shutdowns

    International Nuclear Information System (INIS)

    VanKuiken, J.C., Buehring, W.A.; Hamilton, S.; Kavicky, J.A.; Cavallo, J.D.; Veselka, T.D.; Willing, D.L.

    1993-10-01

    Average replacement power costs are estimated for potential permanent shutdowns of nuclear electricity-generating units. Replacement power costs are considered to include replacement energy, capacity, and reliability cost components. These estimates were developed to assist the US Nuclear Regulatory Commission in evaluating regulatory issues that potentially affect changes in serious reactor accident frequencies. Cost estimates were derived from long-term production-cost and capacity expansion simulations of pooled utility-system operations. Factors that affect replacement power cost, such as load growth, replacement sources of generation, and capital costs for replacement capacity, were treated in the analysis. Costs are presented for a representative reactor and for selected subcategories of reactors, based on estimates for 112 individual reactors

  15. Preliminary investigation on reliability assessment of passive safety system

    International Nuclear Information System (INIS)

    Huang Changfan; Kuang Bo

    2012-01-01

    The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)

  16. Reliability Evaluation on Creep Life Prediction of Alloy 617 for a Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Kim, Woo-Gon; Hong, Sung-Deok; Kim, Yong-Wan; Park, Jae-Young; Kim, Seon-Jin

    2012-01-01

    This paper evaluates the reliability of creep rupture life under service conditions of Alloy 617, which is considered as one of the candidate materials for use in a very high temperature reactor (VHTR) system. A Z-parameter, which represents the deviation of creep rupture data from the master curve, was used for the reliability analysis of the creep rupture data of Alloy 617. A Service-condition Creep Rupture Interference (SCRI) model, which can consider both the scattering of the creep rupture data and the fluctuations of temperature and stress under any service conditions, was also used for evaluating the reliability of creep rupture life. The statistical analysis showed that the scattering of creep rupture data based on Z-parameter was supported by normal distribution. The values of reliability decreased rapidly with increasing amplitudes of temperature and stress fluctuations. The results established that the reliability decreased with an increasing service time.

  17. A survey on the human reliability analysis methods for the design of Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W

    2000-03-01

    Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design.

  18. A survey on the human reliability analysis methods for the design of Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W.

    2000-03-01

    Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design

  19. Reliability assessment of underground shaft closure

    International Nuclear Information System (INIS)

    Fossum, A.F.; Munson, D.E.

    1994-01-01

    The intent of the WIPP, being constructed in the bedded geologic salt deposits of Southeastern New Mexico, is to provide the technological basis for the safe disposal of radioactive Transuranic (TRU) wastes generated by the defense programs of the United States. In determining this technological basis, advanced reliability and structural analysis techniques are used to determine the probability of time-to-closure of a hypothetical underground shaft located in an argillaceous salt formation and filled with compacted crushed salt. Before being filled with crushed salt for sealing, the shaft provides access to an underground facility. Reliable closure of the shaft depends upon the sealing of the shaft through creep closure and recompaction of crushed backfill. Appropriate methods are demonstrated to calculate cumulative distribution functions of the closure based on laboratory determined random variable uncertainty in salt creep properties

  20. Risk and reliability assessment for telecommunications networks

    Energy Technology Data Exchange (ETDEWEB)

    Wyss, G.D.; Schriner, H.K.; Gaylor, T.R.

    1996-08-01

    Sandia National Laboratories has assembled an interdisciplinary team to explore the applicability of probabilistic logic modeling (PLM) techniques to model network reliability for a wide variety of communications network architectures. The authors have found that the reliability and failure modes of current generation network technologies can be effectively modeled using fault tree PLM techniques. They have developed a ``plug-and-play`` fault tree analysis methodology that can be used to model connectivity and the provision of network services in a wide variety of current generation network architectures. They have also developed an efficient search algorithm that can be used to determine the minimal cut sets of an arbitrarily-interconnected (non-hierarchical) network without the construction of a fault tree model. This paper provides an overview of these modeling techniques and describes how they are applied to networks that exhibit hybrid network structures (i.e., a network in which some areas are hierarchical and some areas are not hierarchical).

  1. Structural reliability analysis and seismic risk assessment

    International Nuclear Information System (INIS)

    Hwang, H.; Reich, M.; Shinozuka, M.

    1984-01-01

    This paper presents a reliability analysis method for safety evaluation of nuclear structures. By utilizing this method, it is possible to estimate the limit state probability in the lifetime of structures and to generate analytically the fragility curves for PRA studies. The earthquake ground acceleration, in this approach, is represented by a segment of stationary Gaussian process with a zero mean and a Kanai-Tajimi Spectrum. All possible seismic hazard at a site represented by a hazard curve is also taken into consideration. Furthermore, the limit state of a structure is analytically defined and the corresponding limit state surface is then established. Finally, the fragility curve is generated and the limit state probability is evaluated. In this paper, using a realistic reinforced concrete containment as an example, results of the reliability analysis of the containment subjected to dead load, live load and ground earthquake acceleration are presented and a fragility curve for PRA studies is also constructed

  2. Suprahyoid Muscle Complex: A Reliable Neural Assessment Tool For Dysphagia?

    DEFF Research Database (Denmark)

    Kothari, Mohit; Stubbs, Peter William; Pedersen, Asger Roer

    be a non-invasive reliable neural assessment tool for patients with dysphagia. Objective: To investigate the possibility of using the suprahyoid muscle complex (SMC) using surface electromyography (sEMG) to assess changes to neural pathways by determining the reliability of measurements in healthy...

  3. Reliability of assessment of adherence to an antimicrobial treatment guideline

    NARCIS (Netherlands)

    Mol, PGM; Gans, ROB; Panday, PVN; Degener, JE; Laseur, M; Haaijer-Ruskamp, FM

    Assessment procedures for adherence to a guideline must be reliable and credible. The aim of this study was to explore the reliability of assessment of adherence, taking account of the professional backgrounds of the observers. A secondary analysis explored the impact of case characteristics on

  4. Reliability assessment using the concept of possibility

    International Nuclear Information System (INIS)

    Ping, R.; Zhao, G.

    1994-01-01

    The role of the possibility theory in the quantitative approach to the reliability of large scale systems, in particular man-machine systems, is discussed. According to the original idea due to professor L.A. Zadeh, an error rate may be associated with a possibility distribution and a probability distribution. They describe the potentiallity and the actuality for error. This is especially important for the analysis of big accidents

  5. Scram reliability under seismic conditions at the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Roglans, J.; Wang, C.Y.; Hill, D.J.

    1993-01-01

    A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

  6. The evaluation of equipment and Instrumentation Reliability Factors on Power Reactor

    International Nuclear Information System (INIS)

    Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Santosa, Kussigit; Darlis; Sudiyono, Bambang; Yuniyanta, Sasongko; Sudarmin

    1999-01-01

    Equipment and instrumentation reliability on type power reactor control room was determined by its pattern and design. the principle of ergonomy applied on equipment and instrumentation layout in this ABWR type reactor are geometric pattern appropriate with economic body motion, average anthropometry data of operator especially operator hand-reach, range of vision, angle of vision, lighting, color arrangement and harmony as will as operator case in operating the equipment system. Limitation criteria of the parameter mentioned above are based on EPRI NP-3659, NURG 0700, and NUREG/CR-3331 documents. Besides that, the (working) physical environment parameter factor of the control room must be designed in order to fulfil the standard criteria of ergonomic condition based on NUREG-0800. The reliability evaluation of equipment and instrumentation system also occurs observed from man machine interaction side which happen between operator and equipment and instrumentation in the ABWR type power reactor control room. From the MMI analysis can be known the working failure possibility which is caused by the operator. The evaluation result of equipment and instrumentation reliability on ABWR type power reactor control room showed that the design of this ABWR control room is good and fulfils the ergonomy standard criteria have been determined

  7. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V.

    1993-11-01

    The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992

  8. Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1991-08-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 6 of the ASME Code. This is a progress report covering the pro grammatic work from April 1989 through September 1989. 12 refs., 4 figs. 5 tabs

  9. Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1989-11-01

    Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs

  10. Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1989-10-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvement that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements based on material properties, service conditions, and NDE capabilities and uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1987 through March 1988. 21 refs., 28 figs., 2 tabs

  11. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-07-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWR's); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from April 1991 through September 1991

  12. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.

    1993-09-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992

  13. Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

    International Nuclear Information System (INIS)

    Mohamed, F.; Hassan, A.; Yahaya, R.; Rahman, I.; Maskin, M.; Praktom, P.; Charlie, F.

    2015-01-01

    Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification

  14. Contributions to the fundamental optimization and higher reliability of nuclear reactor complex control systems

    International Nuclear Information System (INIS)

    Tapu, C.

    1979-01-01

    The reference control system automatically adjusted to the reactor state and the reliability analysis of the control systems which survive the first defect with given solution for detecting all kinds of defects in the system with parallel redundancy, supply additional data on the high performance and largely available solutions. The results of the paper have been applied in the following cases: The reference control system automatically adapted to the reactor state has been tested with the VVR-S reactor and was patented. The solutions advanced for period and reactivity monitoring by the numerical technique were worked up during the training in France-CEN-Saclay and have been applied on the Ulysse reactor at INSTN (CERN-Saclay). At the same time, the papers published on these subjects have been cited in ''Revue General d'Electricite (February 1976)'' and in ''MCH/MENT 10 - Materiel Electronique Nucleaire pour tableau de commande et de controle (1976)'' issued by CEA France, as marking a progress in the control of nuclear power plants. The reliability analysis of the control system with parallel redundancy was performed during a specialization in France and on the basis of this analysis, the control system for the high flux reactor in Grenoble was selected. (author)

  15. Extension and update of the reliability data for fire protection equipment in German light water reactors; Ergaenzung und Aktualisierung von Zuverlaessigkeitskenngroessen fuer Brandschutzeinrichtungen in deutschen Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Forell, Burkhard; Einarsson, Svante

    2016-12-15

    Within the BMUB project 3610R01370 ''Extension and update of reliability data for fire protection equipment in German light water reactors'' plant-specific as well as generic failure rates for the technical reliability of active fire protection features in German nuclear power plants have been calculated. Based on results of previous projects, in this project observation times of components were updated and extended and additional components and functions were assessed. Now, the data evaluated results from a total of six German reference plants with seven reactor units.

  16. Extrapolation Method for System Reliability Assessment

    DEFF Research Database (Denmark)

    Qin, Jianjun; Nishijima, Kazuyoshi; Faber, Michael Havbro

    2012-01-01

    of integrals with scaled domains. The performance of this class of approximation depends on the approach applied for the scaling and the functional form utilized for the extrapolation. A scheme for this task is derived here taking basis in the theory of asymptotic solutions to multinormal probability integrals......The present paper presents a new scheme for probability integral solution for system reliability analysis, which takes basis in the approaches by Naess et al. (2009) and Bucher (2009). The idea is to evaluate the probability integral by extrapolation, based on a sequence of MC approximations...... that the proposed scheme is efficient and adds to generality for this class of approximations for probability integrals....

  17. Reliability and maintainability assessment factors for reliable fault-tolerant systems

    Science.gov (United States)

    Bavuso, S. J.

    1984-01-01

    A long term goal of the NASA Langley Research Center is the development of a reliability assessment methodology of sufficient power to enable the credible comparison of the stochastic attributes of one ultrareliable system design against others. This methodology, developed over a 10 year period, is a combined analytic and simulative technique. An analytic component is the Computer Aided Reliability Estimation capability, third generation, or simply CARE III. A simulative component is the Gate Logic Software Simulator capability, or GLOSS. The numerous factors that potentially have a degrading effect on system reliability and the ways in which these factors that are peculiar to highly reliable fault tolerant systems are accounted for in credible reliability assessments. Also presented are the modeling difficulties that result from their inclusion and the ways in which CARE III and GLOSS mitigate the intractability of the heretofore unworkable mathematics.

  18. Analysis and assessment of water treatment plant reliability

    Directory of Open Access Journals (Sweden)

    Szpak Dawid

    2017-03-01

    Full Text Available The subject of the publication is the analysis and assessment of the reliability of the surface water treatment plant (WTP. In the study the one parameter method of reliability assessment was used. Based on the flow sheet derived from the water company the reliability scheme of the analysed WTP was prepared. On the basis of the daily WTP work report the availability index Kg for the individual elements included in the WTP, was determined. Then, based on the developed reliability scheme showing the interrelationships between elements, the availability index Kg for the whole WTP was determined. The obtained value of the availability index Kg was compared with the criteria values.

  19. Assessment of Sensor Technologies for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, Kofi [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vlim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Britton, Jr, Charles L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wootan, D. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anheier, Jr, N. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, E. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chien, H. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Sheen, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States); Gopalsami, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Heifetz, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Tam, S. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Upadhyaya, B. R. [Univ. of Tennessee, Knoxville, TN (United States); Stanford, A. [Univ. of Tennessee, Knoxville, TN (United States)

    2016-10-01

    Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

  20. N reactor external events probabilistic risk assessment

    International Nuclear Information System (INIS)

    Baxter, J.T.

    1989-01-01

    An external events probabilistic risk assessment of the N Reactor has been completed. The methods used are those currently being proposed for external events analysis in NUREG-1150. Results are presented for the external hazards that survived preliminary screening. They are earthquake, fire, and external flood. Core damage frequencies for these hazards are shown to be comparable to those for commercial pressurized water reactors. Dominant fire sequences are described and related to 10 CFR 50, Appendix R design requirements. Potential remedial measures that reduce fire core damage risk are described including modifications to fire protection systems, procedure changes, and addition of new administrative controls. Dominant seismic sequences are described. The effect of non-safety support system dependencies on seismic risk is presented

  1. Reliability analysis of protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

    Varde, P. V.; Choi, J. G.; Lee, D. Y.; Han, J. B.

    2003-04-01

    Reliability analysis was carried out for the protection system of the Korean Advanced Pressurized Water Reactor - APR 1400. The main focus of this study was the reliability analysis of digital protection system, however, towards giving an integrated statement of complete protection reliability an attempt has been made to include the shutdown devices and other related aspects based on the information available to date. The sensitivity analysis has been carried out for the critical components / functions in the system. Other aspects like importance analysis and human error reliability for the critical human actions form part of this work. The framework provided by this study and the results obtained shows that this analysis has potential to be utilized as part of risk informed approach for future design / regulatory applications

  2. Reliability analysis of diverse safety logic systems of fast breeder reactor

    International Nuclear Information System (INIS)

    Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

    2006-01-01

    Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)

  3. Power probability density function control and performance assessment of a nuclear research reactor

    International Nuclear Information System (INIS)

    Abharian, Amir Esmaeili; Fadaei, Amir Hosein

    2014-01-01

    Highlights: • In this paper, the performance assessment of static PDF control system is discussed. • The reactor PDF model is set up based on the B-spline functions. • Acquaints of Nu, and Th-h. equations solve concurrently by reformed Hansen’s method. • A principle of performance assessment is put forward for the PDF of the NR control. - Abstract: One of the main issues in controlling a system is to keep track of the conditions of the system function. The performance condition of the system should be inspected continuously, to keep the system in reliable working condition. In this study, the nuclear reactor is considered as a complicated system and a principle of performance assessment is used for analyzing the performance of the power probability density function (PDF) of the nuclear research reactor control. First, the model of the power PDF is set up, then the controller is designed to make the power PDF for tracing the given shape, that make the reactor to be a closed-loop system. The operating data of the closed-loop reactor are used to assess the control performance with the performance assessment criteria. The modeling, controller design and the performance assessment of the power PDF are all applied to the control of Tehran Research Reactor (TRR) power in a nuclear process. In this paper, the performance assessment of the static PDF control system is discussed, the efficacy and efficiency of the proposed method are investigated, and finally its reliability is proven

  4. 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

  5. Assessing the Impact of Imperfect Diagnosis on Service Reliability

    DEFF Research Database (Denmark)

    Grønbæk, Lars Jesper; Schwefel, Hans-Peter; Kjærgaard, Jens Kristian

    2010-01-01

    , representative diagnosis performance metrics have been defined and their closed-form solutions obtained for the Markov model. These equations enable model parameterization from traces of implemented diagnosis components. The diagnosis model has been integrated in a reliability model assessing the impact...... of the diagnosis functions for the studied reliability problem. In a simulation study we finally analyze trade-off properties of diagnosis heuristics from literature, map them to the analytic Markov model, and investigate its suitability for service reliability optimization....

  6. Development and Reliability Analysis of HTR-PM Reactor Protection System

    International Nuclear Information System (INIS)

    Li Duo; Guo Chao; Xiong Huasheng

    2014-01-01

    High Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM) digital Reactor Protection System (RPS) is a dedicated system, which is designed and developed according to HTR-PM NPP protection specifications. To decrease the probability of accident trips and increase the system reliability, HTR-PM RPS has such features as a framework of four redundant channels, two diverse sub-systems in each channel, and two level two-out-of-four logic voters. Reliability analysis of HTR-PM RPS is based on fault tree model. A fault tree is built based on HTR-PM RPS Failure Modes and Effects Analysis (FMEA), and special analysis is focused on the sub-tree of redundant channel ''2-out-of-4'' logic and the fault tree under one channel is bypassed. The qualitative analysis of fault tree, such as RPS weakness according to minimal cut sets, is summarized in the paper. (author)

  7. Study on highly reliable digital communication technology of reactor nuclear measuring equipment

    International Nuclear Information System (INIS)

    Gu Pengfei; Huang Xiaojin

    2007-01-01

    To meet the need of highly reliable of reactor nuclear measuring equipment, in allusion to the idiographic request of nuclear measuring equipment, the actual technical development and the application in industrial field, we design a kind of redundancy communication net based on PROFIBUS, and a kind of communication interface module based on redundancy PROFIBUS communication, which link the nuclear measuring equipment and PROFIBUS communication net, and also lay a foundation for advanced research. (authors)

  8. The protection system of nuclear reactors. Example of a very high reliability system. Its evolution

    International Nuclear Information System (INIS)

    Weill, J.

    1980-06-01

    The present state of reactor protection is described and mention is made of certain evolutionary trends towards completely automated systems which either help the operator to take decisions in the event of an accident or take and execute these decisions for him. To do so, the use of models and recourse to complex data processing systems is necessary. This ensemble reflects an evolution of the reliability of the equipment towards that of the software [fr

  9. Advanced Test Reactor outage risk assessment

    International Nuclear Information System (INIS)

    Thatcher, T.A.; Atkinson, S.A.

    1997-01-01

    Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance

  10. Probabilistic simulation applications to reliability assessments

    International Nuclear Information System (INIS)

    Miller, Ian; Nutt, Mark W.; Hill, Ralph S. III

    2003-01-01

    Probabilistic risk/reliability (PRA) analyses for engineered systems are conventionally based on fault-tree methods. These methods are mature and efficient, and are well suited to systems consisting of interacting components with known, low probabilities of failure. Even complex systems, such as nuclear power plants or aircraft, are modeled by the careful application of these approaches. However, for systems that may evolve in complex and nonlinear ways, and where the performance of components may be a sensitive function of the history of their working environments, fault-tree methods can be very demanding. This paper proposes an alternative method of evaluating such systems, based on probabilistic simulation using intelligent software objects to represent the components of such systems. Using a Monte Carlo approach, simulation models can be constructed from relatively simple interacting objects that capture the essential behavior of the components that they represent. Such models are capable of reflecting the complex behaviors of the systems that they represent in a natural and realistic way. (author)

  11. Meeting of Specialists on the Reliability of Decay Heat Removal Systems for Fast Reactors. Summary Report

    International Nuclear Information System (INIS)

    1975-10-01

    The Specialists Meeting on Reliability of Decay Heat Removal Systems proposed for Fast Reactors was sponsored by the UKAEA Safety & Reliability Directorate and held at Harwell between 28th April and 1st May, 1975. The meeting was attended by delegates from six countries - (USA, Federal Republic of Germany, France, Japan, USSR and the UK). A list of participants is included in an Appendix to this report. The subject matter of the meeting was concerned with the degree to which the ability to maintain decay heat removal from a fast reactor after shutdown in normal and abnormal circumstances could be guaranteed by design provisions and substantiated by reliability analysis techniques, operational testing etc. Consideration of conditions prevailing after a hypothetical core melt down incident were not included in the subject matter. The deliberations of the meeting were focussed at each working session on a defined theme and its dependant topics as shown in the detailed Agenda included in this report. Although provision had been made in the Agenda for a limited amount of discussion of the decay heat rejection problems of Gas Cooled Fast Reactors, delegates had no contributions to offer on this subject. During each session a Recording Secretary prepared a summary of the main points made by national delegates and of the resulting recommendations and conclusions. These draft summaries were made available to delegates during subsequent sessions of the meeting and approved by them for inclusion in the Summary, General Conclusions and Recommendations provided under Table of Contents (item 3 and 4)

  12. A Reliability Assessment Method for the VHTR Safety Systems

    International Nuclear Information System (INIS)

    Lee, Hyung Sok; Jae, Moo Sung; Kim, Yong Wan

    2011-01-01

    The Passive safety system by very high temperature reactor which has attracted worldwide attention in the last century is the reliability safety system introduced for the improvement in the safety of the next generation nuclear power plant design. The Passive system functionality does not rely on an external source of energy, but on an intelligent use of the natural phenomena, such as gravity, conduction and radiation, which are always present. Because of these features, it is difficult to evaluate the passive safety on the risk analysis methodology having considered the existing active system failure. Therefore new reliability methodology has to be considered. In this study, the preliminary evaluation and conceptualization are tried, applying the concept of the load and capacity from the reliability physics model, designing the new passive system analysis methodology, and the trial applying to paper plant.

  13. Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

    Varde, P. V.; Lee, D. Y.; Han, J. B.

    2003-03-01

    A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

  14. A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods

    International Nuclear Information System (INIS)

    Benko, Pedro Luiz

    1997-01-01

    A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

  15. Systems reliability analysis: applications of the SPARCS System-Reliability Assessment Computer Program

    International Nuclear Information System (INIS)

    Locks, M.O.

    1978-01-01

    SPARCS-2 (Simulation Program for Assessing the Reliabilities of Complex Systems, Version 2) is a PL/1 computer program for assessing (establishing interval estimates for) the reliability and the MTBF of a large and complex s-coherent system of any modular configuration. The system can consist of a complex logical assembly of independently failing attribute (binomial-Bernoulli) and time-to-failure (Poisson-exponential) components, without regard to their placement. Alternatively, it can be a configuration of independently failing modules, where each module has either or both attribute and time-to-failure components. SPARCS-2 also has an improved super modularity feature. Modules with minimal-cut unreliabiliy calculations can be mixed with those having minimal-path reliability calculations. All output has been standardized to system reliability or probability of success, regardless of the form in which the input data is presented, and whatever the configuration of modules or elements within modules

  16. 75 FR 66038 - Planning Resource Adequacy Assessment Reliability Standard

    Science.gov (United States)

    2010-10-27

    ..., providing a common framework for resource adequacy analysis, assessment, and documentation) effectively and...://www.nerc.com/commondocs.php?cd=2 . D. Proposed Effective Date 24. Proposed regional Reliability...

  17. A Mechanistic Reliability Assessment of RVACS and Metal Fuel Inherent Reactivity Feedbacks

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Brunett, Acacia J.; Passerini, Stefano; Grelle, Austin

    2017-09-24

    GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory (Argonne) participated in a two year collaboration to modernize and update the probabilistic risk assessment (PRA) for the PRISM sodium fast reactor. At a high level, the primary outcome of the project was the development of a next-generation PRA that is intended to enable risk-informed prioritization of safety- and reliability-focused research and development. A central Argonne task during this project was a reliability assessment of passive safety systems, which included the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedbacks of the metal fuel core. Both systems were examined utilizing a methodology derived from the Reliability Method for Passive Safety Functions (RMPS), with an emphasis on developing success criteria based on mechanistic system modeling while also maintaining consistency with the Fuel Damage Categories (FDCs) of the mechanistic source term assessment. This paper provides an overview of the reliability analyses of both systems, including highlights of the FMEAs, the construction of best-estimate models, uncertain parameter screening and propagation, and the quantification of system failure probability. In particular, special focus is given to the methodologies to perform the analysis of uncertainty propagation and the determination of the likelihood of violating FDC limits. Additionally, important lessons learned are also reviewed, such as optimal sampling methodologies for the discovery of low likelihood failure events and strategies for the combined treatment of aleatory and epistemic uncertainties.

  18. Assessment of Power Quality Problems for TRIGA PUSPATI Reactor (RTP)

    International Nuclear Information System (INIS)

    Mohd Fazli Zakaria; Ramachandaramurthy, V.K.

    2016-01-01

    The electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality is necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. (author)

  19. Reliability Prediction Of System And Component Of Process System Of RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Sitorus Pane, Jupiter

    2001-01-01

    The older the reactor the higher the probability of the system and components suffer from loss of function or degradation. This phenomenon occurred because of wear, corrosion, and fatigue. Study on component reliability was generally performed deterministically and statistically. This paper would describe an analysis of using statistical method, i.e. regression Cox, in order to predict the reliability of the components and their environmental influence's factors. The result showed that the dynamics, non safety related, and mechanic components have higher risk of failure, whereas static, safety related, and electric have lower risk of failures. The relative risk value for variable of components dynamics, quality, dummy 1 and dummy 2 are of 1.54, 1.59, 1.50, and 0.83 compare to other components type with each variable. Component with the higher risk have lower reliability than lower one

  20. Software coding for reliable data communication in a reactor safety system

    International Nuclear Information System (INIS)

    Maghsoodi, R.

    1978-01-01

    A software coding method is proposed to improve the communication reliability of a microprocessor based fast-reactor safety system. This method which replaces the conventional coding circuitry, applies a program to code the data which is communicated between the processors via their data memories. The system requirements are studied and the suitable codes are suggested. The problems associated with hardware coders, and the advantages of software coding methods are discussed. The product code which proves a faster coding time over the cyclic code is chosen as the final code. Then the improvement of the communication reliability is derived for a processor and its data memory. The result is used to calculate the reliability improvement of the processing channel as the basic unit for the safety system. (author)

  1. Thermionic conversion reactor technology assessment. Final report

    International Nuclear Information System (INIS)

    1984-02-01

    The in-core thermionic space nuclear power supply may be the only identified reactor-power concept that can meet the SP-100 size functional requirements with demonstrated state-of-the-art reactor system and space-qualified power system component temperatures. The SP-100 configuration limits provide a net 40 m 2 of primary non-deployed radiator area. If a reasonable 7-year degradation allowance of 15% to 20% is provided then the beginning of life (BOL) net power output requirement is about 120 kWe. Consequently, the SP-100 power system must produce a P/A of 2.7 kWe/m 2 . This non-deployed radiator area power density performance can only be reasonably achieved by the thermionic in-core convertr system, the potassium Rankine turbine system and the Stirling engine system. The purpose of this study is to examine past and current tests and data, and to assess the potential for successful development of suitable fueled-thermionic converters that will meet SP-100 and growth requirements. The basis for the assessment will be provided and the recommended key developments plan set forth

  2. Piping reliability model development, validation and its applications to light water reactor piping

    International Nuclear Information System (INIS)

    Woo, H.H.

    1983-01-01

    A brief description is provided of a three-year effort undertaken by the Lawrence Livermore National Laboratory for the piping reliability project. The ultimate goal of this project is to provide guidance for nuclear piping design so that high-reliability piping systems can be built. Based on the results studied so far, it is concluded that the reliability approach can undoubtedly help in understanding not only how to assess and improve the safety of the piping systems but also how to design more reliable piping systems

  3. Reliability and comparison of acromion assessment techniques on x-ray and magnetic resonance imaging (reliability of acromion assessment techniques)

    International Nuclear Information System (INIS)

    Viskontas, D.G.; MacDermid, J.C.; Drosdowech, D.S.; Garvin, G.J.; Romano, W.M.; Faber, K.J.

    2005-01-01

    To determine the reliability and correlation of plain radiography and magnetic resonance imaging (MRI) in the assessment of acromion morphology. Materials and Methods: Acromion morphology was assessed using the lateral acromion angle (LAA) and the acromion-humeral interval (AHI). Thirty patients who had x-rays and MRI for impingement syndrome were included. Six blinded observers assessed the acromion morphology subjectively and objectively. Results: Neither acromion assessment technique demonstrated a positive correlation (kappa and intraclass coefficient 0.55) when measured objectively by experienced observers. Conclusion: The LAA and the AHI are both reliable acromion assessment techniques on X-ray and MRI when measured objectively and by experienced observers. (author)

  4. Assessment of nuclear reactor concepts for low power space applications

    Science.gov (United States)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  5. Sequential optimization and reliability assessment method for metal forming processes

    International Nuclear Information System (INIS)

    Sahai, Atul; Schramm, Uwe; Buranathiti, Thaweepat; Chen Wei; Cao Jian; Xia, Cedric Z.

    2004-01-01

    Uncertainty is inevitable in any design process. The uncertainty could be due to the variations in geometry of the part, material properties or due to the lack of knowledge about the phenomena being modeled itself. Deterministic design optimization does not take uncertainty into account and worst case scenario assumptions lead to vastly over conservative design. Probabilistic design, such as reliability-based design and robust design, offers tools for making robust and reliable decisions under the presence of uncertainty in the design process. Probabilistic design optimization often involves double-loop procedure for optimization and iterative probabilistic assessment. This results in high computational demand. The high computational demand can be reduced by replacing computationally intensive simulation models with less costly surrogate models and by employing Sequential Optimization and reliability assessment (SORA) method. The SORA method uses a single-loop strategy with a series of cycles of deterministic optimization and reliability assessment. The deterministic optimization and reliability assessment is decoupled in each cycle. This leads to quick improvement of design from one cycle to other and increase in computational efficiency. This paper demonstrates the effectiveness of Sequential Optimization and Reliability Assessment (SORA) method when applied to designing a sheet metal flanging process. Surrogate models are used as less costly approximations to the computationally expensive Finite Element simulations

  6. Safety assessment of Department of Energy nuclear reactors

    International Nuclear Information System (INIS)

    1981-03-01

    One of the first tasks of the NFPQT Committee was to determine which DOE reactors would be assessed. The Committee determined that in view of the limited time available to conduct the assessment, 13 DOE reactors were of such size (physical, power or fission product inventory) to warrant review. This determination was approved by the Under Secretary. A decision was also made in the cases of three weapons material production reactors, C, K and P, to concentrate on the K reactor only, since all three are of the same basic design, have the same operating features, are all at the same site, and are all operated by the same contractor. The assessment was accomplished in the following ways: reviewing the results of assessments conducted by the DOE organizations with reactor safety responsibilities, which were undertaken in compliance with the request of the various program directors; reviewing selected documents that were requested by the Committee and assembled at DOE Headquarters; interviewing DOE Headquarters and Field Office personnel; and conducting on-site reviews of four reactors located at four different sites. The four reactors for on-site reviews were: Advanced Test Reactor (ATR); K Production Reactor; High Flux Beam Reactor (HFBR); and High Flux Isotope Reactor (HFIR). Specific findings and recommendations from the assessment are presented

  7. Unavailability Analysis of the Reactor Core Protection System using Reliability Block Diagram

    International Nuclear Information System (INIS)

    Shin, Hyun Kook; Kim, Sung Ho; Choi, Woong Suk; Kim, Jae Hack

    2006-01-01

    The reactor core of nuclear power plants needs to be monitored for the early detection of core abnormal conditions to protect plants from a severe accident. The core protection calculator system (CPCS) has been provided to calculate the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) based on measured parameters of reactor and coolant system. The original CPCS for OPR 1000 has been designed and implemented based on the concurrent 3205 computer system whose components are obsolete. The CPCS based on Westinghouse Common-Q system has recently been implemented for the Shin-Kori Nuclear Power Plant, Units 1 and 2(SKN 1 and 2). An R and D project has been launched to develop new core protection system called as RCOPS (Reactor Core Protection System) with the partnership of KOPEC and Doosan Heavy Industries and Construction Co. RCOPS is implemented on the HFC-6000 safety class programmable logic controller (PLC). In this paper, the reliability of RCOPS is analyzed using the reliability block diagram (RBD) method. The calculated results are compared with that of the CPCS for SKN 1 and 2

  8. Translation, reliability, and clinical utility of the Melbourne Assessment 2.

    Science.gov (United States)

    Gerber, Corinna N; Plebani, Anael; Labruyère, Rob

    2017-10-12

    The aims were to (i) provide a German translation of the Melbourne Assessment 2 (MA2), a quantitative test to measure unilateral upper limb function in children with neurological disabilities and (ii) to evaluate its reliability and aspects of clinical utility. After its translation into German and approval of the back translation by the original authors, the MA2 was performed and videotaped twice with 30 children with neuromotor disorders. For each participant, two raters scored the video of the first test for inter-rater reliability. To determine test-retest reliability, one rater additionally scored the video of the second test while the other rater repeated the scoring of the first video to evaluate intra-rater reliability. Time needed for rater training, test administration, and scoring was recorded. The four subscale scores showed excellent intra-, inter-rater, and test-retest reliability with intraclass correlation coefficients of 0.90-1.00 (95%-confidence intervals 0.78-1.00). Score items revealed substantial to almost perfect intra-rater reliability (weighted kappa k w  = 0.66-1.00) for the more affected side. Score item inter-rater and test-retest reliability of the same extremity were, with one exception, moderate to almost perfect (k w  = 0.42-0.97; k w  = 0.40-0.89). Furthermore, the MA2 was feasible and acceptable for patients and clinicians. The MA2 showed excellent subscale and moderate to almost perfect score item reliability. Implications for Rehabilitation There is a lack of high-quality studies about psychometric properties of upper limb measurement tools in the neuropediatric population. The Melbourne Assessment 2 is a promising tool for reliable measurement of unilateral upper limb movement quality in the neuropediatric population. The Melbourne Assessment 2 is acceptable and practicable to therapists and patients for routine use in clinical care.

  9. Insights from the interim reliability evaluation program pertinent to reactor safety issues

    International Nuclear Information System (INIS)

    Carlson, D.D.

    1983-01-01

    The Interim Reliability Evaluation Program (IREP) consisted of concurrent probabilistic analyses of four operating nuclear power plants. This paper presents and integrated view of the results of the analyses drawing insights pertinent to reactor safety. The importance to risk of accident sequences initiated by transients and small loss-of-coolant accidents was confirmed. Support systems were found to contribute significantly to the sets of dominant accident sequences, either due to single failures which could disable one or more mitigating systems or due to their initiating plant transients. Human errors in response to accidents also were important risk contributors. Consideration of operator recovery actions influences accident sequence frequency estimates, the list of accident sequences dominating core melt, and the set of dominant risk contributors. Accidents involving station blackout, reactor coolant pump seal leaks and ruptures, and loss-of-coolant accidents requiring manual initiation of coolant injection were found to be risk significant

  10. Is the Frontal Assessment Battery reliable in ALS patients?

    NARCIS (Netherlands)

    Raaphorst, J.; Beeldman, E.; Jaeger, B.; Schmand, B.A.; Berg, L.H. van den; Weikamp, J.G.; Schelhaas, H.J.; Visser, M. de; Haan, R.J. de

    2013-01-01

    The assessment of frontal functions in ALS patients is important because of the overlap with the behavioural variant of frontotemporal dementia (bvFTD). We investigated the applicability and reliability of the Frontal Assessment Battery (FAB) within a cohort of predominantly prevalent ALS patients.

  11. Photographic assessment of burn size and depth: reliability and validity

    NARCIS (Netherlands)

    Hop, M.; Moues, C.; Bogomolova, K.; Nieuwenhuis, M.; Oen, I.; Middelkoop, E.; Breederveld, R.; de Baar, M.

    2014-01-01

    Objective: The aim of this study was to examine the reliability and validity of using photographs of burns to assess both burn size and depth. Method: Fifty randomly selected photographs taken on day 0-1 post burn were assessed by seven burn experts and eight referring physicians. Inter-rater

  12. Valid and Reliable Science Content Assessments for Science Teachers

    Science.gov (United States)

    Tretter, Thomas R.; Brown, Sherri L.; Bush, William S.; Saderholm, Jon C.; Holmes, Vicki-Lynn

    2013-01-01

    Science teachers' content knowledge is an important influence on student learning, highlighting an ongoing need for programs, and assessments of those programs, designed to support teacher learning of science. Valid and reliable assessments of teacher science knowledge are needed for direct measurement of this crucial variable. This paper…

  13. Reliability and validity of subjective assessment of lumbar lordosis in ...

    African Journals Online (AJOL)

    Radiological assessment of lumbar lordotic curve aids in early diagnosis of conditions even before neurologic changes set in. Objective: To ascertain the level of reliability and validity of subjective assessment of lumbar lordosis in conventional radiography. Design: A blinded, repeated-measures diagnostic test was carried ...

  14. A computational Bayesian approach to dependency assessment in system reliability

    International Nuclear Information System (INIS)

    Yontay, Petek; Pan, Rong

    2016-01-01

    Due to the increasing complexity of engineered products, it is of great importance to develop a tool to assess reliability dependencies among components and systems under the uncertainty of system reliability structure. In this paper, a Bayesian network approach is proposed for evaluating the conditional probability of failure within a complex system, using a multilevel system configuration. Coupling with Bayesian inference, the posterior distributions of these conditional probabilities can be estimated by combining failure information and expert opinions at both system and component levels. Three data scenarios are considered in this study, and they demonstrate that, with the quantification of the stochastic relationship of reliability within a system, the dependency structure in system reliability can be gradually revealed by the data collected at different system levels. - Highlights: • A Bayesian network representation of system reliability is presented. • Bayesian inference methods for assessing dependencies in system reliability are developed. • Complete and incomplete data scenarios are discussed. • The proposed approach is able to integrate reliability information from multiple sources at multiple levels of the system.

  15. Sequential decision reliability concept and failure rate assessment

    International Nuclear Information System (INIS)

    Ciftcioglu, O.

    1990-11-01

    Conventionally, a reliability concept is considered together with both each basic unit and their integration in a complicated large scale system such as a nuclear power plant (NPP). Basically, as the plant's operational status is determined by the information obtained from various sensors, the plant's reliability and the risk assessment is closely related to the reliability of the sensory information and hence the sensor components. However, considering the relevant information-processing systems, e.g. fault detection processors, there exists a further question about the reliability of such systems, specifically the reliability of the systems' decision-based outcomes by means of which the further actions are performed. To this end, a general sequential decision reliability concept and the failure rate assessment methodology is introduced. The implications of the methodology are investigated and the importance of the decision reliability concept in system operation is demonstrated by means of sensory signals in real-time from the Borssele NPP in the Netherlands. (author). 21 refs.; 8 figs

  16. Reliability of sonographic assessment of tendinopathy in tennis elbow.

    Science.gov (United States)

    Poltawski, Leon; Ali, Syed; Jayaram, Vijay; Watson, Tim

    2012-01-01

    To assess the reliability and compute the minimum detectable change using sonographic scales to quantify the extent of pathology and hyperaemia in the common extensor tendon in people with tennis elbow. The lateral elbows of 19 people with tennis elbow were assessed sonographically twice, 1-2 weeks apart. Greyscale and power Doppler images were recorded for subsequent rating of abnormalities. Tendon thickening, hypoechogenicity, fibrillar disruption and calcification were each rated on four-point scales, and scores were summed to provide an overall rating of structural abnormality; hyperaemia was scored on a five point scale. Inter-rater reliability was established using the intraclass correlation coefficient (ICC) to compare scores assigned independently to the same set of images by a radiologist and a physiotherapist with training in musculoskeletal imaging. Test-retest reliability was assessed by comparing scores assigned by the physiotherapist to images recorded at the two sessions. The minimum detectable change (MDC) was calculated from the test-retest reliability data. ICC values for inter-rater reliability ranged from 0.35 (95% CI: 0.05, 0.60) for fibrillar disruption to 0.77 (0.55, 0.88) for overall greyscale score, and 0.89 (0.79, 0.95) for hyperaemia. Test-retest reliability ranged from 0.70 (0.48, 0.84) for tendon thickening to 0.82 (0.66, 0.90) for overall greyscale score and 0.86 (0.73, 0.93) for calcification. The MDC for the greyscale total score was 2.0/12 and for the hyperaemia score was 1.1/5. The sonographic scoring system used in this study may be used reliably to quantify tendon abnormalities and change over time. A relatively inexperienced imager can conduct the assessment and use the rating scales reliably.

  17. Design and reliability of a didactic inphographic rubric assessment

    OpenAIRE

    Guzmán-cedillo, Yunuen Ixchel

    2017-01-01

    The objective of this study is to describe design, validity process and reliability of a rubric assessment to evaluate didactic infographics quality. Participants were fifteen judges who participate in different moments of elaboration rubric process; it was made in three process phases: design, settings and reliability determination. Content validity was obtained by percentage agreement between 3 judges by component of the rubric; likewise a Krippendorff’s alpha were applied (a = .710) in pi...

  18. Reliability of histologic assessment in patients with eosinophilic oesophagitis.

    Science.gov (United States)

    Warners, M J; Ambarus, C A; Bredenoord, A J; Verheij, J; Lauwers, G Y; Walsh, J C; Katzka, D A; Nelson, S; van Viegen, T; Furuta, G T; Gupta, S K; Stitt, L; Zou, G; Parker, C E; Shackelton, L M; D Haens, G R; Sandborn, W J; Dellon, E S; Feagan, B G; Collins, M H; Jairath, V; Pai, R K

    2018-04-01

    The validity of the eosinophilic oesophagitis (EoE) histologic scoring system (EoEHSS) has been demonstrated, but only preliminary reliability data exist. Formally assess the reliability of the EoEHSS and additional histologic features. Four expert gastrointestinal pathologists independently reviewed slides from adult patients with EoE (N = 45) twice, in random order, using standardised training materials and scoring conventions for the EoEHSS and additional histologic features agreed upon during a modified Delphi process. Intra- and inter-rater reliability for scoring the EoEHSS, a visual analogue scale (VAS) of overall histopathologic disease severity, and additional histologic features were assessed using intra-class correlation coefficients (ICCs). Almost perfect intra-rater reliability was observed for the composite EoEHSS scores and the VAS. Inter-rater reliability was also almost perfect for the composite EoEHSS scores and substantial for the VAS. Of the EoEHSS items, eosinophilic inflammation was associated with the highest ICC estimates and consistent with almost perfect intra- and inter-rater reliability. With the exception of dyskeratotic epithelial cells and surface epithelial alteration, ICC estimates for the remaining EoEHSS items were above the benchmarks for substantial intra-rater, and moderate inter-rater reliability. Estimation of peak eosinophil count and number of lamina propria eosinophils were associated with the highest ICC estimates among the exploratory items. The composite EoEHSS and most component items are associated with substantial reliability when assessed by central pathologists. Future studies should assess responsiveness of the score to change after a therapeutic intervention to facilitate its use in clinical trials. © 2018 John Wiley & Sons Ltd.

  19. Soil slurry reactors for the assessment of contaminant biodegradation

    Science.gov (United States)

    Toscano, G.; Colarieti, M. L.; Greco, G.

    2012-04-01

    Slurry reactors are frequently used in the assessment of feasibility of biodegradation in natural soil systems. The rate of contaminant removal is usually quantified by zero- or first-order kinetics decay constants. The significance of such constants for the evaluation of removal rate in the field could be questioned because the slurry reactor is a water-saturated, well-stirred system without resemblance with an unsaturated fixed bed of soil. Nevertheless, a kinetic study with soil slurry reactors can still be useful by means of only slightly more sophisticated kinetic models than zero-/first-order decay. The use of kinetic models taking into account the role of degrading biomass, even in the absence of reliable experimental methods for its quantification, provides further insight into the effect of nutrient additions. A real acceleration of biodegradation processes is obtained only when the degrading biomass is in the growth condition. The apparent change in contaminant removal course can be useful to diagnose biomass growth without direct biomass measurement. Even though molecular biology techniques are effective to assess the presence of potentially degrading microorganism in a "viable-but-nonculturable" state, the attainment of conditions for growth is still important to the development of enhanced remediation techniques. The methodology is illustrated with reference to data gathered for two test sites, Oslo airport Gardermoen in Norway (continuous contamination by aircraft deicing fluids) and the Trecate site in Italy (aged contamination by crude oil spill). This research is part of SoilCAM project (Soil Contamination, Advanced integrated characterisation and time-lapse Monitoring 2008-2012, EU-FP7).

  20. Generic magnetic fusion reactor cost assessment

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    The Fusion Energy Division of the Oak Ridge National Laboratory discusses ''generic'' magnetic fusion reactors. The author comments on DT burning magnetic fusion reactor models being possibly operational in the 21st century. Representative parameters from D-T reactor studies are given, as well as a shematic diagram of a generic fusion reactor. Values are given for winding pack current density for existing and future superconducting coils. Topics included are the variation of the cost of electricity (COE), the dependence of the COE on the net electric power of the reactor, and COE formula definitions

  1. Design of power auto-regulating system's high reliability controller for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    An Zhencai; Liu Longzhi; Chen Yuan

    1996-01-01

    The paper mainly introduces power auto-regulating system's high reliability controller for 200 MW Nuclear Heating Reactor. The controller is implemented with excellent performance 16 bit single chip microcomputer 8097. Master controller and 10 digit samplers are blocked. Each and every block's hardware is identical. These blocks communicate each other through 8 bit BUS and operate synchronously by united clock and reset signal and are designed with three redundancies. The identity comparison principle through two-out-of three is also introduced. The test proves that designing scheme is feasible

  2. Improvement in the reliability of shells for light water reactors by manufacturing from hollow ingots

    International Nuclear Information System (INIS)

    Bocquet, P.; Blondeau, R.; Poitrault, I.; Badeau, J.P.; Dumont, R.

    1989-01-01

    The example of forging shells for PWR type reactors is proposed to show how the choice of the manufacturing process may be of prime importance for the component integrity by the reduction of the detrimental effects of segregations. The forging shells (20MnMoNi55) manufactured from hollow ingot are free of any segregation in the critical area located just at the internal surface and sub-surface. Manufacturing problems associated to these segregations in shells issued from conventional ingots, in particular welding difficulties for cladding, have been reduced or eleminated. The reliability of these components present an improved resistance to irradiation embrittlement. (DG)

  3. Reliable and valid assessment of Lichtenstein hernia repair skills

    DEFF Research Database (Denmark)

    Carlsen, C G; Lindorff Larsen, Karen; Funch-Jensen, P

    2014-01-01

    PURPOSE: Lichtenstein hernia repair is a common surgical procedure and one of the first procedures performed by a surgical trainee. However, formal assessment tools developed for this procedure are few and sparsely validated. The aim of this study was to determine the reliability and validity...... of an assessment tool designed to measure surgical skills in Lichtenstein hernia repair. METHODS: Key issues were identified through a focus group interview. On this basis, an assessment tool with eight items was designed. Ten surgeons and surgical trainees were video recorded while performing Lichtenstein hernia...... a significant difference between the three groups which indicates construct validity, p skills can be assessed blindly by a single rater in a reliable and valid fashion with the new procedure-specific assessment tool. We recommend this tool for future assessment...

  4. Reliability and Probabilistic Risk Assessment - How They Play Together

    Science.gov (United States)

    Safie, Fayssal M.; Stutts, Richard G.; Zhaofeng, Huang

    2015-01-01

    PRA methodology is one of the probabilistic analysis methods that NASA brought from the nuclear industry to assess the risk of LOM, LOV and LOC for launch vehicles. PRA is a system scenario based risk assessment that uses a combination of fault trees, event trees, event sequence diagrams, and probability and statistical data to analyze the risk of a system, a process, or an activity. It is a process designed to answer three basic questions: What can go wrong? How likely is it? What is the severity of the degradation? Since 1986, NASA, along with industry partners, has conducted a number of PRA studies to predict the overall launch vehicles risks. Planning Research Corporation conducted the first of these studies in 1988. In 1995, Science Applications International Corporation (SAIC) conducted a comprehensive PRA study. In July 1996, NASA conducted a two-year study (October 1996 - September 1998) to develop a model that provided the overall Space Shuttle risk and estimates of risk changes due to proposed Space Shuttle upgrades. After the Columbia accident, NASA conducted a PRA on the Shuttle External Tank (ET) foam. This study was the most focused and extensive risk assessment that NASA has conducted in recent years. It used a dynamic, physics-based, integrated system analysis approach to understand the integrated system risk due to ET foam loss in flight. Most recently, a PRA for Ares I launch vehicle has been performed in support of the Constellation program. Reliability, on the other hand, addresses the loss of functions. In a broader sense, reliability engineering is a discipline that involves the application of engineering principles to the design and processing of products, both hardware and software, for meeting product reliability requirements or goals. It is a very broad design-support discipline. It has important interfaces with many other engineering disciplines. Reliability as a figure of merit (i.e. the metric) is the probability that an item will

  5. Assessment of structural reliability of precast concrete buildings

    Directory of Open Access Journals (Sweden)

    Koyankin Alexandr

    2018-01-01

    Full Text Available Precast housing construction is currently being under rapid development, however, reliability of building structures made from precast reinforced concrete cannot be assessed rationally due to insufficient research data on that subject. In this regard, experimental and numerical studies were conducted to assess structural reliability of precast buildings as described in the given paper. Experimental studies of full-scale and model samples were conducted; numerical studies were held based on finite element models using “Lira” software. The objects under study included fragment of flooring of a building under construction, full-size fragment of flooring, full-scale models of precast cross-beams-to-columns joints and joints between hollow-core floor slabs and precast and cast-in-place cross-beams. Conducted research enabled to perform an objective assessment of structural reliability of precast buildings.

  6. A new measurement of workload in Web application reliability assessment

    Directory of Open Access Journals (Sweden)

    CUI Xia

    2015-02-01

    Full Text Available Web application has been popular in various fields of social life.It becomes more and more important to study the reliability of Web application.In this paper the definition of Web application failure is firstly brought out,and then the definition of Web application reliability.By analyzing data in the IIS server logs and selecting corresponding usage and information delivery failure data,the paper study the feasibility of Web application reliability assessment from the perspective of Web software system based on IIS server logs.Because the usage for a Web site often has certain regularity,a new measurement of workload in Web application reliability assessment is raised.In this method,the unit is removed by weighted average technique;and the weights are assessed by setting objective function and optimization.Finally an experiment was raised for validation.The experiment result shows the assessment of Web application reliability base on the new workload is better.

  7. Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

    2014-12-15

    The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

  8. ASSESSING AND COMBINING RELIABILITY OF PROTEIN INTERACTION SOURCES

    Science.gov (United States)

    LEACH, SONIA; GABOW, AARON; HUNTER, LAWRENCE; GOLDBERG, DEBRA S.

    2008-01-01

    Integrating diverse sources of interaction information to create protein networks requires strategies sensitive to differences in accuracy and coverage of each source. Previous integration approaches calculate reliabilities of protein interaction information sources based on congruity to a designated ‘gold standard.’ In this paper, we provide a comparison of the two most popular existing approaches and propose a novel alternative for assessing reliabilities which does not require a gold standard. We identify a new method for combining the resultant reliabilities and compare it against an existing method. Further, we propose an extrinsic approach to evaluation of reliability estimates, considering their influence on the downstream tasks of inferring protein function and learning regulatory networks from expression data. Results using this evaluation method show 1) our method for reliability estimation is an attractive alternative to those requiring a gold standard and 2) the new method for combining reliabilities is less sensitive to noise in reliability assignments than the similar existing technique. PMID:17990508

  9. Assessment of the reliability of ultrasonic inspection methods

    International Nuclear Information System (INIS)

    Haines, N.F.; Langston, D.B.; Green, A.J.; Wilson, R.

    1982-01-01

    The reliability of NDT techniques has remained an open question for many years. A reliable technique may be defined as one that, when rigorously applied by a number of inspection teams, consistently finds then correctly sizes all defects of concern. In this paper we report an assessment of the reliability of defect detection by manual ultrasonic methods applied to the inspection of thick section pressure vessel weldments. Initially we consider the available data relating to the inherent physical capabilities of ultrasonic techniques to detect cracks in weldment and then, independently, we assess the likely variability in team to team performance when several teams are asked to follow the same specified test procedure. The two aspects of 'capability' and 'variability' are brought together to provide quantitative estimates of the overall reliability of ultrasonic inspection of thick section pressure vessel weldments based on currently existing data. The final section of the paper considers current research programmes on reliability and presents a view on how these will help to further improve NDT reliability. (author)

  10. Reliable and valid assessment of performance in thoracoscopy

    DEFF Research Database (Denmark)

    Konge, Lars; Lehnert, Per; Hansen, Henrik Jessen

    2012-01-01

    BACKGROUND: As we move toward competency-based education in medicine, we have lagged in developing competency-based evaluation methods. In the era of minimally invasive surgery, there is a need for a reliable and valid tool dedicated to measure competence in video-assisted thoracoscopic surgery....... The purpose of this study is to create such an assessment tool, and to explore its reliability and validity. METHODS: An expert group of physicians created an assessment tool consisting of 10 items rated on a five-point rating scale. The following factors were included: economy and confidence of movement...

  11. Reliability assessment for safety critical systems by statistical random testing

    International Nuclear Information System (INIS)

    Mills, S.E.

    1995-11-01

    In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs

  12. Reliability assessment for safety critical systems by statistical random testing

    Energy Technology Data Exchange (ETDEWEB)

    Mills, S E [Carleton Univ., Ottawa, ON (Canada). Statistical Consulting Centre

    1995-11-01

    In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs.

  13. Light-water reactors reference system classification for the European reliability data system (ERDS)

    International Nuclear Information System (INIS)

    Melis, M.; Mancini, G.

    1982-01-01

    The reference system classification represents a basic stage in the organization of the European reliability data system (ERDS) for light-water reactors, a project actually in development at the Joint Research Centre, Ispra. This project is concerned with operational reliability data collection from the various ''national'' data banks, and centralization in a European reliability data system, so improving the significance of the resulting reliability evaluations. In the framework of the ERDS project, the reference system classification provides a LWR functional break-down and represents a plant-unique identification in the process of homogenization of event-data coming from the various ''national'' organizations. The report, after a brief description of the main objectives of the ERDS project, reviews the criteria followed in the elaboration of the reference system classification; then the detailed classification is presented. The nuclear power station is subdivided in about 180 systems. To each system a sheet is associated, containing: a comprehensive description of system-functions and boundaries; a descritpion of the plant operating mode, linked to the various system functions; a list of the main interface system; and finally, a list of the main components, including type and safety classification

  14. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  15. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  16. Reliability assessment of complex electromechanical systems under epistemic uncertainty

    International Nuclear Information System (INIS)

    Mi, Jinhua; Li, Yan-Feng; Yang, Yuan-Jian; Peng, Weiwen; Huang, Hong-Zhong

    2016-01-01

    The appearance of macro-engineering and mega-project have led to the increasing complexity of modern electromechanical systems (EMSs). The complexity of the system structure and failure mechanism makes it more difficult for reliability assessment of these systems. Uncertainty, dynamic and nonlinearity characteristics always exist in engineering systems due to the complexity introduced by the changing environments, lack of data and random interference. This paper presents a comprehensive study on the reliability assessment of complex systems. In view of the dynamic characteristics within the system, it makes use of the advantages of the dynamic fault tree (DFT) for characterizing system behaviors. The lifetime of system units can be expressed as bounded closed intervals by incorporating field failures, test data and design expertize. Then the coefficient of variation (COV) method is employed to estimate the parameters of life distributions. An extended probability-box (P-Box) is proposed to convey the present of epistemic uncertainty induced by the incomplete information about the data. By mapping the DFT into an equivalent Bayesian network (BN), relevant reliability parameters and indexes have been calculated. Furthermore, the Monte Carlo (MC) simulation method is utilized to compute the DFT model with consideration of system replacement policy. The results show that this integrated approach is more flexible and effective for assessing the reliability of complex dynamic systems. - Highlights: • A comprehensive study on the reliability assessment of complex system is presented. • An extended probability-box is proposed to convey the present of epistemic uncertainty. • The dynamic fault tree model is built. • Bayesian network and Monte Carlo simulation methods are used. • The reliability assessment of a complex electromechanical system is performed.

  17. Reliable and valid assessment of Lichtenstein hernia repair skills.

    Science.gov (United States)

    Carlsen, C G; Lindorff-Larsen, K; Funch-Jensen, P; Lund, L; Charles, P; Konge, L

    2014-08-01

    Lichtenstein hernia repair is a common surgical procedure and one of the first procedures performed by a surgical trainee. However, formal assessment tools developed for this procedure are few and sparsely validated. The aim of this study was to determine the reliability and validity of an assessment tool designed to measure surgical skills in Lichtenstein hernia repair. Key issues were identified through a focus group interview. On this basis, an assessment tool with eight items was designed. Ten surgeons and surgical trainees were video recorded while performing Lichtenstein hernia repair, (four experts, three intermediates, and three novices). The videos were blindly and individually assessed by three raters (surgical consultants) using the assessment tool. Based on these assessments, validity and reliability were explored. The internal consistency of the items was high (Cronbach's alpha = 0.97). The inter-rater reliability was very good with an intra-class correlation coefficient (ICC) = 0.93. Generalizability analysis showed a coefficient above 0.8 even with one rater. The coefficient improved to 0.92 if three raters were used. One-way analysis of variance found a significant difference between the three groups which indicates construct validity, p fashion with the new procedure-specific assessment tool. We recommend this tool for future assessment of trainees performing Lichtenstein hernia repair to ensure that the objectives of competency-based surgical training are met.

  18. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-11-01

    Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

  19. Time-variant reliability assessment through equivalent stochastic process transformation

    International Nuclear Information System (INIS)

    Wang, Zequn; Chen, Wei

    2016-01-01

    Time-variant reliability measures the probability that an engineering system successfully performs intended functions over a certain period of time under various sources of uncertainty. In practice, it is computationally prohibitive to propagate uncertainty in time-variant reliability assessment based on expensive or complex numerical models. This paper presents an equivalent stochastic process transformation approach for cost-effective prediction of reliability deterioration over the life cycle of an engineering system. To reduce the high dimensionality, a time-independent reliability model is developed by translating random processes and time parameters into random parameters in order to equivalently cover all potential failures that may occur during the time interval of interest. With the time-independent reliability model, an instantaneous failure surface is attained by using a Kriging-based surrogate model to identify all potential failure events. To enhance the efficacy of failure surface identification, a maximum confidence enhancement method is utilized to update the Kriging model sequentially. Then, the time-variant reliability is approximated using Monte Carlo simulations of the Kriging model where system failures over a time interval are predicted by the instantaneous failure surface. The results of two case studies demonstrate that the proposed approach is able to accurately predict the time evolution of system reliability while requiring much less computational efforts compared with the existing analytical approach. - Highlights: • Developed a new approach for time-variant reliability analysis. • Proposed a novel stochastic process transformation procedure to reduce the dimensionality. • Employed Kriging models with confidence-based adaptive sampling scheme to enhance computational efficiency. • The approach is effective for handling random process in time-variant reliability analysis. • Two case studies are used to demonstrate the efficacy

  20. Researching and improving the reliability of reactor protection system of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zuyue; Sheng Jiannan

    1997-01-01

    Due to the original design defects of the Reactor Protection System (RPS) of Qinshan Nuclear Power Plant, this system has brought about a number of reactor shutdown accidents and Engineered Safety Features (ESF) mis-activation events which have seriously endangered safe and steady operation of the nuclear power plant. So over three years have been spent on research on the reform of the original design on the premise that the general wiring of the system should remain the same and that the system size should also remain small to be contained in the original cabinets. The following improvements were made: (1) Increase the system's anti-disturbance capability. The system's zero volt bus floating designs were modified to surmount the disturbance resulting from the bad isolation performance of impedance-isolated amplifier; Double grounds have been added to logical modules to surmount the disturbance resulting from zero volt floating bus during the replacement of single module with two connectors; The opto-coupling circuit in its oscillation input stage of Engineered Safety Features have been improved to increase its reliability. (2)Modify to output activation part of the system. The new type of output relays were selected and the relay activation circuits were redesigned in which switcher activation mode is used instead of amplifier activation mode so as to increase the reliability of relay operation and reduce the power consumption; CMOS buffer gates in the input and output stage of the circuit were used to match TTL circuits to CMOS circuits of the system

  1. Savannah River Site K-Reactor Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

  2. Reliable and Valid Assessment of Point-of-care Ultrasonography

    DEFF Research Database (Denmark)

    Todsen, Tobias; Tolsgaard, Martin Grønnebæk; Olsen, Beth Härstedt

    2015-01-01

    physicians' OSAUS scores with diagnostic accuracy. RESULTS: The generalizability coefficient was high (0.81) and a D-study demonstrated that 1 assessor and 5 cases would result in similar reliability. The construct validity of the OSAUS scale was supported by a significant difference in the mean scores......OBJECTIVE: To explore the reliability and validity of the Objective Structured Assessment of Ultrasound Skills (OSAUS) scale for point-of-care ultrasonography (POC US) performance. BACKGROUND: POC US is increasingly used by clinicians and is an essential part of the management of acute surgical...... conditions. However, the quality of performance is highly operator-dependent. Therefore, reliable and valid assessment of trainees' ultrasonography competence is needed to ensure patient safety. METHODS: Twenty-four physicians, representing novices, intermediates, and experts in POC US, scanned 4 different...

  3. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

  4. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  5. Windfarm Generation Assessment for ReliabilityAnalysis of Power Systems

    DEFF Research Database (Denmark)

    Negra, Nicola Barberis; Holmstrøm, Ole; Bak-Jensen, Birgitte

    2007-01-01

    Due to the fast development of wind generation in the past ten years, increasing interest has been paid to techniques for assessing different aspects of power systems with a large amount of installed wind generation. One of these aspects concerns power system reliability. Windfarm modelling plays...

  6. Windfarm generation assessment for reliability analysis of power systems

    DEFF Research Database (Denmark)

    Negra, N.B.; Holmstrøm, O.; Bak-Jensen, B.

    2007-01-01

    Due to the fast development of wind generation in the past ten years, increasing interest has been paid to techniques for assessing different aspects of power systems with a large amount of installed wind generation. One of these aspects concerns power system reliability. Windfarm modelling plays...

  7. Methodology for reliability, economic and environmental assessment of wave energy

    International Nuclear Information System (INIS)

    Thorpe, T.W.; Muirhead, S.

    1994-01-01

    As part of the Preliminary Actions in Wave Energy R and D for DG XII's Joule programme, methodologies were developed to facilitate assessment of the reliability, economics and environmental impact of wave energy. This paper outlines these methodologies, their limitations and areas requiring further R and D. (author)

  8. Reliable and Valid Assessment of Clinical Bronchoscopy Performance

    DEFF Research Database (Denmark)

    Konge, Lars; Larsen, Klaus Richter; Clementsen, Paul

    2012-01-01

    : The interrater reliability was high, with Cronbach's a = 0.86. Assessment of 3 bronchoscopies by a single rater had a generalizability coefficient of 0.84. The correlation between experience and performance was good (Pearson correlation = 0.76). There were significant differences between the groups for all...

  9. Assessment of the Reliability of Concrete Slab Bridges

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle; Jensen, F. M.; Middleton, C. R.

    This paper is based on research performed for the Highways Agency, London, UK under the project DPU/9/44 "Revision of Bridge Assessment Rules Based on Whole Life Performance: Concrete Bridges". It contains details of a methodology which can be used to generate Whole Life (WL) reliability profiles....

  10. Assessment of reliability of Greulich and Pyle (gp) method for ...

    African Journals Online (AJOL)

    Background: Greulich and Pyle standards are the most widely used age estimation standards all over the world. The applicability of the Greulich and Pyle standards to populations which differ from their reference population is often questioned. This study aimed to assess the reliability of Greulich and Pyle (GP) method for ...

  11. Reliability of scored patient generated subjective global assessment ...

    African Journals Online (AJOL)

    Objective: Establish the reliability of the scored Patient Generated-Subjective Global Assessment (PG-SGA) in determining nutritional status among Antiretroviral Therapy (ART) naive HIV-infected adults. Methods: A descriptive, cross sectional study among outpatient medical clinics, in The AIDS Support Organization ...

  12. Qualification Users' Perceptions and Experiences of Assessment Reliability

    Science.gov (United States)

    Chamberlain, Suzanne

    2013-01-01

    This paper presents the findings of a study designed to explore qualification users' perceptions and experiences of reliability in the context of national assessment outcomes in England. The study consisted of 17 focus groups conducted across six sectors of qualification users: students, teachers, trainee teachers, job-seekers, employers and…

  13. Reliability of medical audit in quality assessment of medical care

    Directory of Open Access Journals (Sweden)

    Camacho Luiz Antonio Bastos

    1996-01-01

    Full Text Available Medical audit of hospital records has been a major component of quality of care assessment, although physician judgment is known to have low reliability. We estimated interrater agreement of quality assessment in a sample of patients with cardiac conditions admitted to an American teaching hospital. Physician-reviewers used structured review methods designed to improve quality assessment based on judgment. Chance-corrected agreement for the items considered more relevant to process and outcome of care ranged from low to moderate (0.2 to 0.6, depending on the review item and the principal diagnoses and procedures the patients underwent. Results from several studies seem to converge on this point. Comparisons among different settings should be made with caution, given the sensitivity of agreement measurements to prevalence rates. Reliability of review methods in their current stage could be improved by combining the assessment of two or more reviewers, and by emphasizing outcome-oriented events.

  14. Risk assessment and reliability for low level radioactive waste disposal

    International Nuclear Information System (INIS)

    Gregory, P.O.; Jones, G.A.

    1986-01-01

    The reliability of critical design features at low-level radioactive waste disposal facilities is a major concern in the licensing of these structures. To date, no systematic methodology has been adopted to evaluate the geotechnical reliability of Uranium Mill Tailings Remedial Action (UMTRA) disposal facilities currently being designed and/or constructed. This paper discusses and critiques the deterministic methods currently used to evaluate UMTRA reliability. Because deterministic methods may not be applicable in some cases because of the unusually long design life of UMTRA facilities, it is proposed that a probabilistic risk assessment-based methodology be used as a secondary method to aid in the evaluating of geotechnical reliability of critical items. Similar methodologies have proven successful in evaluating the reliability of a variety of conventional earth structures. In this paper, an ''acceptable'' level of risk for UMTRA facilities is developed, an evaluation method is presented, and two example applications of the proposed methodology are provided for a generic UMTRA disposal facility. The proposed technique is shown to be a simple method which might be used to aid in reliability evaluations on a selective basis. Finally, other possible applications and the limitations of the proposed methodology are discussed

  15. Reactor assessments of advanced bumpy torus configurations

    International Nuclear Information System (INIS)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1984-02-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past

  16. Observer reliability in assessing placental maturity by histology.

    OpenAIRE

    Khong, T Y; Staples, A; Bendon, R W; Chambers, H M; Gould, S J; Knowles, S; Shen-Schwarz, S

    1995-01-01

    AIMS--To evaluate the ability of five experienced perinatal pathologists to assess placental maturity reliably by histology. METHODS--Twenty four haematoxylin and eosin slides, six each from placentas of 27, 31, 35, and 39 weeks' gestation, were circulated to five pathologists on three separate occasions. The slides were labelled with the correct or incorrect gestational ages. RESULTS--The mean absolute error over all 360 readings was 2.72 weeks. Only 54% of the slides were assessed within tw...

  17. Reliability of the Hazelbaker Assessment Tool for Lingual Frenulum Function

    Directory of Open Access Journals (Sweden)

    James Jennifer P

    2006-03-01

    Full Text Available Abstract Background About 3% of infants are born with a tongue-tie which may lead to breastfeeding problems such as ineffective latch, painful attachment or poor weight gain. The Hazelbaker Assessment Tool for Lingual Frenulum Function (HATLFF has been developed to give a quantitative assessment of the tongue-tie and recommendation about frenotomy (release of the frenulum. The aim of this study was to assess the inter-rater reliability of the HATLFF. Methods Fifty-eight infants referred to the Breastfeeding Education and Support Services (BESS at The Royal Women's Hospital for assessment of tongue-tie and 25 control infants were assessed by two clinicians independently. Results The Appearance items received kappas between about 0.4 to 0.6, which represents "moderate" reliability. The first three Function items (lateralization, lift and extension of tongue had kappa values over 0.65 which indicates "substantial" agreement. The four Function items relating to infant sucking (spread, cupping, peristalsis and snapback received low kappa values with insignificant p values. There was 96% agreement between the two assessors on the recommendation for frenotomy (kappa 0.92, excellent agreement. The study found that the Function Score can be more simply assessed using only the first three function items (ie not scoring the sucking items, with a cut-off of ≤4 for recommendation of frenotomy. Conclusion We found that the HATLFF has a high reliability in a study of infants with tongue-tie and control infants

  18. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

    2003-04-01

    The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

  19. Environmental impact assessment around TRIGA research reactor

    International Nuclear Information System (INIS)

    Lee, Jeong Ho; Lee, Hyun Duk; Lee, Young Bok; Cheong, Kyu Hoi; Ahn, Jong Sung; Kim, Kug Chan; You, Byung Sun; Kim, Byung Woo; Kim, Sang Bok; Han Moon Hee

    1985-01-01

    Population distribution, atmospheric change, X/Q, characteristics of terrestrial ecosystem around Seoul site were surveyed. Environmental radiation and radioactivities such as grossα, grossβ, Cs-137, Sr-90 and H-3 of various environmental samples were analyzed. The values of environmental radiation dose tended to increase gradually in the light of the recent five years' results of environmental radiation monitoring around the nuclear power plants from 1980 to 1984, however, the changes were not significant. In addition, continuous assessment of environmental radiation monitoring on the roofs of main building and life science building at KAERI showed that the environmental radiation dose tended to increase a little during the night time. Judging from the above results, it is concluded that environmental contamination level by radioactive materials could be ignored in the case of radioisotope production or experiment using radioisotopes except the release of gaseous radioactive materials such as Ar-41 of short half life by the operation of nuclear reactor. (Author)

  20. Risk-assessment techniques and the reactor licensing process

    International Nuclear Information System (INIS)

    Levine, S.

    1979-01-01

    A brief description of the Reactor Safety Study (WASH-1400), concentrating on the engineering aspects of the contribution to reactor accident risks is followed by some comments on how we have applied the insights and techniques developed in this study to prepare a program to improve the safety of nuclear power plants. Some new work we are just beginning on the application of risk-assessment techniques to stablize the reactor licensing process is also discussed

  1. Coupling finite elements and reliability methods - application to safety evaluation of pressurized water reactor vessels

    International Nuclear Information System (INIS)

    Pitner, P.; Venturini, V.

    1995-02-01

    When reliability studies are extended form deterministic calculations in mechanics, it is necessary to take into account input parameters variabilities which are linked to the different sources of uncertainty. Integrals must then be calculated to evaluate the failure risk. This can be performed either by simulation methods, or by approximations ones (FORM/SORM). Model in mechanics often require to perform calculation codes. These ones must then be coupled with the reliability calculations. Theses codes can involve large calculation times when they are invoked numerous times during simulations sequences or in complex iterative procedures. Response surface method gives an approximation of the real response from a reduced number of points for which the finite element code is run. Thus, when it is combined with FORM/SORM methods, a coupling can be carried out which gives results in a reasonable calculation time. An application of response surface method to mechanics reliability coupling for a mechanical model which calls for a finite element code is presented. It corresponds to a probabilistic fracture mechanics study of a pressurized water reactor vessel. (authors). 5 refs., 3 figs

  2. Reliability Centered Maintenance (RCM) Methodology and Application to the Shutdown Cooling System for APR-1400 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faragalla, Mohamed M.; Emmanuel, Efenji; Alhammadi, Ibrahim; Awwal, Arigi M.; Lee, Yong Kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    Shutdown Cooling System (SCS) is a safety-related system that is used in conjunction with the Main Steam and Main or Auxiliary Feedwater Systems to reduce the temperature of the Reactor Coolant System (RCS) in post shutdown periods from the hot shutdown operating temperature to the refueling temperature. In this paper RCM methodology is applied to (SCS). RCM analysis is performed based on evaluation of Failure Modes Effects and Criticality Analysis (FME and CA) on the component, system and plant. The Logic Tree Analysis (LTA) is used to determine the optimum maintenance tasks. The main objectives of RCM is the safety, preserve the System function, the cost-effective maintenance of the plant components and increase the reliability and availability value. The RCM methodology is useful for improving the equipment reliability by strengthening the management of equipment condition, and leads to a significant decrease in the number of periodical maintenance, extended maintenance cycle, longer useful life of equipment, and decrease in overall maintenance cost. It also focuses on the safety of the system by assigning criticality index to the various components and further selecting maintenance activities based on the risk of failure involved. Therefore, it can be said that RCM introduces a maintenance plan designed for maximum safety in an economical manner and making the system more reliable. For the SCP, increasing the number of condition monitoring tasks will improve the availability of the SCP. It is recommended to reduce the number of periodic maintenance activities.

  3. Nonspecialist Raters Can Provide Reliable Assessments of Procedural Skills

    DEFF Research Database (Denmark)

    Mahmood, Oria; Dagnæs, Julia; Bube, Sarah

    2018-01-01

    was significant (p Pearson's correlation of 0.77 for the nonspecialists and 0.75 for the specialists. The test-retest reliability showed the biggest difference between the 2 groups, 0.59 and 0.38 for the nonspecialist raters and the specialist raters, respectively (p ... was chosen as it is a simple procedural skill that is crucial to master in a resident urology program. RESULTS: The internal consistency of assessments was high, Cronbach's α = 0.93 and 0.95 for nonspecialist and specialist raters, respectively (p correlations). The interrater reliability...

  4. Quantitative reliability assessment for safety critical system software

    International Nuclear Information System (INIS)

    Chung, Dae Won; Kwon, Soon Man

    2005-01-01

    An essential issue in the replacement of the old analogue I and C to computer-based digital systems in nuclear power plants is the quantitative software reliability assessment. Software reliability models have been successfully applied to many industrial applications, but have the unfortunate drawback of requiring data from which one can formulate a model. Software which is developed for safety critical applications is frequently unable to produce such data for at least two reasons. First, the software is frequently one-of-a-kind, and second, it rarely fails. Safety critical software is normally expected to pass every unit test producing precious little failure data. The basic premise of the rare events approach is that well-tested software does not fail under normal routine and input signals, which means that failures must be triggered by unusual input data and computer states. The failure data found under the reasonable testing cases and testing time for these conditions should be considered for the quantitative reliability assessment. We will present the quantitative reliability assessment methodology of safety critical software for rare failure cases in this paper

  5. On the reliability of steam generator performance at nuclear power plants with WWER type reactors

    International Nuclear Information System (INIS)

    Styrikovich, M.A.; Margulova, T.Kh.

    1974-01-01

    The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

  6. High reliability flow system - an assessment of pump reliability and optimisation of the number of pumps

    International Nuclear Information System (INIS)

    Butterfield, J.M.

    1981-01-01

    A system is considered where a number of pumps operate in parallel. Normally, all pumps operate, driven by main motors fed from the grid. Each pump has a pony motor fed from an individual battery supply. Each pony motor is normally running, but not engaged to the pump shaft. On demand, e.g. failure of grid supplies, each pony motor is designed to clutch-in automatically when the pump speed falls to a specified value. The probability of all the pony motors failing to clutch-in on demand must be demonstrated with 95% confidence to be less than 10 -8 per demand. This assessment considers how the required reliability of pony motor drives might be demonstrated in practice and the implications on choice of the number of pumps at the design stage. The assessment recognises that not only must the system prove to be extremely reliable, but that demonstration that reliability is adequate must be done during plant commissioning, with practical limits on the amount of testing performed. It is concluded that a minimum of eight pony motors should be provided, eight pumps each with one pony motor (preferred) or five pumps each with two independent pony motors. A minimum of two diverse pony motor systems should be provided. (author)

  7. Reliability Assessment of Transformerless PV Inverters considering Mission Profiles

    Directory of Open Access Journals (Sweden)

    Yongheng Yang

    2015-01-01

    Full Text Available Due to the small volume and high efficiency, transformerless inverters have gained much popularity in grid-connected PV applications, where minimizing leakage current injection is mandatory. This can be achieved by either modifying the modulation schemes or adding extra power switching devices, resulting in an uneven distribution of the power losses on the switching devices. Consequently, the device thermal loading is redistributed and thus may alter the entire inverter reliability performance, especially under a long-term operation. In this consideration, this paper assesses the device reliability of three transformerless inverters under a yearly mission profile (i.e., solar irradiance and ambient temperature. The mission profile is translated to device thermal loading, which is used for lifetime prediction. Comparison results reveal the lifetime mismatches among the power switching devices operating under the same condition, which offers new thoughts for a robust design and a reliable operation of grid-connected transformerless PV inverters with high efficiency.

  8. Reliability-based assessment of polyethylene pipe creep lifetime

    International Nuclear Information System (INIS)

    Khelif, Rabia; Chateauneuf, Alaa; Chaoui, Kamel

    2007-01-01

    Lifetime management of underground pipelines is mandatory for safe hydrocarbon transmission and distribution systems. The use of high-density polyethylene tubes subjected to internal pressure, external loading and environmental variations requires a reliability study in order to define the service limits and the optimal operating conditions. In service, the time-dependent phenomena, especially creep, take place during the pipe lifetime, leading to significant strength reduction. In this work, the reliability-based assessment of pipe lifetime models is carried out, in order to propose a probabilistic methodology for lifetime model selection and to determine the pipe safety levels as well as the most important parameters for pipeline reliability. This study is enhanced by parametric analysis on pipe configuration, gas pressure and operating temperature

  9. Reliability-based assessment of polyethylene pipe creep lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Khelif, Rabia [LaMI-UBP and IFMA, Campus de Clermont-Fd, Les Cezeaux, BP 265, 63175 Aubiere Cedex (France); LR3MI, Departement de Genie Mecanique, Universite Badji Mokhtar, BP 12, Annaba 23000 (Algeria)], E-mail: rabia.khelif@ifma.fr; Chateauneuf, Alaa [LGC-University Blaise Pascal, Campus des Cezeaux, BP 206, 63174 Aubiere Cedex (France)], E-mail: alaa.chateauneuf@polytech.univ-bpclermont.fr; Chaoui, Kamel [LR3MI, Departement de Genie Mecanique, Universite Badji Mokhtar, BP 12, Annaba 23000 (Algeria)], E-mail: chaoui@univ-annaba.org

    2007-12-15

    Lifetime management of underground pipelines is mandatory for safe hydrocarbon transmission and distribution systems. The use of high-density polyethylene tubes subjected to internal pressure, external loading and environmental variations requires a reliability study in order to define the service limits and the optimal operating conditions. In service, the time-dependent phenomena, especially creep, take place during the pipe lifetime, leading to significant strength reduction. In this work, the reliability-based assessment of pipe lifetime models is carried out, in order to propose a probabilistic methodology for lifetime model selection and to determine the pipe safety levels as well as the most important parameters for pipeline reliability. This study is enhanced by parametric analysis on pipe configuration, gas pressure and operating temperature.

  10. Reliability and Validity Assessment of a Linear Position Transducer

    Science.gov (United States)

    Garnacho-Castaño, Manuel V.; López-Lastra, Silvia; Maté-Muñoz, José L.

    2015-01-01

    The objectives of the study were to determine the validity and reliability of peak velocity (PV), average velocity (AV), peak power (PP) and average power (AP) measurements were made using a linear position transducer. Validity was assessed by comparing measurements simultaneously obtained using the Tendo Weightlifting Analyzer Systemi and T-Force Dynamic Measurement Systemr (Ergotech, Murcia, Spain) during two resistance exercises, bench press (BP) and full back squat (BS), performed by 71 trained male subjects. For the reliability study, a further 32 men completed both lifts using the Tendo Weightlifting Analyzer Systemz in two identical testing sessions one week apart (session 1 vs. session 2). Intraclass correlation coefficients (ICCs) indicating the validity of the Tendo Weightlifting Analyzer Systemi were high, with values ranging from 0.853 to 0.989. Systematic biases and random errors were low to moderate for almost all variables, being higher in the case of PP (bias ±157.56 W; error ±131.84 W). Proportional biases were identified for almost all variables. Test-retest reliability was strong with ICCs ranging from 0.922 to 0.988. Reliability results also showed minimal systematic biases and random errors, which were only significant for PP (bias -19.19 W; error ±67.57 W). Only PV recorded in the BS showed no significant proportional bias. The Tendo Weightlifting Analyzer Systemi emerged as a reliable system for measuring movement velocity and estimating power in resistance exercises. The low biases and random errors observed here (mainly AV, AP) make this device a useful tool for monitoring resistance training. Key points This study determined the validity and reliability of peak velocity, average velocity, peak power and average power measurements made using a linear position transducer The Tendo Weight-lifting Analyzer Systemi emerged as a reliable system for measuring movement velocity and power. PMID:25729300

  11. A human reliability assessment screening method for the NRU upgrade project

    International Nuclear Information System (INIS)

    Bremner, F.M.; Alsop, C.J.

    1997-01-01

    The National Research Universal (NRU) reactor is a 130MW, low pressure, heavy water cooled and moderated research reactor. The reactor is used for research, both in support of Canada's CANDU development program, and for a wide variety of other research applications. In addition, NRU plays an important part in the production of medical isotopes, e.g., generating 80% of worldwide supplies of Molybdenum-99. NRU is owned and operated by Atomic Energy of Canada Ltd. (AECL), and is currently undergoing upgrading as part of AECL's continuing commitment to operate their facilities in a safe manner. As part of these upgrades both deterministic and probabilistic safety assessments are being carried out. It was recognized that the assignment of Human Error Probabilities (HEPs) is an important part of the Probabilistic Safety Assessment (PSA) studies, particularly for a facility whose design predates modern ergonomic practices, and which will undergo a series of backfitted modifications whilst continuing to operate. A simple Human Reliability Assessment (HRA) screening method, looking at both pre- and post-accident errors, was used in the initial safety studies. However, following review of this method within AECL and externally by the regulator, it was judged that benefits could be gained for future error reduction by including additional features, as later described in this document. The HRA development project consisted of several stages; needs analysis, literature review, development of method (including testing and evaluation), and implementation. This paper discusses each of these stages in further detail. (author)

  12. Nanoscale deformation measurements for reliability assessment of material interfaces

    Science.gov (United States)

    Keller, Jürgen; Gollhardt, Astrid; Vogel, Dietmar; Michel, Bernd

    2006-03-01

    With the development and application of micro/nano electronic mechanical systems (MEMS, NEMS) for a variety of market segments new reliability issues will arise. The understanding of material interfaces is the key for a successful design for reliability of MEMS/NEMS and sensor systems. Furthermore in the field of BIOMEMS newly developed advanced materials and well known engineering materials are combined despite of fully developed reliability concepts for such devices and components. In addition the increasing interface-to volume ratio in highly integrated systems and nanoparticle filled materials are challenges for experimental reliability evaluation. New strategies for reliability assessment on the submicron scale are essential to fulfil the needs of future devices. In this paper a nanoscale resolution experimental method for the measurement of thermo-mechanical deformation at material interfaces is introduced. The determination of displacement fields is based on scanning probe microscopy (SPM) data. In-situ SPM scans of the analyzed object (i.e. material interface) are carried out at different thermo-mechanical load states. The obtained images are compared by grayscale cross correlation algorithms. This allows the tracking of local image patterns of the analyzed surface structure. The measurement results are full-field displacement fields with nanometer resolution. With the obtained data the mixed mode type of loading at material interfaces can be analyzed with highest resolution for future needs in micro system and nanotechnology.

  13. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    Jung, J.

    1983-12-01

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

  14. Assessment of beam tube performance for the maple research reactor

    International Nuclear Information System (INIS)

    Lee, A.G.

    1986-06-01

    The MAPLE research reactor is a versatile new research facility that can be adapted to meet the requirements of a variety of reactor applications. A particular group of reactor applications involves the use of beams of radiation extracted from the reactor core via tubes that penetrate through the biological shield and terminate in the reflector surrounding the fuelled core. An assessment is given of the neutron and gamma radiation fields entering beam tubes that are located radially or tangentially with respect to the core

  15. Uncertainty evaluation of reliability of shutdown system of a medium size fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zeliang, Chireuding; Singh, Om Pal, E-mail: singhop@iitk.ac.in; Munshi, Prabhat

    2016-11-15

    Highlights: • Uncertainty analysis of reliability of Shutdown System is carried out. • Monte Carlo method of sampling is used. • The effect of various reliability improvement measures of SDS are accounted. - Abstract: In this paper, results are presented on the uncertainty evaluation of the reliability of Shutdown System (SDS) of a Medium Size Fast Breeder Reactor (MSFBR). The reliability analysis results are of Kumar et al. (2005). The failure rate of the components of SDS are taken from International literature and it is assumed that these follow log-normal distribution. Fault tree method is employed to propagate the uncertainty in failure rate from components level to shutdown system level. The beta factor model is used to account different extent of diversity. The Monte Carlo sampling technique is used for the analysis. The results of uncertainty analysis are presented in terms of the probability density function, cumulative distribution function, mean, variance, percentile values, confidence intervals, etc. It is observed that the spread in the probability distribution of SDS failure rate is less than SDS components failure rate and ninety percent values of the failure rate of SDS falls below the target value. As generic values of failure rates are used, sensitivity analysis is performed with respect to failure rate of control and safety rods and beta factor. It is discovered that a large increase in failure rate of SDS rods is not carried to SDS system failure proportionately. The failure rate of SDS is very sensitive to the beta factor of common cause failure between the two systems of SDS. The results of the study provide insight in the propagation of uncertainty in the failure rate of SDS components to failure rate of shutdown system.

  16. A study on a reliability assessment methodology for the VHTR safety systems

    International Nuclear Information System (INIS)

    Lee, Hyung Sok

    2012-02-01

    The passive safety system of a 300MWt VHTR (Very High Temperature Reactor)which has attracted worldwide attention recently is actively considered for designing the improvement in the safety of the next generation nuclear power plant. The passive system functionality does not rely on an external source of the electrical support system,but on an intelligent use of the natural phenomena, such as convection, conduction, radiation, and gravity. It is not easy to evaluate quantitatively the reliability of the passive safety for the risk analysis considering the existing active system failure since the classical reliability assessment method could not be applicable. Therefore a new reliability methodology needs to be developed and applied for evaluating the reliability of the conceptual designed VHTR in this study. The preliminary evaluation and conceptualization are performed using the concept of the load and capacity theory related to the reliability physics model. The method of response surface method (RSM) is also utilized for evaluating the maximum temperature of nuclear fuel in this study. The significant variables and their correlation are considered for utilizing the GAMMA+ code. The proposed method might contribute to designing the new passive system of the VHTR

  17. Mechanical reliability of falling systems. Case of control rods of nuclear reactors

    International Nuclear Information System (INIS)

    Rezkallah, B.

    1991-05-01

    After having briefly recalled the main principles adopted for nuclear reactor safety (barrier based method, emergency stop system, emergency cooling system) and how control rod operation anomalies are described, this research thesis first gives an overview of knowledge in the field of reliability (principles, qualitative and quantitative methods, failure trees, performance function, shock modelling and mode analysis). Then, the author presents the modelling of the drop and rotation of a control rod. The computation technique is the modal synthesis. The author highlights the influence of physical (friction, initial conditions) and numerical (modal truncation) parameters. He reports a probabilistic analysis of the problem and proposes a stochastic model for the different shock forces (notably friction). He compares the results with those obtained with the DEMT software. The last part reports the investigation of a rare case: the jamming or untimely stop of the control rod device

  18. Mammography image assessment; validity and reliability of current scheme

    International Nuclear Information System (INIS)

    Hill, C.; Robinson, L.

    2015-01-01

    Mammographers currently score their own images according to criteria set out by Regional Quality Assurance. The criteria used are based on the ‘Perfect, Good, Moderate, Inadequate’ (PGMI) marking criteria established by the National Health Service Breast Screening Programme (NHSBSP) in their Quality Assurance Guidelines of 2006 1 . This document discusses the validity and reliability of the current mammography image assessment scheme. Commencing with a critical review of the literature this document sets out to highlight problems with the national approach to the use of marking schemes. The findings suggest that ‘PGMI’ scheme is flawed in terms of reliability and validity and is not universally applied across the UK. There also appear to be differences in schemes used by trainees and qualified mammographers. Initial recommendations are to be made in collaboration with colleagues within the National Health Service Breast Screening Programme (NHSBSP), Higher Education Centres, College of Radiographers and the Royal College of Radiologists in order to identify a mammography image appraisal scheme that is fit for purpose. - Highlights: • Currently no robust evidence based marking tools in use for the assessment of images in mammography. • Is current system valid, reliable and robust? • How can the current image assessment tool be improved? • Should students and qualified mammographers use the same tool? • What marking criteria are available for image assessment?

  19. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

    2005-02-01

    The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

  20. Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR)

    International Nuclear Information System (INIS)

    Gilmore, W.E.; Blackman, H.S.; Ryan, T.G.

    1986-01-01

    The Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR) program is a multiyear effort sponsored by the NRC and is being conducted at the Idaho National Engineering Laboratory (INEL). The goal of this program is to establish and operate computerized data base management tools for the human reliability data bank specification developed by Comer and Donovan. The NRC and the risk analysis community recognized that implementing a fully functional library would not be feasible, or practical, without the aid of computerized tools for management and manipulation of its data sources. The end users of the NUCLARR can be classified into three categories according to specific needs. The first category is those users interested in reviewing individual data sources for a given situation. The second category of users selects multiple data sources for a specific case, summarizing the information, and performing comparative studies. The last category of users interfaces the NUCLARR with other programming applications, such as other data banks, and simulation models of risk assessment. Project status is provided in the paper

  1. Evaluation and improvement in nondestructive examination (NDE) reliability for in-service inspection of light water reactors

    International Nuclear Information System (INIS)

    Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1988-01-01

    The evaluation and improvement of NDE Reliability for In-service Inspection (ISI) of Light Water Reactors (NDE Reliability) Program at Pacific Northwest Laboratory (PNL) was established to determine the reliability of current ISI techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this NRC program are to: determine the reliability of ultrasonic ISI performed on commercial light-water reactor (LWR) primary systems; determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to ensure a suitably low failure probability using probabilistic fracture mechanics analysis; evaluate the degree of reliability improvement that could be achieved using improved and advanced NDE technique; and recommend revisions to ASME Code, Section XI, and Regulatory Requirements, based on material properties, service conditions, and NDE uncertainties, that will ensure suitably low failure probabilities. The program consists of three basic tasks: a Piping task, a Pressure Vessel task, and an Evaluation and Improvement in NDE Reliability task. The major efforts were concentrated in the Piping task and the Evaluation and Improvement in NDE Reliability task

  2. Reliability Assessment for Low-cost Unmanned Aerial Vehicles

    Science.gov (United States)

    Freeman, Paul Michael

    Existing low-cost unmanned aerospace systems are unreliable, and engineers must blend reliability analysis with fault-tolerant control in novel ways. This dissertation introduces the University of Minnesota unmanned aerial vehicle flight research platform, a comprehensive simulation and flight test facility for reliability and fault-tolerance research. An industry-standard reliability assessment technique, the failure modes and effects analysis, is performed for an unmanned aircraft. Particular attention is afforded to the control surface and servo-actuation subsystem. Maintaining effector health is essential for safe flight; failures may lead to loss of control incidents. Failure likelihood, severity, and risk are qualitatively assessed for several effector failure modes. Design changes are recommended to improve aircraft reliability based on this analysis. Most notably, the control surfaces are split, providing independent actuation and dual-redundancy. The simulation models for control surface aerodynamic effects are updated to reflect the split surfaces using a first-principles geometric analysis. The failure modes and effects analysis is extended by using a high-fidelity nonlinear aircraft simulation. A trim state discovery is performed to identify the achievable steady, wings-level flight envelope of the healthy and damaged vehicle. Tolerance of elevator actuator failures is studied using familiar tools from linear systems analysis. This analysis reveals significant inherent performance limitations for candidate adaptive/reconfigurable control algorithms used for the vehicle. Moreover, it demonstrates how these tools can be applied in a design feedback loop to make safety-critical unmanned systems more reliable. Control surface impairments that do occur must be quickly and accurately detected. This dissertation also considers fault detection and identification for an unmanned aerial vehicle using model-based and model-free approaches and applies those

  3. An analysis of operational experience during low power and shutdown and a plan for addressing human reliability assessment issues

    International Nuclear Information System (INIS)

    Barriere, M.; Luckas, W.; Whitehead, D.; Ramey-Smith, A.

    1994-06-01

    Recent nuclear power plant events (e.g. Chernobyl, Diablo Canyon, and Vogtle) and US Nuclear Regulatory Commission (NRC) reports (e.g. NUREG-1449) have led to concerns regarding human reliability during low power and shutdown (LP ampersand S) conditions and limitations of human reliability analysis (HRA) methodologies in adequately representing the LP ampersand S environment. As a result of these concerns, the NRC initiated two parallel research projects to assess the influence of LP ampersand S conditions on human reliability through an analysis of operational experience at pressurized water reactors (PWRs) an boiling water reactors (BWRs). These research projects, performed by Brookhaven National Laboratory for PWRS, and Sandia National Laboratories for BWRs, identified unique aspects of human performance during LP ampersand S conditions and provided a program plan for research and development necessary to improve existing HRA methodologies. This report documents the results of the analysis of LP ampersand S operating experience and describes the improved HRA program plan

  4. Reliability-based condition assessment of steel containment and liners

    International Nuclear Information System (INIS)

    Ellingwood, B.; Bhattacharya, B.; Zheng, R.

    1996-11-01

    Steel containments and liners in nuclear power plants may be exposed to aggressive environments that may cause their strength and stiffness to decrease during the plant service life. Among the factors recognized as having the potential to cause structural deterioration are uniform, pitting or crevice corrosion; fatigue, including crack initiation and propagation to fracture; elevated temperature; and irradiation. The evaluation of steel containments and liners for continued service must provide assurance that they are able to withstand future extreme loads during the service period with a level of reliability that is sufficient for public safety. Rational methodologies to provide such assurances can be developed using modern structural reliability analysis principles that take uncertainties in loading, strength, and degradation resulting from environmental factors into account. The research described in this report is in support of the Steel Containments and Liners Program being conducted for the US Nuclear Regulatory Commission by the Oak Ridge National Laboratory. The research demonstrates the feasibility of using reliability analysis as a tool for performing condition assessments and service life predictions of steel containments and liners. Mathematical models that describe time-dependent changes in steel due to aggressive environmental factors are identified, and statistical data supporting the use of these models in time-dependent reliability analysis are summarized. The analysis of steel containment fragility is described, and simple illustrations of the impact on reliability of structural degradation are provided. The role of nondestructive evaluation in time-dependent reliability analysis, both in terms of defect detection and sizing, is examined. A Markov model provides a tool for accounting for time-dependent changes in damage condition of a structural component or system. 151 refs

  5. A model for assessing human cognitive reliability in PRA studies

    International Nuclear Information System (INIS)

    Hannaman, G.W.; Spurgin, A.J.; Lukic, Y.

    1985-01-01

    This paper summarizes the status of a research project sponsored by EPRI as part of the Probabilistic Risk Assessment (PRA) technology improvement program and conducted by NUS Corporation to develop a model of Human Cognitive Reliability (HCR). The model was synthesized from features identified in a review of existing models. The model development was based on the hypothesis that the key factors affecting crew response times are separable. The inputs to the model consist of key parameters the values of which can be determined by PRA analysts for each accident situation being assessed. The output is a set of curves which represent the probability of control room crew non-response as a function of time for different conditions affecting their performance. The non-response probability is then a contributor to the overall non-success of operating crews to achieve a functional objective identified in the PRA study. Simulator data and some small scale tests were utilized to illustrate the calibration of interim HCR model coefficients for different types of cognitive processing since the data were sparse. The model can potentially help PRA analysts make human reliability assessments more explicit. The model incorporates concepts from psychological models of human cognitive behavior, information from current collections of human reliability data sources and crew response time data from simulator training exercises

  6. Statistical reliability assessment of software-based systems

    International Nuclear Information System (INIS)

    Korhonen, J.; Pulkkinen, U.; Haapanen, P.

    1997-01-01

    Plant vendors nowadays propose software-based systems even for the most critical safety functions. The reliability estimation of safety critical software-based systems is difficult since the conventional modeling techniques do not necessarily apply to the analysis of these systems, and the quantification seems to be impossible. Due to lack of operational experience and due to the nature of software faults, the conventional reliability estimation methods can not be applied. New methods are therefore needed for the safety assessment of software-based systems. In the research project Programmable automation systems in nuclear power plants (OHA), financed together by the Finnish Centre for Radiation and Nuclear Safety (STUK), the Ministry of Trade and Industry and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. This volume in the OHA-report series deals with the statistical reliability assessment of software based systems on the basis of dynamic test results and qualitative evidence from the system design process. Other reports to be published later on in OHA-report series will handle the diversity requirements in safety critical software-based systems, generation of test data from operational profiles and handling of programmable automation in plant PSA-studies. (orig.) (25 refs.)

  7. Interrater reliability of videotaped observational gait-analysis assessments.

    Science.gov (United States)

    Eastlack, M E; Arvidson, J; Snyder-Mackler, L; Danoff, J V; McGarvey, C L

    1991-06-01

    The purpose of this study was to determine the interrater reliability of videotaped observational gait-analysis (VOGA) assessments. Fifty-four licensed physical therapists with varying amounts of clinical experience served as raters. Three patients with rheumatoid arthritis who demonstrated an abnormal gait pattern served as subjects for the videotape. The raters analyzed each patient's most severely involved knee during the four subphases of stance for the kinematic variables of knee flexion and genu valgum. Raters were asked to determine whether these variables were inadequate, normal, or excessive. The temporospatial variables analyzed throughout the entire gait cycle were cadence, step length, stride length, stance time, and step width. Generalized kappa coefficients ranged from .11 to .52. Intraclass correlation coefficients (2,1) and (3,1) were slightly higher. Our results indicate that physical therapists' VOGA assessments are only slightly to moderately reliable and that improved interrater reliability of the assessments of physical therapists utilizing this technique is needed. Our data suggest that there is a need for greater standardization of gait-analysis training.

  8. The European reliability data system. An organized information exchange on the operation of European nuclear reactors

    International Nuclear Information System (INIS)

    Mancini, G.; Amesz, J.; Bastianini, P.; Capobianchi, S.

    1983-01-01

    The paper revises the aims and objectives of the European Reliability Data System (ERDS), a centralized system collecting and organizing, at European level, information related to the operation of LWRs. The ERDS project was started in 1977 and after a preliminary feasibility study that ended in 1979 is now proceeding towards the final design and implementation stages. ERDS exploits information collected in national data systems and information deriving from single reactor sources. The paper describes first the development of the four data banks constituting the system: Component Event Data Bank, CEDB; Abnormal Occurrences Reporting System, AORS; Operating Unit Status Report, OUSR; and Generic Reliability Parameter Data Bank, GRPDB. Several typical aspects concerning the project are then outlined from the need of homogeneization of data and therefore the need for setting up reference classifications, to the problem of data transcoding and input into the system. Furthermore, the need is stressed of involving much more deeply nuclear power plant operators into the process of data acquisition by providing them with a useful feedback from the data analysis. (author)

  9. Some aspects of the reliability-based design of reactor containment structures

    International Nuclear Information System (INIS)

    Schueller, G.I.

    1975-01-01

    It is generally recognized that the load which a structure is likely to experience during its design life as well as its resistance are to be represented by random variables. A rational design procedure for reactor containment structures can therefore only be carried out within a probabilistic framework. Internal load conditions caused by system failure such as loss-of-coolant accident, pressure loads etc., and external load conditions caused for instance by impact due to aircraft crashes, external pressure waves and natural hazards such as earthquakes, floods, hurricanes are described by extreme value distributions of the Fisher-Tippett types. Statistical and physical arguments are given to support their application. The occurrence of these rare events with respect to time is modeled by a Poisson process. The yield strength of the containment structure for both steel (liner) and reinforced concrete shells is also modeled by extreme value distributions (of the smallest values). The failure criterion considered here is that of collapse determined by plastic yieldline formation. A failure mechanism as considered here describes a particular regime of plastic line formation. The probability of failure of a structure under a single load application of load types likely to occur during the design life of the structure is to be determined by integrating over all possible mechanisms. Finally Freudenthal's reliability function is utilized to combine the information derived above so that a containment design for given design lifes and reliabilities is possible. (orig.) [de

  10. 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

  11. MAPLE research reactor safety uncertainty assessment methodology

    International Nuclear Information System (INIS)

    Sills, H.E.; Duffey, R.B.; Andres, T.H.

    1999-01-01

    The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

  12. The prediction of reliability and residual life of reactor pressure components

    International Nuclear Information System (INIS)

    Nemec, J.; Antalovsky, S.

    1978-01-01

    The paper deals with the problem of PWR pressure components reliability and residual life evaluation and prediction. A physical model of damage cumulation which serves as a theoretical basis for all considerations presents two major aspects. The first one describes the dependence of the degree of damage in the crack leading-edge in pressure components on the reactor system load-time history, i.e. on the number of transient loads. Both stages, fatigue crack initiation and growth through the wall until the critical length is reached, are investigated. The crack is supposed to initiate at the flaws in a strength weld joint or in the bimetallic weld of the base ferritic steel and the austenitic stainless overlay cladding. The growth rates of developed cracks are analysed in respect to different load-time histories. Important cyclic properties of some steels are derived from the low-cycle fatigue theory. The second aspect is the load-time history-dependent process of precipitation, deformation and radiation aging, characterized entirely by the critical crack-length value mentioned above. The fracture point, defined by the equation ''crack-length=critical value'' and hence the residual life, can be evaluated using this model and verified by in-service inspection. The physical model described is randomized by considering all the parameters of the model as random. Monte Carlo methods are applied and fatigue crack initiation and growth is simulated. This permits evaluation of the reliability and residual life of the component. The distributions of material and load-time history parameters are needed for such simulation. Both the deterministic and computer-simulated probabilistic predictions of reliability and residual life are verified by prior-to-failure sequential testing of data coming from in-service NDT periodical inspections. (author)

  13. The seismic assessment of fast reactor cores in the UK

    International Nuclear Information System (INIS)

    Duthie, J.C.; Dostal, M.

    1988-01-01

    The design of the UK Commercial Demonstration Fast Reactor (CDFR) has evolved over a number of years. The design has to meet two seismic requirements: (i) the reactor must cause no hazard to the public during or after the Safe Shutdown Earthquake (SSE); (ii) there must be no sudden reduction in safety for an earthquake exceeding the SSE. The core is a complicated component in the whole reactor. It is usually represented in a very simplified manner in the seismic assessment of the whole reactor station. From this calculation, a time history or response spectrum can be generated for the diagrid, which supports the core, and for the above core structure, which supports the main absorber rods. These data may then be used to perform a detailed assessment of the reactor core. A new simplified model of the core response may then be made and used in a further calculation of the whole reactor. The calculation of the core response only, is considered in the remainder of this paper. One important feature of the fast reactor core, compared with other reactors, is that the components are relatively thin and flexible to promote neutron economy and heat transfer. A further important feature is that there are very small gaps between the wrapper tubes. This leads to very strong fluid-coupling effects. These effects are likely to be beneficial, but adequate techniques to calculate them are only just being developed. 9 refs, figs

  14. Soil-structure interaction effects on the reliability evaluation of reactor containments

    International Nuclear Information System (INIS)

    Pires, J.; Hwang, H.; Reich, M.

    1986-01-01

    The probability-based method for the seismic reliability assessment of nuclear structures, which has been developed at Brookhaven National Laboratory (BNL), is extended to include the effects of soil-structure interaction. A reinforced concrete containment building is analyzed in order to examine soil-structure interaction effects on: (1) structural fragilities; (2) floor response spectra statistics; and (3) correlation coefficients for total acceleration responses at specified structural locations

  15. Development of reliability and probabilistic safety assessment program RiskA

    International Nuclear Information System (INIS)

    Wu, Yican

    2015-01-01

    Highlights: • There are four parts in the structure of RiskA. User input part lets users input the PSA model and some necessary data by GUI or model transformation tool. In calculation engine part, fault tree analysis, event tree analysis, uncertainty analysis, sensitivity analysis, importance analysis and failure mode and effects analysis are supplied. User output part outputs the analysis results, user customized reports and some other data. The last part includes reliability database, some other common tools and help documents. • RiskA has several advanced features. Extensible framework makes it easy to add any new functions, making RiskA to be a large platform of reliability and probabilistic safety assessment. It is very fast to analysis fault tree in RiskA because many advanced algorithm improvement were made. Many model formats can be imported and exported, which made the PSA model in the commercial software can be easily transformed to adapt RiskA platform. Web-based co-modeling let several users in different places work together whenever they are online. • The comparison between RiskA and other mature PSA codes (e.g. CAFTA, RiskSpectrum, XFTA) has demonstrated that the calculation and analysis of RiskA is correct and efficient. Based on the development of this code package, many applications of safety and reliability analysis of some research reactors and nuclear power plants were performed. The development of RiskA appears to be of realistic and potential value for academic research and practical operation safety management of nuclear power plants in China and abroad. - Abstract: PSA (probabilistic safety assessment) software, the indispensable tool in nuclear safety assessment, has been widely used. An integrated reliability and PSA program named RiskA has been developed by FDS Team. RiskA supplies several standard PSA modules including fault tree analysis, event tree analysis, uncertainty analysis, failure mode and effect analysis and reliability

  16. Design and reliability of a didactic inphographic rubric assessment

    Directory of Open Access Journals (Sweden)

    Yunuen Ixchel GUZMÁN-CEDILLO

    2017-12-01

    Full Text Available The objective of this study is to describe design, validity process and reliability of a rubric assessment to evaluate didactic infographics quality. Participants were fifteen judges who participate in different moments of elaboration rubric process; it was made in three process phases: design, settings and reliability determination. Content validity was obtained by percentage agreement between 3 judges by component of the rubric; likewise a Krippendorff’s alpha were applied (a = .710 in pilot assessment with 5 infographics in order to set possible writings contradictions between components and criteria of performance. The intern consistence was determined by Cronbach’s alpha (? = .806 in 22 infographics gradation. An Intraclass correlation coefficient icc (a = .909 was applied to 6 judges qualifications also a Krippendorff’s alpha (a = .538 both of them in ordinal levels. The rubric is composed by 9 components, 3 performance levels, definitions of each component and assignments how to use the rubric. Results suggest the rubric is valid and reliable to grade quality of didactic infographic.

  17. L-Reactor operation, Savannah River Plant: environmental assessment

    International Nuclear Information System (INIS)

    1982-08-01

    The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

  18. Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

    International Nuclear Information System (INIS)

    Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

    2010-01-01

    In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

  19. Reliability and Validity Assessment of a Linear Position Transducer

    Directory of Open Access Journals (Sweden)

    Manuel V. Garnacho-Castaño

    2015-03-01

    Full Text Available The objectives of the study were to determine the validity and reliability of peak velocity (PV, average velocity (AV, peak power (PP and average power (AP measurements were made using a linear position transducer. Validity was assessed by comparing measurements simultaneously obtained using the Tendo Weightlifting Analyzer Systemi and T-Force Dynamic Measurement Systemr (Ergotech, Murcia, Spain during two resistance exercises, bench press (BP and full back squat (BS, performed by 71 trained male subjects. For the reliability study, a further 32 men completed both lifts using the Tendo Weightlifting Analyzer Systemz in two identical testing sessions one week apart (session 1 vs. session 2. Intraclass correlation coefficients (ICCs indicating the validity of the Tendo Weightlifting Analyzer Systemi were high, with values ranging from 0.853 to 0.989. Systematic biases and random errors were low to moderate for almost all variables, being higher in the case of PP (bias ±157.56 W; error ±131.84 W. Proportional biases were identified for almost all variables. Test-retest reliability was strong with ICCs ranging from 0.922 to 0.988. Reliability results also showed minimal systematic biases and random errors, which were only significant for PP (bias -19.19 W; error ±67.57 W. Only PV recorded in the BS showed no significant proportional bias. The Tendo Weightlifting Analyzer Systemi emerged as a reliable system for measuring movement velocity and estimating power in resistance exercises. The low biases and random errors observed here (mainly AV, AP make this device a useful tool for monitoring resistance training.

  20. Influence of temperature measurement accuracy and reliability on WWER-440 reactor operation

    International Nuclear Information System (INIS)

    Petenyi, V.; Ricany, J.

    2001-01-01

    The WWER-440 reactor power is controlled by coolant heat-up measurements installed on hot and cold circulation loops (enthalpy rise). For power distribution determination the thermocouples installed in reactor vessel above the fuel assemblies are mainly utilised. The paper shortly presents some interesting observations of temperature measurements influencing the reactor power operation of revealed changes in reactor core behaviour. (Authors)

  1. 105-B Reactor museum feasibility assessment (Phase 2) project

    International Nuclear Information System (INIS)

    Heckel, R. P.

    2000-01-01

    This 105-B Reactor Museum feasibility assessment project report documents project activities that have been performed, including a review and assessment of previously existing information, a walk-through of the facility, an assessment of potential hazards, and selection of mitigative measures deemed to be appropriate to allow unescorted access by members of the public to a specified primary tour route

  2. Reliability Assessment of Transformerless PV Inverters Considering Mission Profiles

    DEFF Research Database (Denmark)

    Yang, Yongheng; Wang, Huai; Blaabjerg, Frede

    2015-01-01

    reliability of three transformerless inverters under a yearly mission profile (i.e., solar irradiance and ambient temperature). The mission profile is translated to device thermal loading, which is used for lifetime prediction. Compar¬ison results reveal the lifetime mismatches among the power switching......Due to the small volume and high efficiency, transformerless inverters have gained much popularity in grid-connected PV applications, where minimizing leakage current injection is mandatory. This can be achieved either by modifying the modulation schemes or adding extra power switching devices......, resulting in an uneven distribution of the power losses on the switching devices. Consequently, the device thermal loading is redis-tributed, and thus may alter the entire inverter reliability performance, especially under a long-term operation. In this consideration, this paper assesses the device...

  3. Improving risk assessment by defining consistent and reliable system scenarios

    Directory of Open Access Journals (Sweden)

    B. Mazzorana

    2009-02-01

    Full Text Available During the entire procedure of risk assessment for hydrologic hazards, the selection of consistent and reliable scenarios, constructed in a strictly systematic way, is fundamental for the quality and reproducibility of the results. However, subjective assumptions on relevant impact variables such as sediment transport intensity on the system loading side and weak point response mechanisms repeatedly cause biases in the results, and consequently affect transparency and required quality standards. Furthermore, the system response of mitigation measures to extreme event loadings represents another key variable in hazard assessment, as well as the integral risk management including intervention planning. Formative Scenario Analysis, as a supplement to conventional risk assessment methods, is a technique to construct well-defined sets of assumptions to gain insight into a specific case and the potential system behaviour. By two case studies, carried out (1 to analyse sediment transport dynamics in a torrent section equipped with control measures, and (2 to identify hazards induced by woody debris transport at hydraulic weak points, the applicability of the Formative Scenario Analysis technique is presented. It is argued that during scenario planning in general and with respect to integral risk management in particular, Formative Scenario Analysis allows for the development of reliable and reproducible scenarios in order to design more specifically an application framework for the sustainable assessment of natural hazards impact. The overall aim is to optimise the hazard mapping and zoning procedure by methodologically integrating quantitative and qualitative knowledge.

  4. Significance assessment of small-medium sized reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

    2002-12-01

    Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

  5. Reliability assessment of passive isolation condenser system of AHWR using APSRA methodology

    International Nuclear Information System (INIS)

    Nayak, A.K.; Jain, Vikas; Gartia, M.R.; Prasad, Hari; Anthony, A.; Bhatia, S.K.; Sinha, R.K.

    2009-01-01

    In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) is used for evaluation of reliability of passive isolation condenser system of the Indian Advanced Heavy Water Reactor (AHWR). As per the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to perform the design basis function. The methodology first determines the operational characteristics of the system and the failure conditions based on a predetermined failure criterion. The parameters that could degrade the system performance are identified and considered for analysis. Different modes of failure and their cause are identified. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the isolation condenser system performance. Once the failure surface of the system is predicted, the causes of failure are examined through root diagnosis, which occur mainly due to failure of mechanical components. Reliability of the system is evaluated through a classical PSA treatment based on the failure probability of the components using generic data

  6. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qiming, E-mail: liqiming@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Tian, Jian; Zhou, Chong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Wang, Naxiu, E-mail: wangnaxiu@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-06-15

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail.

  7. A new method to evaluate the sealing reliability of the flanged connections for Molten Salt Reactors

    International Nuclear Information System (INIS)

    Li, Qiming; Tian, Jian; Zhou, Chong; Wang, Naxiu

    2015-01-01

    Highlights: • We novelly valuate the sealing reliability of the flanged connections for MSRs. • We focus on the passive decrease of the leak impetus in flanged connections. • The modified flanged connections are acquired a sealing ability of self-adjustment. • Effects of redesigned flange configurations on molten salt leakage are discussed. - Abstract: The Thorium based Molten Salt Reactor (TMSR) project is a future Generation IV nuclear reactor system proposed by the Chinese Academy of Sciences with the strategic goal of meeting the growing energy needs in the Chinese economic development and social progress. It is based on liquid salts served as both fuel and primary coolant and consequently great challenges are brought into the sealing of the flanged connections. In this study, an improved prototype flange assembly is performed on the strength of the Freeze-Flange initially developed by Oak Ridge National Laboratory (ORNL). The calculation results of the finite element model established to analyze the temperature profile of the Freeze-Flange agree well with the experimental data, which indicates that the numerical simulation method is credible. For further consideration, the ideal-gas thermodynamic model, together with the mathematical approximation, is novelly borrowed to theoretically evaluate the sealing performance of the modified Freeze-Flange and the traditional double gaskets bolted flange joint. This study focuses on the passive decrease of the leak driving force due to multiple gaskets introduced in flanged connections for MSR. The effects of the redesigned flange configuration on molten salt leakage resistance are discussed in detail

  8. The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

    International Nuclear Information System (INIS)

    Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

    2014-01-01

    The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

  9. Reliability, compliance, and security in web-based course assessments

    Directory of Open Access Journals (Sweden)

    Scott Bonham

    2008-04-01

    Full Text Available Pre- and postcourse assessment has become a very important tool for education research in physics and other areas. The web offers an attractive alternative to in-class paper administration, but concerns about web-based administration include reliability due to changes in medium, student compliance rates, and test security, both question leakage and utilization of web resources. An investigation was carried out in introductory astronomy courses comparing pre- and postcourse administration of assessments using the web and on paper. Overall no difference was seen in performance due to the medium. Compliance rates fluctuated greatly, and factors that seemed to produce higher rates are identified. Notably, email reminders increased compliance by 20%. Most of the 559 students complied with requests to not copy, print, or save questions nor use web resources; about 1% did copy some question text and around 2% frequently used other windows or applications while completing the assessment.

  10. A comparative evaluation of five human reliability assessment techniques

    International Nuclear Information System (INIS)

    Kirwan, B.

    1988-01-01

    A field experiment was undertaken to evaluate the accuracy, usefulness, and resources requirements of five human reliability quantification techniques (Techniques for Human Error Rate Prediction (THERP); Paired Comparisons, Human Error Assessment and Reduction Technique (HEART), Success Liklihood Index Method (SLIM)-Multi Attribute Utility Decomposition (MAUD), and Absolute Probability Judgement). This was achieved by assessing technique predictions against a set of known human error probabilities, and by comparing their predictions on a set of five realistic Probabilisitc Risk Assessment (PRA) human error. On a combined measure of accuracy THERP and Absolute Probability Judgement performed best, whilst HEART showed indications of accuracy and was lower in resources usage than other techniques. HEART and THERP both appear to benefit from using trained assessors in order to obtain the best results. SLIM and Paired Comparisons require further research on achieving a robust calibration relationship between their scale values and absolute probabilities. (author)

  11. Assessment of Astrid reactor pit design options

    International Nuclear Information System (INIS)

    Verpoest, Thomas; Villedieu, Alexandre; Robin, Jean-Charles

    2014-01-01

    Answering the French Act of the 28. of June 2006 about nuclear materials and waste management, the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) Project has the objectives to demonstrate the industrial feasibility based on identified domains (safety, operability, economy) of Sodium-cooled Fast Reactor and to perform transmutation demonstrations. The pre-conceptual design, started in 2010, considers several reactor pit design options. One of the objectives is to define a reference configuration for the ASTRID project which is able to answer safety and design requirements. The components addressed in this article are: the safety vessel and the Decay Heat Removal system through the main vessel. The core catcher associated to the different configurations studied in this article is an internal core catcher (inside the main vessel). This article deals with the different locations of the DHR through the main vessel and the type of the safety vessel (supported versus suspended vessel). These options are studied in order to establish the advantages and drawbacks of the different configurations in terms of economy, safety, In Service Inspection and Repair (ISIR), operability, robustness, and project risk (authors)

  12. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

    Baxter, D.C.; Dabiri, A.E.

    1983-01-01

    A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

  13. Reliability of Volumetry and Perimetry to Assess Knee Volume.

    Science.gov (United States)

    Nunes, Guilherme S; Yamashitafuji, Igor; Wageck, Bruna; Teixeira, Guilherme Garcia; Karloh, Manuela; de Noronha, Marcos

    2016-08-24

    The treatment of edema after a knee injury is usually 1 of the main objectives during rehabilitation. To assess the success of treatment, 2 methods are commonly used in clinical practice: volumetry and perimetry. To investigate the intra- and interassessor reliability of volumetry and perimetry to assess knee volume. Cross-sectional. Laboratory. 45 healthy participants (26 women) with mean age of 22.4 ± 2.8 y. Knee volume was assessed by 3 assessors (A, B, and C) with 3 methods (lower-limb volumetry [LLV], knee volumetry [KV], and knee perimetry [KP]). Assessor A was the most-experienced assessor, and assessor C, the least experienced. LLV and KV were performed with participants in the orthostatic position, while KP was performed with participants in supine. For the interassessor analysis, the ICC2,1 was high (.82) for KV and very high for LLV (.99) and KP (.99). For the intra-assessor analysis, ICC2,1 ranged from moderate to high for KV (.69-.83) and was very high for LLV (.99) and KP (.97-.99). KV, LLV, and KP are reliable methods, both intra- and interassessor, to measure knee volume.

  14. Assessing Reliability of Cold Spray Sputter Targets in Photovoltaic Manufacturing

    Science.gov (United States)

    Hardikar, Kedar; Vlcek, Johannes; Bheemreddy, Venkata; Juliano, Daniel

    2017-10-01

    Cold spray has been used to manufacture more than 800 Cu-In-Ga (CIG) sputter targets for deposition of high-efficiency photovoltaic thin films. It is a preferred technique since it enables high deposit purity and transfer of non-equilibrium alloy states to the target material. In this work, an integrated approach to reliability assessment of such targets with deposit weight in excess of 50 lb. is undertaken, involving thermal-mechanical characterization of the material in as-deposited condition, characterization of the interface adhesion on cylindrical substrate in as-deposited condition, and developing means to assess target integrity under thermal-mechanical loads during the physical vapor deposition (PVD) sputtering process. Mechanical characterization of cold spray deposited CIG alloy is accomplished through the use of indentation testing and adaptation of Brazilian disk test. A custom lever test was developed to characterize adhesion along the cylindrical interface between the CIG deposit and cylindrical substrate, overcoming limitations of current standards. A cohesive zone model for crack initiation and propagation at the deposit interface is developed and validated using the lever test and later used to simulate the potential catastrophic target failure in the PVD process. It is shown that this approach enables reliability assessment of sputter targets and improves robustness.

  15. An Assessment of the VHTR Safety Distance Using the Reliability Physics Model

    International Nuclear Information System (INIS)

    Lee, Joeun; Kim, Jintae; Jae, Moosung

    2015-01-01

    In Korea planning the production of hydrogen using high temperature from nuclear power is in progress. To produce hydrogen from nuclear plants, supplying temperature above 800 .deg. C is required. Therefore, Very High Temperature Reactor (VHTR) which is able to provide about 950 .deg. C is suitable. In situation of high temperature and corrosion where hydrogen might be released easily, hydrogen production facility using VHTR has a danger of explosion. Moreover explosion not only has a bad influence upon facility itself but also on VHTR. Those explosions result in unsafe situation that cause serious damage. However, In terms of thermal-hydraulics view, long distance makes low efficiency Thus, in this study, a methodology for the safety assessment of safety distance between the hydrogen production facilities and the VHTR is developed with reliability physics model. Based on the standard safety criteria which is a value of 1 x 10 -6 , the safety distance between the hydrogen production facilities and the VHTR using reliability physics model are calculated to be a value of 60m - 100m. In the future, assessment for characteristic of VHTR, the capacity to resist pressure from outside hydrogen explosion and the overpressure for the large amount of detonation volume in detail is expected to identify more precise safety distance using this reliability physics model

  16. Method for assessing reliability of a network considering probabilistic safety assessment

    International Nuclear Information System (INIS)

    Cepin, M.

    2005-01-01

    A method for assessment of reliability of the network is developed, which uses the features of the fault tree analysis. The method is developed in a way that the increase of the network under consideration does not require significant increase of the model. The method is applied to small examples of network consisting of a small number of nodes and a small number of their connections. The results give the network reliability. They identify equipment, which is to be carefully maintained in order that the network reliability is not reduced, and equipment, which is a candidate for redundancy, as this would improve network reliability significantly. (author)

  17. Assessing high reliability via Bayesian approach and accelerated tests

    International Nuclear Information System (INIS)

    Erto, Pasquale; Giorgio, Massimiliano

    2002-01-01

    Sometimes the assessment of very high reliability levels is difficult for the following main reasons: - the high reliability level of each item makes it impossible to obtain, in a reasonably short time, a sufficient number of failures; - the high cost of the high reliability items to submit to life tests makes it unfeasible to collect enough data for 'classical' statistical analyses. In the above context, this paper presents a Bayesian solution to the problem of estimation of the parameters of the Weibull-inverse power law model, on the basis of a limited number (say six) of life tests, carried out at different stress levels, all higher than the normal one. The over-stressed (i.e. accelerated) tests allow the use of experimental data obtained in a reasonably short time. The Bayesian approach enables one to reduce the required number of failures adding to the failure information the available a priori engineers' knowledge. This engineers' involvement conforms to the most advanced management policy that aims at involving everyone's commitment in order to obtain total quality. A Monte Carlo study of the non-asymptotic properties of the proposed estimators and a comparison with the properties of maximum likelihood estimators closes the work

  18. Reliability assessment of distribution power systems including distributed generations

    International Nuclear Information System (INIS)

    Megdiche, M.

    2004-12-01

    Nowadays, power systems have reached a good level of reliability. Nevertheless, considering the modifications induced by the connections of small independent producers to distribution networks, there's a need to assess the reliability of these new systems. Distribution networks present several functional characteristics, highlighted by the qualitative study of the failures, as dispersed loads at several places, variable topology and some electrotechnical phenomena which must be taken into account to model the events that can occur. The adopted reliability calculations method is Monte Carlo simulations, the probabilistic method most powerful and most flexible to model complex operating of the distribution system. We devoted a first part on the case of a 20 kV feeder to which a cogeneration unit is connected. The method was applied to a software of stochastic Petri nets simulations. Then a second part related to the study of a low voltage power system supplied by dispersed generations. Here, the complexity of the events required to code the method in an environment of programming allowing the use of power system calculations (load flow, short-circuit, load shedding, management of units powers) in order to analyse the system state for each new event. (author)

  19. Seismic reliability assessment methodology for CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Stephens, M.J.; Nessim, M.A.; Hong, H.P.

    1995-05-01

    A study was undertaken to develop a reliability-based methodology for the assessment of existing CANDU concrete containment structures with respect to seismic loading. The focus of the study was on defining appropriate specified values and partial safety factors for earthquake loading and resistance parameters. Key issues addressed in the work were the identification of an approach to select design earthquake spectra that satisfy consistent safety levels, and the use of structure-specific data in the evaluation of structural resistance. (author). 23 refs., 9 tabs., 15 figs

  20. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  1. Complex method to calculate objective assessments of information systems protection to improve expert assessments reliability

    Science.gov (United States)

    Abdenov, A. Zh; Trushin, V. A.; Abdenova, G. A.

    2018-01-01

    The paper considers the questions of filling the relevant SIEM nodes based on calculations of objective assessments in order to improve the reliability of subjective expert assessments. The proposed methodology is necessary for the most accurate security risk assessment of information systems. This technique is also intended for the purpose of establishing real-time operational information protection in the enterprise information systems. Risk calculations are based on objective estimates of the adverse events implementation probabilities, predictions of the damage magnitude from information security violations. Calculations of objective assessments are necessary to increase the reliability of the proposed expert assessments.

  2. Reliable scar scoring system to assess photographs of burn patients.

    Science.gov (United States)

    Mecott, Gabriel A; Finnerty, Celeste C; Herndon, David N; Al-Mousawi, Ahmed M; Branski, Ludwik K; Hegde, Sachin; Kraft, Robert; Williams, Felicia N; Maldonado, Susana A; Rivero, Haidy G; Rodriguez-Escobar, Noe; Jeschke, Marc G

    2015-12-01

    Several scar-scoring scales exist to clinically monitor burn scar development and maturation. Although scoring scars through direct clinical examination is ideal, scars must sometimes be scored from photographs. No scar scale currently exists for the latter purpose. We modified a previously described scar scale (Yeong et al., J Burn Care Rehabil 1997) and tested the reliability of this new scale in assessing burn scars from photographs. The new scale consisted of three parameters as follows: scar height, surface appearance, and color mismatch. Each parameter was assigned a score of 1 (best) to 4 (worst), generating a total score of 3-12. Five physicians with burns training scored 120 representative photographs using the original and modified scales. Reliability was analyzed using coefficient of agreement, Cronbach alpha, intraclass correlation coefficient, variance, and coefficient of variance. Analysis of variance was performed using the Kruskal-Wallis test. Color mismatch and scar height scores were validated by analyzing actual height and color differences. The intraclass correlation coefficient, the coefficient of agreement, and Cronbach alpha were higher for the modified scale than those of the original scale. The original scale produced more variance than that in the modified scale. Subanalysis demonstrated that, for all categories, the modified scale had greater correlation and reliability than the original scale. The correlation between color mismatch scores and actual color differences was 0.84 and between scar height scores and actual height was 0.81. The modified scar scale is a simple, reliable, and useful scale for evaluating photographs of burn patients. Copyright © 2015 Elsevier Inc. All rights reserved.

  3. A review of the probabilistic safety assessment application to the TR-2 research reactor

    International Nuclear Information System (INIS)

    Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

    2001-01-01

    A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

  4. Reliability assessment using Bayesian networks. Case study on quantative reliability estimation of a software-based motor protection relay

    International Nuclear Information System (INIS)

    Helminen, A.; Pulkkinen, U.

    2003-06-01

    In this report a quantitative reliability assessment of motor protection relay SPAM 150 C has been carried out. The assessment focuses to the methodological analysis of the quantitative reliability assessment using the software-based motor protection relay as a case study. The assessment method is based on Bayesian networks and tries to take the full advantage of the previous work done in a project called Programmable Automation System Safety Integrity assessment (PASSI). From the results and experiences achieved during the work it is justified to claim that the assessment method presented in the work enables a flexible use of qualitative and quantitative elements of reliability related evidence in a single reliability assessment. At the same time the assessment method is a concurrent way of reasoning one's beliefs and references about the reliability of the system. Full advantage of the assessment method is taken when using the method as a way to cultivate the information related to the reliability of software-based systems. The method can also be used as a communicational instrument in a licensing process of software-based systems. (orig.)

  5. The obscure factor analysis on the vibration reliability of the internals of nuclear power plant reactor and anti-vibration measures

    International Nuclear Information System (INIS)

    Fu Geyan; Zhu Qirong

    1998-11-01

    It is pointed out that the main reason making nuclear power plants reactors leak is the vibration of internals of reactors. The factors which lead the vibration all have randomness and obscureness. The obscure reliability theory is introduced to the vibration system of internals of nuclear power reactor. Based on a quantity of designing and moving data, the obscure factors effecting the vibration reliability of the internals of nuclear power plant reactor are analyzed and the anti-vibration reliability criteria and the evaluating model are given. And the anti-vibration reliability measures are advanced from different quarters of the machine design and building, the thermohydraulics design, the control of reactivity, etc.. They may benefit the theory and practice for building and perfecting the vibration obscure reliability model of the reactor internals

  6. Assessment of the integrity of WWER type reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

  7. Method and apparatus for a nuclear reactor for increasing reliability to scram control elements

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1976-01-01

    A description is given of a method and apparatus for increasing the reliability of linear drive devices of a nuclear reactor to scram the control elements held in a raised position thereby. Each of the plurality of linear drive devices includes a first type of holding means associated with the drive means of the linear drive device and a second type of holding means distinct and operatively dissimilar from the first type. The system of linear drive devices having both types of holding means are operated in such a manner that the control elements of a portion of the linear drive devices are only held in a raised position by the first holding means and the control elements of the remaining portion of linear drive devices are held in a raised position by only the second type of holding means. Since the two types of holding means are distinct from one another and are operatively dissimilar, the probability of failure of both systems to scram as a result of common mode failure will be minimized. Means may be provided to positively detect disengagement of the first type of holding means and engagement of the second type of holding means for those linear drive devices being operative to hold the control elements in a raised position with the second type of holding means

  8. Afterheat assessment of a conceptual fusion reactor

    International Nuclear Information System (INIS)

    Jayatissa, S.P.; Goddard, A.J.H.

    1987-01-01

    Structural activation and decay heat deposition calculations have been undertaken for the DEMO fusion reactor design. The DEMO design was based on an earlier conceptual design of a blanket sector which could breed tritium and generate electricity. These calculations have taken account of the redistribution of energy by the transport of γ radiation. Calculated heat deposition patterns have been used as data for simplified heat transfer calculations to judge temperature rises in relation to materials limits in a severe accident involving complete coolant flow failure. (author)

  9. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Choi, Y.A.; Feltus, M.A.

    1995-01-01

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  10. Application of nonparametric statistics to material strength/reliability assessment

    International Nuclear Information System (INIS)

    Arai, Taketoshi

    1992-01-01

    An advanced material technology requires data base on a wide variety of material behavior which need to be established experimentally. It may often happen that experiments are practically limited in terms of reproducibility or a range of test parameters. Statistical methods can be applied to understanding uncertainties in such a quantitative manner as required from the reliability point of view. Statistical assessment involves determinations of a most probable value and the maximum and/or minimum value as one-sided or two-sided confidence limit. A scatter of test data can be approximated by a theoretical distribution only if the goodness of fit satisfies a test criterion. Alternatively, nonparametric statistics (NPS) or distribution-free statistics can be applied. Mathematical procedures by NPS are well established for dealing with most reliability problems. They handle only order statistics of a sample. Mathematical formulas and some applications to engineering assessments are described. They include confidence limits of median, population coverage of sample, required minimum number of a sample, and confidence limits of fracture probability. These applications demonstrate that a nonparametric statistical estimation is useful in logical decision making in the case a large uncertainty exists. (author)

  11. Application of fuzzy synthetic assessment to assess human factors design level on reactor control panel

    International Nuclear Information System (INIS)

    Peng Xuecheng

    1999-01-01

    Reactor control panel design level on human factors must be considered by designer. The author evaluated the human factor design level of arrangement and combinations including the switch buttons, meter dials and indication lamps on Minjiang Reactor and High-Flux Engineer Test Reactor (HFETR) critical device by application of fuzzy synthetic assessment method in mathematics. From the assessment results, the advantages and shortcomings are fount, and some modification suggestions have also been proposed

  12. Capabilities and limitations of fracture mechanics methods in the assessment of integrity of light water reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Burdekin, F M

    1988-12-31

    This document deals with fracture mechanics methods used for the assessment of Light Water Reactor (LWR) components. The background to analysis methods using elastic plastic parameters is described. Several results obtained with these methods are presented as well as results of reliability analysis methods. (TEC). 27 refs.

  13. Performance Assessment of Passive Gaseous Provisions (PGAP). Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-07-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000 on the basis of IAEA General Conference resolution GC(44)/RES/21. INPRO helps to ensure the availability of sustainable nuclear energy in the 21st century and seeks to bring together all interested Member States - both technology holders and technology users - to consider joint actions to achieve desired innovations. To contribute to an international consensus on the definition of the reliability of passive systems that involve natural circulation, and on a methodology to assess this reliability, INPRO initiated a collaborative project on Performance Assessment of Passive Gaseous Provisions (PGAP) in 2007. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones, not only to enhance the operational safety of the reactors but also to mitigate the consequences of a severe accident should one occur. However, the reliability of passive safety systems is crucial and must be assessed before they are used extensively in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are a priori unknown. The functions of many passive systems are based on thermohydraulic principles, which until recently were considered as not being subject to any kind of failure. Hence, large and consistent efforts are required to quantify the reliability of such systems. Three participants from three INPRO Member States were involved in this collaborative project. Reliability methods for passive systems (RMPS) and assessment of passive system reliability (APSRA) methodologies were used by the participants to assess the performance and reliability of the passive decay heat removal system of the French gas cooled fast reactor design for station blackout and a loss of coolant accident combined with loss of off-site power, respectively. This publication presents the

  14. Reliability assessment using degradation models: bayesian and classical approaches

    Directory of Open Access Journals (Sweden)

    Marta Afonso Freitas

    2010-04-01

    Full Text Available Traditionally, reliability assessment of devices has been based on (accelerated life tests. However, for highly reliable products, little information about reliability is provided by life tests in which few or no failures are typically observed. Since most failures arise from a degradation mechanism at work for which there are characteristics that degrade over time, one alternative is monitor the device for a period of time and assess its reliability from the changes in performance (degradation observed during that period. The goal of this article is to illustrate how degradation data can be modeled and analyzed by using "classical" and Bayesian approaches. Four methods of data analysis based on classical inference are presented. Next we show how Bayesian methods can also be used to provide a natural approach to analyzing degradation data. The approaches are applied to a real data set regarding train wheels degradation.Tradicionalmente, o acesso à confiabilidade de dispositivos tem sido baseado em testes de vida (acelerados. Entretanto, para produtos altamente confiáveis, pouca informação a respeito de sua confiabilidade é fornecida por testes de vida no quais poucas ou nenhumas falhas são observadas. Uma vez que boa parte das falhas é induzida por mecanismos de degradação, uma alternativa é monitorar o dispositivo por um período de tempo e acessar sua confiabilidade através das mudanças em desempenho (degradação observadas durante aquele período. O objetivo deste artigo é ilustrar como dados de degradação podem ser modelados e analisados utilizando-se abordagens "clássicas" e Bayesiana. Quatro métodos de análise de dados baseados em inferência clássica são apresentados. A seguir, mostramos como os métodos Bayesianos podem também ser aplicados para proporcionar uma abordagem natural à análise de dados de degradação. As abordagens são aplicadas a um banco de dados real relacionado à degradação de rodas de trens.

  15. Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

    International Nuclear Information System (INIS)

    Atkins, D.F.; Lillie, A.F.

    1983-01-01

    As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

  16. Reactor technology assessment and selection utilizing systems engineering approach

    Science.gov (United States)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  17. Reliability worth assessment of radial systems with distributed generation

    OpenAIRE

    Bellart Llavall, Francesc Xavier

    2010-01-01

    With recent advances in technology, utilities generation (DG) on the distribution systems. Reliability worth is very important in power system planning and operation. Having a DG ensures reli increase the reliability worth. This research project presents the study of a radial distribution system and the impact of placing DG in order to increase the reliability worth. where a DG have to be placed. The reliability improvement is measured by different reliability indices tha...

  18. Modeling issues associated with production reactor safety assessment

    International Nuclear Information System (INIS)

    Stack, D.W.; Thomas, W.R.

    1990-01-01

    This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

  19. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  20. Preliminary nuclear power reactor technology qualitative assessment for Malaysia

    International Nuclear Information System (INIS)

    Shamsul Amri Sulaiman

    2011-01-01

    Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

  1. Proliferation resistance assessment of high temperature gas reactors

    International Nuclear Information System (INIS)

    Chikamatsu N, M. A.; Puente E, F.

    2014-10-01

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  2. Assessment of materials needs for fusion reactors

    International Nuclear Information System (INIS)

    Allison, G.S.

    1976-07-01

    This report has the goal of presenting for the CTR designer and material supplier potentially significant problem areas in materials manufacturing and in structural material resources projected for potential application in fusion power reactor construction. The projected material requirements are based on presently available bills-of-materials for conceptual CTR designs used for constructing a hypothetical fusion power generating capacity of 10 6 MW(e) maturing exponentially over a 20-year period. The projected elemental requirements, the ratio of these requirements to the projected total U.S. demand, and the salient problems currently identified with the CTR use of these elements are summarized. The projected requirements are based upon a ''model'' industry, which is described, and the estimated potential use of molybdenum, niobium, vanadium, and tantalum as blanket structural materials

  3. Assessment of materials needs for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Allison, G.S. (comp.)

    1976-07-01

    This report has the goal of presenting for the CTR designer and material supplier potentially significant problem areas in materials manufacturing and in structural material resources projected for potential application in fusion power reactor construction. The projected material requirements are based on presently available bills-of-materials for conceptual CTR designs used for constructing a hypothetical fusion power generating capacity of 10/sup 6/ MW(e) maturing exponentially over a 20-year period. The projected elemental requirements, the ratio of these requirements to the projected total U.S. demand, and the salient problems currently identified with the CTR use of these elements are summarized. The projected requirements are based upon a ''model'' industry, which is described, and the estimated potential use of molybdenum, niobium, vanadium, and tantalum as blanket structural materials.

  4. Development of the GO-FLOW reliability analysis methodology for nuclear reactor system

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi; Kobayashi, Michiyuki

    1994-01-01

    Probabilistic Safety Assessment (PSA) is important in the safety analysis of technological systems and processes, such as, nuclear plants, chemical and petroleum facilities, aerospace systems. Event trees and fault trees are the basic analytical tools that have been most frequently used for PSAs. Several system analysis methods can be used in addition to, or in support of, the event- and fault-tree analysis. The need for more advanced methods of system reliability analysis has grown with the increased complexity of engineered systems. The Ship Research Institute has been developing a new reliability analysis methodology, GO-FLOW, which is a success-oriented system analysis technique, and is capable of evaluating a large system with complex operational sequences. The research has been supported by the special research fund for Nuclear Technology, Science and Technology Agency, from 1989 to 1994. This paper describes the concept of the Probabilistic Safety Assessment (PSA), an overview of various system analysis techniques, an overview of the GO-FLOW methodology, the GO-FLOW analysis support system, procedure of treating a phased mission problem, a function of common cause failure analysis, a function of uncertainty analysis, a function of common cause failure analysis with uncertainty, and printing out system of the results of GO-FLOW analysis in the form of figure or table. Above functions are explained by analyzing sample systems, such as PWR AFWS, BWR ECCS. In the appendices, the structure of the GO-FLOW analysis programs and the meaning of the main variables defined in the GO-FLOW programs are described. The GO-FLOW methodology is a valuable and useful tool for system reliability analysis, and has a wide range of applications. With the development of the total system of the GO-FLOW, this methodology has became a powerful tool in a living PSA. (author) 54 refs

  5. Assessment of microelectronics packaging for high temperature, high reliability applications

    Energy Technology Data Exchange (ETDEWEB)

    Uribe, F.

    1997-04-01

    This report details characterization and development activities in electronic packaging for high temperature applications. This project was conducted through a Department of Energy sponsored Cooperative Research and Development Agreement between Sandia National Laboratories and General Motors. Even though the target application of this collaborative effort is an automotive electronic throttle control system which would be located in the engine compartment, results of this work are directly applicable to Sandia`s national security mission. The component count associated with the throttle control dictates the use of high density packaging not offered by conventional surface mount. An enabling packaging technology was selected and thermal models defined which characterized the thermal and mechanical response of the throttle control module. These models were used to optimize thick film multichip module design, characterize the thermal signatures of the electronic components inside the module, and to determine the temperature field and resulting thermal stresses under conditions that may be encountered during the operational life of the throttle control module. Because the need to use unpackaged devices limits the level of testing that can be performed either at the wafer level or as individual dice, an approach to assure a high level of reliability of the unpackaged components was formulated. Component assembly and interconnect technologies were also evaluated and characterized for high temperature applications. Electrical, mechanical and chemical characterizations of enabling die and component attach technologies were performed. Additionally, studies were conducted to assess the performance and reliability of gold and aluminum wire bonding to thick film conductor inks. Kinetic models were developed and validated to estimate wire bond reliability.

  6. Reliability of a semi-quantitative method for dermal exposure assessment (DREAM)

    NARCIS (Netherlands)

    Wendel de Joode, B. van; Hemmen, J.J. van; Meijster, T.; Major, V.; London, L.; Kromhout, H.

    2005-01-01

    Valid and reliable semi-quantitative dermal exposure assessment methods for epidemiological research and for occupational hygiene practice, applicable for different chemical agents, are practically nonexistent. The aim of this study was to assess the reliability of a recently developed

  7. Space-reactor electric systems: subsystem technology assessment

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-01-01

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified

  8. Risk-assessment methodology for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.

    1976-04-01

    The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed

  9. Probabilistic risk assessment course documentation. Volume 3. System reliability and analysis techniques, Session A - reliability

    International Nuclear Information System (INIS)

    Lofgren, E.V.

    1985-08-01

    This course in System Reliability and Analysis Techniques focuses on the quantitative estimation of reliability at the systems level. Various methods are reviewed, but the structure provided by the fault tree method is used as the basis for system reliability estimates. The principles of fault tree analysis are briefly reviewed. Contributors to system unreliability and unavailability are reviewed, models are given for quantitative evaluation, and the requirements for both generic and plant-specific data are discussed. Also covered are issues of quantifying component faults that relate to the systems context in which the components are embedded. All reliability terms are carefully defined. 44 figs., 22 tabs

  10. Modified Tuck Jump Assessment: Reliability and Training of Raters

    Directory of Open Access Journals (Sweden)

    Craig A. Smith, Nicole J. Chimera, Monica R. Lininger, Meghan Warren

    2017-09-01

    Full Text Available We are writing with regard to “Intra- and inter-rater reliability of the modified tuck jump assessment,” by Fort-Vanmeerhaeghe et al. (2017 published in the Journal of Sports Science & Medicine. The authors reported on the reliability of the modified Tuck Jump Assessment (TJA. The purpose of the article was twofold: to introduce a new scoring methodology and to report on the interrater and intrarater reliability. The authors found the modified TJA to have excellent interrater reliability (ICC = 0.94, 95% CI = 0.88-0.97 and intrarater reliability (rater 1 ICC = 0.94, 95% CI = 0.88-0.9; rater 2 ICC = 0.96, 95% CI = 0.92-0.98 with experienced raters (n = 2 in a sample of 24 elite volleyball athletes. Overall, we found the study to be well conducted and valuable to the field of injury screening; however, the study did not adequately explain how the raters were trained in the modified TJA to improve consistency of scoring, or the modifications of the individual flaw “excessive contact noise at landing.” This information is necessary to improve the clinical utility of the TJA and direct future reliability studies. The TJA has been changed at least three times in the literature: from the initial introduction (Myer et al., 2006 to the most referenced and detailed protocol (Myer et al., 2011 to the publication under discussion (Fort-Vanmeerhaeghe et al., 2017. The initial test protocol was based upon clinical expertise and has evolved over time as new research emerged and problems arose with the original TJA. Initially, the TJA was scored on a visual analog scale (Myer et al., 2006, changed to a dichotomous scale (0 for no flaw or 1 for flaw present (Myer et al., 2011 and most recently modified using an ordinal scale (Fort-Vanmeerhaeghe et al., 2017. A significant disparity in the reported interrater and intrarater reliability arose with the dichotomously scored TJA, between those involved in the development of the TJA (Herrington et al., 2013

  11. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    International Nuclear Information System (INIS)

    Khalil, Rah Man; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan

    2012-01-01

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article

  12. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khalil, Rah Man; Heo, Gyun Young [Kyung Hee Univ., Seoul (Korea, Republic of); Son, Han Seong [Joongbu Univ., Chungnam (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article.

  13. Usage models in reliability assessment of software-based systems

    Energy Technology Data Exchange (ETDEWEB)

    Haapanen, P.; Pulkkinen, U. [VTT Automation, Espoo (Finland); Korhonen, J. [VTT Electronics, Espoo (Finland)

    1997-04-01

    This volume in the OHA-project report series deals with the statistical reliability assessment of software based systems on the basis of dynamic test results and qualitative evidence from the system design process. Other reports to be published later on in the OHA-project report series will handle the diversity requirements in safety critical software-based systems, generation of test data from operational profiles and handling of programmable automation in plant PSA-studies. In this report the issues related to the statistical testing and especially automated test case generation are considered. The goal is to find an efficient method for building usage models for the generation of statistically significant set of test cases and to gather practical experiences from this method by applying it in a case study. The scope of the study also includes the tool support for the method, as the models may grow quite large and complex. (32 refs., 30 figs.).

  14. Usage models in reliability assessment of software-based systems

    International Nuclear Information System (INIS)

    Haapanen, P.; Pulkkinen, U.; Korhonen, J.

    1997-04-01

    This volume in the OHA-project report series deals with the statistical reliability assessment of software based systems on the basis of dynamic test results and qualitative evidence from the system design process. Other reports to be published later on in the OHA-project report series will handle the diversity requirements in safety critical software-based systems, generation of test data from operational profiles and handling of programmable automation in plant PSA-studies. In this report the issues related to the statistical testing and especially automated test case generation are considered. The goal is to find an efficient method for building usage models for the generation of statistically significant set of test cases and to gather practical experiences from this method by applying it in a case study. The scope of the study also includes the tool support for the method, as the models may grow quite large and complex. (32 refs., 30 figs.)

  15. Time-dependent reliability analysis and condition assessment of structures

    International Nuclear Information System (INIS)

    Ellingwood, B.R.

    1997-01-01

    Structures generally play a passive role in assurance of safety in nuclear plant operation, but are important if the plant is to withstand the effect of extreme environmental or abnormal events. Relative to mechanical and electrical components, structural systems and components would be difficult and costly to replace. While the performance of steel or reinforced concrete structures in service generally has been very good, their strengths may deteriorate during an extended service life as a result of changes brought on by an aggressive environment, excessive loading, or accidental loading. Quantitative tools for condition assessment of aging structures can be developed using time-dependent structural reliability analysis methods. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process

  16. Quantification of human reliability in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.; Dankg, Vinh N.

    1996-01-01

    Human performance may substantially influence the reliability and safety of complex technical systems. For this reason, Human Reliability Analysis (HRA) constitutes an important part of Probabilistic Safety Assessment (PSAs) or Quantitative Risk Analyses (QRAs). The results of these studies as well as analyses of past accidents and incidents clearly demonstrate the importance of human interactions. The contribution of human errors to the core damage frequency (CDF), as estimated in the Swedish nuclear PSAs, are between 15 and 88%. A survey of the FRAs in the Swiss PSAs shows that also for the Swiss nuclear power plants the estimated HE contributions are substantial (49% of the CDF due to internal events in the case of Beznau and 70% in the case of Muehleberg; for the total CDF, including external events, 25% respectively 20%). Similar results can be extracted from the PSAs carried out for French, German, and US plants. In PSAs or QRAs, the adequate treatment of the human interactions with the system is a key to the understanding of accident sequences and their relative importance to overall risk. The main objectives of HRA are: first, to ensure that the key human interactions are systematically identified and incorporated into the safety analysis in a traceable manner, and second, to quantify the probabilities of their success and failure. Adopting a structured and systematic approach to the assessment of human performance makes it possible to provide greater confidence that the safety and availability of human-machine systems is not unduly jeopardized by human performance problems. Section 2 discusses the different types of human interactions analysed in PSAs. More generally, the section presents how HRA fits in the overall safety analysis, that is, how the human interactions to be quantified are identified. Section 3 addresses the methods for quantification. Section 4 concludes the paper by presenting some recommendations and pointing out the limitations of the

  17. Assessment of Data Reliability of Wireless Sensor Network for Bioinformatics

    Directory of Open Access Journals (Sweden)

    Ting Dong

    2017-09-01

    Full Text Available As a focal point of biotechnology, bioinformatics integrates knowledge from biology, mathematics, physics, chemistry, computer science and information science. It generally deals with genome informatics, protein structure and drug design. However, the data or information thus acquired from the main areas of bioinformatics may not be effective. Some researchers combined bioinformatics with wireless sensor network (WSN into biosensor and other tools, and applied them to such areas as fermentation, environmental monitoring, food engineering, clinical medicine and military. In the combination, the WSN is used to collect data and information. The reliability of the WSN in bioinformatics is the prerequisite to effective utilization of information. It is greatly influenced by factors like quality, benefits, service, timeliness and stability, some of them are qualitative and some are quantitative. Hence, it is necessary to develop a method that can handle both qualitative and quantitative assessment of information. A viable option is the fuzzy linguistic method, especially 2-tuple linguistic model, which has been extensively used to cope with such issues. As a result, this paper introduces 2-tuple linguistic representation to assist experts in giving their opinions on different WSNs in bioinformatics that involve multiple factors. Moreover, the author proposes a novel way to determine attribute weights and uses the method to weigh the relative importance of different influencing factors which can be considered as attributes in the assessment of the WSN in bioinformatics. Finally, an illustrative example is given to provide a reasonable solution for the assessment.

  18. Data driven CAN node reliability assessment for manufacturing system

    Science.gov (United States)

    Zhang, Leiming; Yuan, Yong; Lei, Yong

    2017-01-01

    The reliability of the Controller Area Network(CAN) is critical to the performance and safety of the system. However, direct bus-off time assessment tools are lacking in practice due to inaccessibility of the node information and the complexity of the node interactions upon errors. In order to measure the mean time to bus-off(MTTB) of all the nodes, a novel data driven node bus-off time assessment method for CAN network is proposed by directly using network error information. First, the corresponding network error event sequence for each node is constructed using multiple-layer network error information. Then, the generalized zero inflated Poisson process(GZIP) model is established for each node based on the error event sequence. Finally, the stochastic model is constructed to predict the MTTB of the node. The accelerated case studies with different error injection rates are conducted on a laboratory network to demonstrate the proposed method, where the network errors are generated by a computer controlled error injection system. Experiment results show that the MTTB of nodes predicted by the proposed method agree well with observations in the case studies. The proposed data driven node time to bus-off assessment method for CAN networks can successfully predict the MTTB of nodes by directly using network error event data.

  19. Human reliability assessment on the basis of operating experience

    International Nuclear Information System (INIS)

    Straeter, O.

    1997-01-01

    For development of methodology, available models for qualitative assessment of human errors (e.g. by Swain, Hacker, Rasmussen) and a variety of known systematic approaches for quantitiative assessment of inadequate human action (e.g. THERP, ASEP, HCR, SLIM) were taken as a basis to establish a job specification, which in turn was used for developing a method for acquisition, characterisation and evaluation of errors. This method encompasses the two processes of event analysis and event evaluation: The first step comprises analysis of events by analysis of information describing the conditions and scenarios of relevance to the inadequate human action examined. In addition to the description of process sequences, information is taken into account on possible conditions that may bring about failure. As an assessment of human reliability requires manifold approaches for evaluation, a connectionistic procedure was developed for evaluation of the compilation of events based on a debate about various approaches from the domain of artificial intelligence (AI). This procedure yields both qualitative and quantitative information through a homogenous approach. (orig./GL) [de

  20. Seventh regular meeting of the International Working Group on Reliability of Reactor Pressure Components, Vienna, 3-5 September 1985

    International Nuclear Information System (INIS)

    1986-07-01

    The seventh regular meeting of the IAEA International Working Group on Reliability of Reactor Pressure Components was held at the Agency's Headquarters in Vienna from 3 to 5 September 1985. The representatives of Member States and of the Commission of the European Communities reported the status of the research programmes in this field (12 presentations). A separate abstract was prepared for each of the presentations

  1. How to use the ENEA data bank for the classification and reliability processing of fast reactor component event data

    International Nuclear Information System (INIS)

    Righini, R.

    1987-01-01

    This report describes the input and inquiry procedures for the Data Bank set-up by ENEA for reliability studies on fast reactors. With reference to the structure and to the codes to be applied in the data entry and in the inquiry, see Report (2) in references. The data contained into the Bank are absolutely confidential. The input and inquiry procedures describes in this report may be applied only by the user who have previously specified the suitable password

  2. Impact of data source on travel time reliability assessment.

    Science.gov (United States)

    2014-08-01

    Travel time reliability measures are becoming an increasingly important input to the mobility and : congestion management studies. In the case of Maryland State Highway Administration, reliability : measures are key elements in the agencys Annual ...

  3. Probablistic risk assessment methodology application to Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Babar, A.K.; Grover, R.B.; Mehra, V.K.; Gangwal, D.K.; Chakraborty, G.

    1987-01-01

    Probabilistic risk assessment in the context of nuclear power plants is associated with models that predict the offsite radiological releases resulting from reactor accidents. Level 1 PRA deals with the identification of accident sequences relevant to the design of a system and also with their quantitative estimation. It is characterised by event tree, fault tree analysis. The initiating events applicable to pressurised heavy water reactors have been considered and the dominating initiating events essential for detailed studies are identified in this paper. Reliability analysis and the associated problems encountered during the case studies are mentioned briefly. It is imperative to validate the failure data used for analysis. Bayesian technique has been employed for the same and a brief account is included herein. A few important observations, e.g. effects of the presence of moderator, made during the application of probabilistic risk assessment methodology are also discussed. (author)

  4. Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

    International Nuclear Information System (INIS)

    Kalchev, B.

    1995-01-01

    The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E -3 , 1/RY to 1.07E -3 , 1/RY. 5 refs., 2 tabs., 5 figs

  5. Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalchev, B [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E{sup -3}, 1/RY to 1.07E{sup -3}, 1/RY. 5 refs., 2 tabs., 5 figs.

  6. Portfolio assessment during medical internships: How to obtain a reliable and feasible assessment procedure?

    Science.gov (United States)

    Michels, Nele R M; Driessen, Erik W; Muijtjens, Arno M M; Van Gaal, Luc F; Bossaert, Leo L; De Winter, Benedicte Y

    2009-12-01

    A portfolio is used to mentor and assess students' clinical performance at the workplace. However, students and raters often perceive the portfolio as a time-consuming instrument. In this study, we investigated whether assessment during medical internship by a portfolio can combine reliability and feasibility. The domain-oriented reliability of 61 double-rated portfolios was measured, using a generalisability analysis with portfolio tasks and raters as sources of variation in measuring the performance of a student. We obtained reliability (Phi coefficient) of 0.87 with this internship portfolio containing 15 double-rated tasks. The generalisability analysis showed that an acceptable level of reliability (Phi = 0.80) was maintained when the amount of portfolio tasks was decreased to 13 or 9 using one and two raters, respectively. Our study shows that a portfolio can be a reliable method for the assessment of workplace learning. The possibility of reducing the amount of tasks or raters while maintaining a sufficient level of reliability suggests an increase in feasibility of portfolio use for both students and raters.

  7. Assessment of structural materials inside the reactor pool of the Dalat research reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

    2010-01-01

    Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

  8. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    International Nuclear Information System (INIS)

    1993-01-01

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs

  9. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs.

  10. Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

    International Nuclear Information System (INIS)

    Ergun, E; Kadiroglu, O.S.

    1999-01-01

    The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

  11. Reliability of a single objective measure in assessing sleepiness.

    Science.gov (United States)

    Sunwoo, Bernie Y; Jackson, Nicholas; Maislin, Greg; Gurubhagavatula, Indira; George, Charles F; Pack, Allan I

    2012-01-01

    To evaluate reliability of single objective tests in assessing sleepiness. Subjects who completed polysomnography underwent a 4-nap multiple sleep latency test (MSLT) the following day. Prior to each nap opportunity on MSLT, subjects performed the psychomotor vigilance test (PVT) and divided attention driving task (DADT). Results of single versus multiple test administrations were compared using the intraclass correlation coefficient (ICC) and adjusted for test administration order effects to explore time of day effects. Measures were explored as continuous and binary (i.e., impaired or not impaired). Community-based sample evaluated at a tertiary, university-based sleep center. 372 adult commercial vehicle operators oversampled for increased obstructive sleep apnea risk. N/A. AS CONTINUOUS MEASURES, ICC WERE AS FOLLOWS: MSLT 0.45, PVT median response time 0.69, PVT number of lapses 0.51, 10-min DADT tracking error 0.87, 20-min DADT tracking error 0.90. Based on binary outcomes, ICC were: MSLT 0.63, PVT number of lapses 0.85, 10-min DADT 0.95, 20-min DADT 0.96. Statistically significant time of day effects were seen in both the MSLT and PVT but not the DADT. Correlation between ESS and different objective tests was strongest for MSLT, range [-0.270 to -0.195] and persisted across all time points. Single DADT and PVT administrations are reliable measures of sleepiness. A single MSLT administration can reasonably discriminate individuals with MSL < 8 minutes. These results support the use of a single administration of some objective tests of sleepiness when performed under controlled conditions in routine clinical care.

  12. The collection, storage and use of equipment performance data for the safety and reliability assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Fothergill, C.D.H.

    1975-01-01

    It has been characteristic of the Nuclear Industry that it should grow up in an atmosphere where reliability and operational safety considerations have been of vital importance. Consequently all aspects of Nuclear Power Reactor design, construction and operation (in the U.K.A.E.A.) are subjected to rigorous reliability assessments, beginning with the automatic protective devices and the safety shut-down systems. This has resulted in the setting up of large and small private data stores to support this upsurgence of Safety and Reliability assessment work. Unfortunately, much of the information being stored and published falls short of the minimum requirements of Safety Assessors and Reliability Analysts who need to make use of it. That there is still an urgent need for more work to be done in the Reliability Data field is universally acknowledged. The characteristics which make up good quality reliability data must be defined and achievable minimum standards must be set for its identification, collection, storage and retrieval. To this end the United Kingdom Atomic Energy Authority have set up the Systems Reliability Service Data Bank. This includes a computerized storage facility comprised of two principal data stores: (i) Reliability Data Store, (ii) Event Data Store. The figures available in the Reliability Data Store range from those relating to the lifetimes of minute components to those obtained from the assessment of whole plants and complete assemblies. These data have been accumulated from many reliable sources both inside and outside the Nuclear Industry, including the transfer of 'live' data generated from the results of reliability surveillance exercises associated with Event Data collection. Computer techniques developed specifically for the Reliability Data Store enable further 'processing' of these data to be carried out. The Event Data Store consists of three discrete computerized data stores, each one providing the necessary storage, retrieval and

  13. Considerations concerning the reliability of reactor safety equipment; Considerations sur la fiabilite des ensembles de securite de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J; Guyot, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [French] On rappelle les circonstances qui favorisent au C.E.A. la collecte d'une information valable des resultats de la maintenance. L'importance des donnees a traiter a rendu necessaire l'utilisation d'une calculatrice poux l'analyse automatique des resultats recueillis. On se limitera ici aux aspects particuliers de la fiabilite du point de vue de l'electronique pour le controle et la commande de reacteurs nucleaires: pannes sures et pannes non sures; probabilite de survie dans le cas de la securite des reacteurs; facteur de disponibilite. Les schemas de principe des ensembles de securite definis pour deux types de reacteurs (reacteur de puissance et reacteur experimental de faible puissance) sont indiques. On

  14. Reliability assessment of restructured power systems using reliability network equivalent and pseudo-sequential simulation techniques

    International Nuclear Information System (INIS)

    Ding, Yi; Wang, Peng; Goel, Lalit; Billinton, Roy; Karki, Rajesh

    2007-01-01

    This paper presents a technique to evaluate reliability of a restructured power system with a bilateral market. The proposed technique is based on the combination of the reliability network equivalent and pseudo-sequential simulation approaches. The reliability network equivalent techniques have been implemented in the Monte Carlo simulation procedure to reduce the computational burden of the analysis. Pseudo-sequential simulation has been used to increase the computational efficiency of the non-sequential simulation method and to model the chronological aspects of market trading and system operation. Multi-state Markov models for generation and transmission systems are proposed and implemented in the simulation. A new load shedding scheme is proposed during generation inadequacy and network congestion to minimize the load curtailment. The IEEE reliability test system (RTS) is used to illustrate the technique. (author)

  15. Reliability analysis of the service water system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Tayt-Sohn, L.C.; Oliveira, L.F.S. de.

    1984-01-01

    A reliability analysis of the service water system is done aiming to use in the evaluation of the non reliability of the Component Cooling System (SRC) for great loss of cooling accidents in nuclear power plants. (E.G.) [pt

  16. Reliability analysis of the service water system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Fleming, P.V.; Frutuoso e Melo, P.F.F.; Tayt-Sohn, L.C.

    1983-01-01

    A reliability analysis of the service water system is done aiming to use in the evaluation of the non reliability of the component cooling system (SRC) for great loss of cooling accidents in nuclear power plants. (E.G.) [pt

  17. Guidelines for Self-assessment of Research Reactor Safety

    International Nuclear Information System (INIS)

    2018-01-01

    Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

  18. Advanced propulsion engine assessment based on a cermet reactor

    Science.gov (United States)

    Parsley, Randy C.

    1993-01-01

    A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

  19. Assessment of the enhanced DHRS configuration for MYRRHA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bubelis, E.; Jaeger, W. [KIT, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bandini, G. [ENEA, via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Alemberti, A.; Palmero, M. [ANSALDO, Corso Perrone 25, 16152 Genova (Italy)

    2016-10-15

    Highlights: • Innovative decay heat removal system (DHRS). • Heavy liquid metal cooled reactor. • Avoiding of lead bismuth eutectic (LBE) freezing. • Numerical assessment and proof of operational principles of innovative DHRS. - Abstract: This paper deals with the assessment of an innovative decay heat removal system for the MYRRHA reactor, based on the analysis of the selected transients with two different system codes. The application to liquid metal cooled reactors has the disadvantage of adding overcooling transients to the transient spectrum. Under these circumstances, freezing of the coolant can occur if no corrective or operator actions are taken in the medium and long term. Therefore, ANSALDO Nucleare invented an enhanced decay heat removal system which avoids the risk of freezing. The numerical assessment and proof of operational principles are performed by KIT and ENEA. The simulation results show that the freezing can be avoided. Moreover, both institutions calculate similar behavior during overcooling transients. This study will help to implement the novel decay heat removal system and the overall safety philosophy of innovative reactor concepts.

  20. Catalyzed deuterium fueled reversed-field pinch reactor assessment

    International Nuclear Information System (INIS)

    Dobrott, D.

    1985-01-01

    This study is part of a Department of Energy supported alternate fusion fuels program at Science Applications International Corporation. The purpose of this portion of the study is to perform an assessment of a conceptual compact reversed-field pinch reactor (CRFPR) that is fueled by the catalyzed-deuterium (Cat-d) fuel cycle with respect to physics, technology, safety, and cost. The Cat-d CRFPR is compared to a d-t fueled fusion reactor with respect to several issues in this study. The comparison includes cost, reactor performance, and technology requirements for a Cat-d fueled CRFPR and a comparable cost-optimized d-t fueled conceptual design developed by LANL

  1. Cost assessment of demo fusion reactor with considering maintenance

    International Nuclear Information System (INIS)

    Hashizume, Hidetoshi; Kitagoh, Kazutoshi

    2003-01-01

    The purpose of this study is to perform cost assessment of nuclear fusion reactors in order to draw up commercial plants. A fusion reactor may have a complex configuration to achieve high beta value, which leads to low and instable availability when maintenance is taken into account. Therefore, reactor's availability must be evaluated with considering the influence of the configuration complexity. Furthermore the availability has the strong impact on COE (Cost of Electricity), that is, a fusion reactor with low availability will not be accepted as a commercial plant. Therefore, we developed a new method to calculate availabilities with random numbers, in which the complexity of reactor's configuration could become considered. In addition, we considered the reduction of superconducting coil's maintenance time by introducing remountable magnet system because the coil maintenance requires quite long time in the present technology. The results show that the availability becomes relatively large if the short maintenance time of coils could be achieved, for example, by remountable magnetic systems. (author)

  2. Proven power reactor systems - novel features and developments in operation performance, safety and reliability

    International Nuclear Information System (INIS)

    Bugl, J.

    1975-01-01

    As the development of nuclear reactors for the generation of electric power started after the end of the Second World War, the prospective use of diverse materials as fuel, moderator and coolant resulted in a wide diversity of design possibilities. Of the 10 nuclear reactor types which were being considered most seriously in those days, only a few have achieved acceptance. This development is best illustrated by listing the nuclear power plants in service, under construction and on order at present, separately by reactor types (table). In the lead at present and for some years to come are the thermal reactors and especially the light water reactors (LWR). In the LWR group the lead is held by the pressurised water reactor (PWR) which accounts for 44% of the installed capacity of all the nuclear power plants in service at present. In the early 1980s this share will increase to 58%, whereas the share of the boiling water reactor (BWR) will increase to only 28% from 23% at present. (orig./TK) [de

  3. Human Reliability Analysis in Support of Risk Assessment for Positive Train Control

    Science.gov (United States)

    2003-06-01

    This report describes an approach to evaluating the reliability of human actions that are modeled in a probabilistic risk assessment : (PRA) of train control operations. This approach to human reliability analysis (HRA) has been applied in the case o...

  4. Reliability-based Assessment of Fatigue Life for Bridges

    DEFF Research Database (Denmark)

    Toft, Henrik Stensgaard; Sørensen, John Dalsgaard

    2012-01-01

    The reliability level for bridges is discussed based on a comparison of the reliability levels proposed and used by e.g. JCSS, ISO, NKB and Eurocodes. The influence of reserve capacity by which failure of a specific detail does not lead to structural collapse is investigated. The results show...

  5. Construction of Valid and Reliable Test for Assessment of Students

    Science.gov (United States)

    Osadebe, P. U.

    2015-01-01

    The study was carried out to construct a valid and reliable test in Economics for secondary school students. Two research questions were drawn to guide the establishment of validity and reliability for the Economics Achievement Test (EAT). It is a multiple choice objective test of five options with 100 items. A sample of 1000 students was randomly…

  6. Assessment of power reliability and improvement potential by using ...

    African Journals Online (AJOL)

    This paper presents the use of smart reclosers for improving reliability of a distribution system of one of the major cities of Ethiopia. As frequent power interruptions are posing a huge problem to the life of the people and the economy, finding a solution to the problem is very essential. Electric reliability has affected social well ...

  7. Application of material databases for improved reliability of reactor pressure vessels

    International Nuclear Information System (INIS)

    Griesbach, T.J.; Server, W.L.; Beaudoin, B.F.; Burgos, B.N.

    1994-01-01

    A vital part of reactor vessel Life Cycle Management program must begin with an accurate characterization of the vessel material properties. Uncertainties in vessel material properties or use of bounding values may result in unnecessary conservatisms in vessel integrity calculations. These conservatisms may be eliminated through a better understanding of the material properties in reactor vessels, both in the unirradiated and irradiated conditions. Reactor vessel material databases are available for quantifying the chemistry and Charpy shift behavior of individual heats of reactor vessel materials. Application of the databases for vessels with embrittlement concerns has proven to be an effective embrittlement management tool. This paper presents details of database development and applications which demonstrate the value of using material databases for improving material chemistry and for maximizing the data from integrated material surveillance programs

  8. Human performance analysis in the frame of probabilistic safety assessment of research reactors

    International Nuclear Information System (INIS)

    Farcasiu, Mita; Nitoi, Mirela; Apostol, Minodora; Turcu, I.; Florescu, Gh.

    2005-01-01

    Full text: The analysis of operating experience has identified the importance of human performance in reliability and safety of research reactors. In Probabilistic Safety Assessment (PSA) of nuclear facilities, human performance analysis (HPA) is used in order to estimate human error contribution to the failure of system components or functions. HPA is a qualitative and quantitative analysis of human actions identified for error-likely situations or accident-prone situations. Qualitative analysis is used to identify all man-machine interfaces that can lead to an accident, types of human interactions which may mitigate or exacerbate the accident, types of human errors and performance shaping factors. Quantitative analysis is used to develop estimates of human error probability as effects of human performance in reliability and safety. The goal of this paper is to accomplish a HPA in the PSA frame for research reactors. Human error probabilities estimated as results of human actions analysis could be included in system event tree and/or system fault tree. The achieved sensitivity analyses determine human performance sensibility at systematically variations both for dependencies level between human actions and for operator stress level. The necessary information was obtained from operating experience of research reactor TRIGA from INR Pitesti. The required data were obtained from generic data bases. (authors)

  9. Reliability assessment of Indian Point Unit 3 containment structure

    International Nuclear Information System (INIS)

    Kawakami, J.; Hwang, H.; Chang, M.T.; Reich, M.

    1984-01-01

    In the current design criteria, the load combinations specified for design of concrete containment structures are in the deterministic formats. However, by applying the probability-based reliability method developed by BNL to the concrete containment structures designed according to the criteria, it is possible to evaluate the reliability levels implied in the current design criteria. For this purpose, the reliability analysis is applied to the Indian Point Unit No. 3 containment. The details of the containment structure such as the geometries and the rebar arrangements, etc., are taken from the working drawings and the final safety analysis reports. Three kinds of loads are considered in the reliability analysis. They are, dead load (D), accidental pressure due to a large LOCA (P), and earthquake ground acceleration (E). Reliability analysis of the containment subjected to all combinations of loads is performed. Results are presented in this report

  10. Reliability Assessment and Reliability-Based Inspection and Maintenance of Offshore Wind Turbines

    DEFF Research Database (Denmark)

    Ramírez, José G. Rangel; Sørensen, John Dalsgaard

    2009-01-01

    Probabilistic methodologies represent an important tool to identify the suitable strategy to inspect and deal with the deterioration in structures such as offshore wind turbines (OWT). Reliability based methods such as Risk Based Inspection (RBI) planning may represent a proper methodology to opt...

  11. Reliability assessment of ballistic jump squats and bench throws.

    Science.gov (United States)

    Alemany, Joseph A; Pandorf, Clay E; Montain, Scott J; Castellani, John W; Tuckow, Alexander P; Nindl, Bradley C

    2005-02-01

    The purpose of this investigation was to determine the test-retest reliability and coefficient of variation of 2 novel physical performance tests. Ten healthy men (22.0 +/- 3.0 years, 87.0 +/- 8.0 kg, 20.0 +/- 5.0% body fat) performed 30 continuous and dynamic jump squats (JS) and bench throws (BT) on 4 separate occasions. The movements were performed under loaded conditions utilizing 30% of subject's predetermined 1 repetition maximum in the back squat and bench press. Mean power (MP; W), peak power (PP; W), mean velocity (MV; m.s(-1)), peak velocity (PV; m.s(-1)), and total work (TW; J) were assessed using a ballistic measurement system (Innervations Inc., Muncie, IN). Data were analyzed using repeated measures analysis of variance with Duncan's post hoc test when mean differences were p < or = 0.05. Intraclass correlation coefficient (ICC) and within-subject coefficient of variation (CV%) were also calculated. All values are presented as mean +/- SE. BT variables were statistically similar across the 4 sessions: MP (350.0 +/- 13.9 W), PP (431.4 +/- 18.5 W) MV (1.6 +/- 0.03 m.s(-1)), PV (2.0 +/- 0.03 m.s(-1)), and TW (199.1 +/- 7.2 J). For JS, session 3 PP (1,669.8 +/- 111.2 W) was significantly greater vs. sessions 1, 2, and 4 (1,601.2 +/- 58.4 W). Session 4 MP (1,403.2 +/- 88.6 W) and MV (1.9 +/- 0.1 m.s(-1)) for JS were significantly lower during sessions 1, 2, and 3 (MP: 1,479.4.5 +/- 44.8 W, MV: 2.0 +/- 0.05 m.s(-1)). TW (834.7 +/- 24.3 J) and PV (2.2 +/- 0.04 m.s(-1)) were statistically similar during all sessions for JS. The CVs ranged from 3.0 to 7.6% for the BT and 3.2 to 5.7% for the JS. ICCs for MP, PP, MV, PV, and TW were 0.92, 0.95, 0.94, 0.91, and 0.95, respectively, during BT. ICCs during JS for MP, PP, MV, PV, and TW were 0.96, 0.98, 0.94, 0.94, and 0.89, respectively. The results of the current study support the use of a 30 continuous and dynamic BT protocol as a reliable upper-body physical performance test, which can be administered with

  12. Lifetime assessment on PWR reactor vessel internals in Korea

    International Nuclear Information System (INIS)

    Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

  13. Reliability analysis of the automatic control and power supply of reactor equipment

    International Nuclear Information System (INIS)

    Monori, Pal; Nagy, J.A.; Meszaros, Zoltan; Konkoly, Laszlo; Szabo, Antal; Nagy, Laszlo

    1988-01-01

    Based on reliability analysis the shortcomings of nuclear facilities are discovered. Fault tree types constructed for the technology of automatic control and for power supply serve as input data of the ORCHARD 2 computer code. In order to charaterize the reliability of the system, availability, failure rates and time intervals between failures are calculated. The results of the reliability analysis of the feedwater system of the Paks Nuclear Power Plant showed that the system consisted of elements of similar reliabilities. (V.N.) 8 figs.; 3 tabs

  14. Time-dependent reliability analysis of nuclear reactor operators using probabilistic network models

    International Nuclear Information System (INIS)

    Oka, Y.; Miyata, K.; Kodaira, H.; Murakami, S.; Kondo, S.; Togo, Y.

    1987-01-01

    Human factors are very important for the reliability of a nuclear power plant. Human behavior has essentially a time-dependent nature. The details of thinking and decision making processes are important for detailed analysis of human reliability. They have, however, not been well considered by the conventional methods of human reliability analysis. The present paper describes the models for the time-dependent and detailed human reliability analysis. Recovery by an operator is taken into account and two-operators models are also presented

  15. Reactor Safety Assessment System--A situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

    Bray, M.A.; Sebo, D.E.; Dixon, B.W.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base that uses the parametric values, the known operator actions, and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant

  16. Reactor Safety Assessment System: a situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

    Bray, M.A.; Sebo, D.E.; Dixon, B.W.

    1985-04-01

    The Reactor Safety Assessment System is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base which uses the parametric values, the known operator actions and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant. 5 figs

  17. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    International Nuclear Information System (INIS)

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

  18. High-reliability logic system evaluation of a programmed multiprocessor solution. Application in the nuclear reactor safety field

    International Nuclear Information System (INIS)

    Lallement, Dominique.

    1979-01-01

    Nuclear reactors are monitored by several systems combined. The hydraulic and mechanical limitations on the equipment and the heat transfer requirements in the core set a reliable working range for the boiler defined with certain safety margins. The control system tends to keep the power plant within this working range. The protection system covers all the electrical and mechanical equipment needed to safeguard the boiler in the event of abnormal transients or accidents accounted for in the design of the plant. On units in service protection is handled by cabled automatic systems. For better reliability and safety operation, greater flexibility of use (modularity, adaptability) and improved start-up criteria by data processing the tendency is to use digital programmed systems. Computers are already present in control systems but their introduction into protection systems meets with some reticence on the part of the nuclear safety authorities. A study on the replacement of conventional by digital protection systems is presented. From choices partly made on the principles which should govern the hardware and software of a protection system the reliability of different structures and elements was examined and an experimental model built with its simulator and test facilities. A prototype based on these options and studies is being built and is to be set up on one of the CEN-G reactors for tests [fr

  19. Network reliability assessment using a cellular automata approach

    International Nuclear Information System (INIS)

    Rocco S, Claudio M.; Moreno, Jose Ali

    2002-01-01

    Two cellular automata (CA) models that evaluate the s-t connectedness and shortest path in a network are presented. CA based algorithms enhance the performance of classical algorithms, since they allow a more reliable and straightforward parallel implementation resulting in a dynamic network evaluation, where changes in the connectivity and/or link costs can readily be incorporated avoiding recalculation from scratch. The paper also demonstrates how these algorithms can be applied for network reliability evaluation (based on Monte-Carlo approach) and for finding s-t path with maximal reliability

  20. A method to assign failure rates for piping reliability assessments

    International Nuclear Information System (INIS)

    Gamble, R.M.; Tagart, S.W. Jr.

    1991-01-01

    This paper reports on a simplified method that has been developed to assign failure rates that can be used in reliability and risk studies of piping. The method can be applied on a line-by-line basis by identifying line and location specific attributes that can lead to piping unreliability from in-service degradation mechanisms and random events. A survey of service experience for nuclear piping reliability also was performed. The data from this survey provides a basis for identifying in-service failure attributes and assigning failure rates for risk and reliability studies

  1. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  2. Site Suitability and Hazard Assessment Guide for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wayne Moe

    2013-10-01

    Commercial nuclear reactor projects in the U.S. have traditionally employed large light water reactors (LWR) to generate regional supplies of electricity. Although large LWRs have consistently dominated commercial nuclear markets both domestically and abroad, the concept of small modular reactors (SMRs) capable of producing between 30 MW(t) and 900 MW(t) to generating steam for electricity is not new. Nor is the idea of locating small nuclear reactors in close proximity to and in physical connection with industrial processes to provide a long-term source of thermal energy. Growing problems associated continued use of fossil fuels and enhancements in efficiency and safety because of recent advancements in reactor technology suggest that the likelihood of near-term SMR technology(s) deployment at multiple locations within the United States is growing. Many different types of SMR technology are viable for siting in the domestic commercial energy market. However, the potential application of a particular proprietary SMR design will vary according to the target heat end-use application and the site upon which it is proposed to be located. Reactor heat applications most commonly referenced in connection with the SMR market include electric power production, district heating, desalinization, and the supply of thermal energy to various processes that require high temperature over long time periods, or a combination thereof. Indeed, the modular construction, reliability and long operational life purported to be associated with some SMR concepts now being discussed may offer flexibility and benefits no other technology can offer. Effective siting is one of the many early challenges that face a proposed SMR installation project. Site-specific factors dealing with support to facility construction and operation, risks to the plant and the surrounding area, and the consequences subsequent to those risks must be fully identified, analyzed, and possibly mitigated before a license

  3. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  4. Structural Reliability Methods for Wind Power Converter System Component Reliability Assessment

    DEFF Research Database (Denmark)

    Kostandyan, Erik; Sørensen, John Dalsgaard

    2012-01-01

    Wind power converter systems are essential subsystems in both off-shore and on-shore wind turbines. It is the main interface between generator and grid connection. This system is affected by numerous stresses where the main contributors might be defined as vibration and temperature loadings....... The temperature variations induce time-varying stresses and thereby fatigue loads. A probabilistic model is used to model fatigue failure for an electrical component in the power converter system. This model is based on a linear damage accumulation and physics of failure approaches, where a failure criterion...... is defined by the threshold model. The attention is focused on crack propagation in solder joints of electrical components due to the temperature loadings. Structural Reliability approaches are used to incorporate model, physical and statistical uncertainties. Reliability estimation by means of structural...

  5. On estimation of reliability of a nuclear power plant with tokamak reactor

    International Nuclear Information System (INIS)

    Klemin, A.I.; Smetannikov, V.P.; Shiverskij, E.A.

    1982-01-01

    The results of the analysis of INTOR plant reliability are presented. The first stage of the analysis consists in the calculation of the INTOR plant structural reliability factors (15 ibs main systems have been considered). For each system the failure flow parameter (W(1/h)) and operational readiness Ksub(r) have been determined which for the plant as a whole besides these factors-technological utilization coefficient Ksub(TU) and mean-cycles-between failures Tsub(o). The second stage of the reliability analysis consists in investigating methods of improving its reliability factors reratively to the one calculated at the first level stage. It is shown that the reliability of the whole plant to the most essential extent is determined by the power supply system reliability. The following as to the influence extent on the INTOR plant reliability is the cryogenic system. Calculations of the INTOR plant reliability factors have given the following values: W=4,5x10 -3 1/h. Tsub(o)=152 h, Ksub(r)=0,71, Ksub(TU)=o,4 g

  6. Failure mode and effect analysis experimental reliability determination for the CANDU reactor equipment

    International Nuclear Information System (INIS)

    Vieru, G.

    1996-01-01

    This paper describes the experimental tests performed in order to prove the reliability parameters for certain equipment manufactured in INR Pitesti, for NPP Cernavoda. The tests were provided by Technical Specifications and test procedures. A comparison, referring to the reliability parameters, between Canadian equipment and INR manufactured equipment ones is also given. The results of tests and conclusions are shown. (author)

  7. Reliability and Probabilistic Risk Assessment - How They Play Together

    Science.gov (United States)

    Safie, Fayssal M.; Stutts, Richard; Huang, Zhaofeng

    2015-01-01

    The objective of this presentation is to discuss the PRA process and the reliability engineering discipline, their differences and similarities, and how they are used as complimentary analyses to support design and flight decisions.

  8. Use of standard reliability levels in design and safety assessment of in-pile loops

    International Nuclear Information System (INIS)

    Bogani, G.; Verre, A.; Balestreri, S.; Colombo, A.G.; Luisi, T.

    1975-01-01

    This paper describes a logic-probabilistic analysis technique for a critical design review and safety assessment of in-pile loops. The examples in this paper refer to the analysis performed for the experimental loops already constructed or under construction in the ESSOR reactor of the Joint Research Centre of Ispra, as irradiation facilities for fuel element research and development tests. The proposed technique is based on the classification into categories of components and protective device malfunctions. Such subdivision into categories was agreed upon by the Italian Safety Authority and Euratom JRC, and adopted for the safety assessment of the ESSOR reactor in-pile loops. For each category, the method makes a link with a corresponding malfunction probability range (probability level). This probability level is defined taking into account design, construction, inspection and maintenance criteria as well as periodic controls; therefore the quality level and consequently the reliability level are thus also defined. The analysis is developed in the following stages: (1) definition of the analysis object (top event) and drawing of the relative fault-tree; (2) loop design analysis and preliminary optimization based on logic criteria; (3) classification into categories of the fault-tree primary events; (4) final loop design analysis and optimization based on defined component quality requirements. Stages 2 and 4 are quite different since stage 2 mainly consists of a redundance optimization, while stage 4 acts on the component quality level in such a way that each minimum cut-set leading to the top has an acceptable probability level. During analysis development, use is made of computer codes which, among other things enable the verification of fault-tree logic makeup, the listing of the minimum cut-sets with and without event categorization, and the evaluation of each cut-set order. (author)

  9. Assessment of residual life of fast breeder test reactor

    International Nuclear Information System (INIS)

    Srinivasan, G.

    2016-01-01

    The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor and has been in operation since 1985. As a part of regulatory requirement for relicensing, residual life assessment had to be carried out. The systems are made of SS 316, and designed for creep and fatigue. The design life for creep is 100,000 h at 550°C. The design fatigue cycle for operation from shutdown to full power varies from component to component. In general, most of the components are designed for 2000 cycles. The reactor has operated mostly below the design temperatures. It is seen that enough creep-fatigue life is available for the non-replaceable, permanent components. The residual life was found to be governed by the residual ductility of the Grid Plate supporting the core after neutron irradiation. Fast flux measurements were carried out at the grid plate location. Samples were irradiated and tensile tested. Results indicate the allowable dpa for a 10% residual ductility criterion as 4.37. This gave a residual life of ~ 6 Effective Full Power Years for the reactor as of Feb 2012. Measures to reduce the neutron dose on the grid plate are being taken. (author)

  10. Overview of the NKS/RAK-1 project 'Strategies for reactor safety' and linkages to piping reliability studies

    International Nuclear Information System (INIS)

    Andersson, Kjell

    1997-01-01

    The NKS/RAK-1 project forms part of a four-year research program (1994-97) in the Nordic countries. The general objective of NKS/RAK-1 project is to explore strategies for reactor safety: to investigate and evaluate the safety work, to increase realism and reliability of safety analysis; and to increase the safety of nuclear installations in selected areas. The project has done extensive interview work at utilities and authorities, and analysed a number of case studies. Brief highlights and overviews of the sub-projects are presented in this paper

  11. Assessment and Improving Methods of Reliability Indices in Bakhtar Regional Electricity Company

    Directory of Open Access Journals (Sweden)

    Saeed Shahrezaei

    2013-04-01

    Full Text Available Reliability of a system is the ability of a system to do prospected duties in future and the probability of desirable operation for doing predetermined duties. Power system elements failures data are the main data of reliability assessment in the network. Determining antiseptic parameters is the goal of reliability assessment by using system history data. These parameters help to recognize week points of the system. In other words, the goal of reliability assessment is operation improving and decreasing of the failures and power outages. This paper is developed to assess reliability indices of Bakhtar Regional Electricity Company up to 1393 and the improving methods and their effects on the reliability indices in this network. DIgSILENT Power Factory software is employed for simulation. Simulation results show the positive effect of improving methods in reliability indices of Bakhtar Regional Electricity Company.

  12. High-temperature gas-cooled reactor safety-reliability program plan

    Energy Technology Data Exchange (ETDEWEB)

    1981-03-01

    The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

  13. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  14. Reliability and testing considerations in the design of nuclear reactor filtration systems

    International Nuclear Information System (INIS)

    O'Nan, A.; Williams, R.P.; Goldsmith, J.M.

    1975-01-01

    The high performance standards set by USAEC-DRL Regulatory Guides for nuclear reactor filtration systems pose difficult problems for on-site leakage tests. These problems are compounded by the crowded conditions inside reactor structures, and by the fact that, until recently, little consideration has been given by system designers to the needs of testing. Techniques for coping with testing problems on existing systems, and suggestions for improving the testability of future systems, are given. Test crew safety considations are discussed, and a pair of easily portable contaminant generators is described. (U.S.)

  15. Reliability assessment of Port Harcourt 33/11kv Distribution System ...

    African Journals Online (AJOL)

    This makes reliability studies an important task besides all the other analyses required for assessing the system performance. The paper presents an analytical approach in the reliability assessment of the Port Harcourt 33/11kV power distribution system. The assessment was performed with the 2009 power outage data ...

  16. Assessment of the thorium fuel cycle in power reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

  17. Reliability of the Bulb Dynamometer for Assessing Grip Strength

    Directory of Open Access Journals (Sweden)

    Colleen Maher

    2018-04-01

    Full Text Available Background: Hand function is an overall indicator of health and is often measured using grip strength. Handheld dynamometry is the most common method of measuring grip strength. The purpose of this study was to determine the inter-rater and test-retest reliability, the reliability of one trial versus three trials, and the preliminary norms for a young adult population using the Baseline® Pneumatic Squeeze Bulb Dynamometer (30 psi. Methods: This study used a one-group methodological design. One hundred and three healthy adults (30 males and 73 females were recruited. Six measurements were collected for each hand per participant. The data was analyzed using Intraclass Correlation Coefficients (ICC two-way effects model (2,2 and paired-samples t-tests. Results: The ICC for inter-rater reliability ranged from 0.955 to 0.977. Conclusion: The results of this study suggest that the bulb dynamometer is a reliable tool to measure grip strength and should be further explored for reliable and valid use in diverse populations and as an alternative to the Jamar dynamometer.

  18. Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

  19. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    International Nuclear Information System (INIS)

    Woo, H.H.; Lu, S.C.

    1981-01-01

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

  20. Assessment of very high-temperature reactors in process applications

    International Nuclear Information System (INIS)

    Spiewak, I.; Jones, J.E. Jr.; Gambill, W.R.; Fox, E.C.

    1976-11-01

    An overview is presented of the technical and economic feasibility for the development of a very high-temperature reactor (VHTR) and associated processes. A critical evaluation of VHTR technology for process temperatures of 1400 and 2000 0 F is made. Additionally, an assessment of potential market impact is made to determine the commercial viability of the reactor system. It is concluded that VHTR process heat in the range of 1400 to 1500 0 F is attainable with near-term technology. However, process heat in excess of 1600 0 F would require considerably more materials development. The potential for the VHTR could include a major contribution to synthetic fuel, hydrogen, steel, and fertilizer production and to systems for transport and storage of high-temperature heat. A recommended development program including projected costs is presented

  1. Environmental assessment for the deactivation of the N Reactor facilities

    International Nuclear Information System (INIS)

    1995-05-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. As required by the National Environmental Policy Act of 1969 (NEPA), this EA complies with Title 40, Code of Federal Regulations (CFR), parts 1500--1508, ''Regulations for Implementing the Procedural Provisions of NEPA. '' It also implements the ''National Environmental Policy Act; Implementing Procedures and Guidelines'' (10 CFR 1021)

  2. Reactor safeguards system assessment and design. Volume I

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.; Bennett, H.A.; Hulme, B.L.

    1978-06-01

    This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore, should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems

  3. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  4. A clinical assessment tool used for physiotherapy students--is it reliable?

    Science.gov (United States)

    Lewis, Lucy K; Stiller, Kathy; Hardy, Frances

    2008-01-01

    Educational institutions providing professional programs such as physiotherapy must provide high-quality student assessment procedures. To ensure that assessment is consistent, assessment tools should have an acceptable level of reliability. There is a paucity of research evaluating the reliability of clinical assessment tools used for physiotherapy students. This study evaluated the inter- and intrarater reliability of an assessment tool used for physiotherapy students during a clinical placement. Five clinical educators and one academic participated in the study. Each rater independently marked 22 student written assessments that had been completed by students after viewing a videotaped patient physiotherapy assessment. The raters repeated the marking process 7 weeks later, with the assessments provided in a randomised order. The interrater reliability (Intraclass Correlation Coefficient) for the total scores was 0.32, representing a poor level of reliability. A high level of intrarater reliability (percentage agreement) was found for the clinical educators, with a difference in section scores of one mark or less on 93.4% of occasions. Further research should be undertaken to reevaluate the reliability of this clinical assessment tool following training. The reliability of clinical assessment tools used in other areas of physiotherapy education should be formally measured rather than assumed.

  5. Component reliability data for use in probabilistic safety assessment

    International Nuclear Information System (INIS)

    1988-10-01

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

  6. Training Raters to Assess Adult ADHD: Reliability of Ratings

    Science.gov (United States)

    Adler, Lenard A.; Spencer, Thomas; Faraone, Stephen V.; Reimherr, Fred W.; Kelsey, Douglas; Michelson, David; Biederman, Joseph

    2005-01-01

    The standardization of ADHD ratings in adults is important given their differing symptom presentation. The authors investigated the agreement and reliability of rater standardization in a large-scale trial of atomoxetine in adults with ADHD. Training of 91 raters for the investigator-administered ADHD Rating Scale (ADHDRS-IV-Inv) occurred prior to…

  7. Re-assessing reliability based on survived loads

    NARCIS (Netherlands)

    Schweckendiek, T.

    2011-01-01

    The reliability of flood defenses is often dictated by large uncertainties in the hydraulic loading and the structural resistance. Additional information decreases uncertainty, however, acquiring it is often costly. One source of information, even though in many cases readily available, is hardly

  8. Assessing Assertion: An Investigation of Construct Validity and Reliability.

    Science.gov (United States)

    Kern, Jeffrey M.; MacDonald, Marian L.

    The reliability and meaning of assertiveness tests were explored using 120 female undergraduates. Several self-report inventories (the College Self-Expression Scale, Conflict Resolution Inventory, and a global rating from one to seven) were administered, as were three anxiety measures (Timed Behavior Checklist, response latency, and response…

  9. Pattern Recognition for Reliability Assessment of Water Distribution Networks

    NARCIS (Netherlands)

    Trifunovi?, N.

    2012-01-01

    The study presented in this manuscript investigates the patterns that describe reliability of water distribution networks focusing to the node connectivity, energy balance, and economics of construction, operation and maintenance. A number of measures to evaluate the network resilience has been

  10. Assessment of the Reliability Profiles for Concrete Bridges

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle

    In this paper calculation of reliability profiles is discussed. ULS as well as SLS limit states are formulated. Corrosion due to chloride penetration is the considered deterioration mechanism. Three models for corrosion are formulated. A definition of service lifetime for concrete bridges...

  11. Reliability Assessment and Reliability-Based Inspection and Maintenance of Offshore Wind Turbines

    DEFF Research Database (Denmark)

    Ramirez, José Rangel

    actions are the most relevant and effective means of control of deterioration. The risk-based inspection planning methodology, based on Bayesian decision theory, represents an important tool to identify the suitable strategy to inspect and control the deterioration in structures such as offshore wind...... performance during the life cycle. The deterioration processes, such as fatigue and corrosion, are typically affecting offshore structural systems. This damage decreases the system performance and increases the risk of failure, thus not fulfilling the established safety criteria. Inspection and maintenance...... to their offshore location, no pollution risks and low human risks since they are unmanned. This allows the allocation of lower reliability level compared to e.g. oil & gas installations. With the incursion to water depths between 20 and 50 meters, the use of jacket and tripod structures represents a feasible...

  12. Technical report on reliability assessment of the digital I and C equipment using reliability block diagram

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Young; Hwang, In Koo; Cha, Kyung Ho; Park, Joo Hyun; Choi, Jong Gyun; Lee, Ki Young; Park, Jong Kyun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    As the operational life-time of nuclear power plants being currently developed is increasing with technology improvements, the life-time of instrumentation and control systems is also required to be met to this plant operational period. The electronic devices are usually expected to be available less than for 10 years. Therefore, an appropriate S and PM (surveillance and preventive maintenance) strategy should be established to secure the reliability and availability of the plants. EPRI-URD requires that the mean time between failures caused by any I and C system failure shall be more than 50 operating years. The methodology to access of the mean time between failures of I and C system is included in this report. 35 refs., 19 figs., 3 tabs. (Author)

  13. A view of technology maturity assessment to realize fusion reactor by Japanese young researchers

    International Nuclear Information System (INIS)

    Kasada, Ryuta; Goto, Takuya; Miyazawa, Junichi; Fujioka, Shinsuke; Hiwatari, Ryoji; Oyama, Naoyuki; Tanigawa, Hiroyasu

    2013-01-01

    Japanese young researchers who have interest in realizing fusion reactor have analyzed Technology Readiness Levels (TRL) in Young Scientists Special Interest Group on Fusion Reactor Realization. In this report, brief introduction to TRL assessment and a view of TRL assessment against fusion reactor projects conducting in Japan. (J.P.N.)

  14. A reliability index for assessment of crack profile reconstructed from ECT signals using a neural-network approach

    International Nuclear Information System (INIS)

    Yusa, Noritaka; Chen, Zhenmao; Miya, Kenzo; Cheng, Weiying

    2002-01-01

    This paper proposes a reliability parameter to enhance an version scheme developed by authors. The scheme is based upon an artificial neural network that simulates mapping between eddy current signals and crack profiles. One of the biggest advantages of the scheme is that it can deal with conductive cracks, which is necessary to reconstruct natural cracks. However, it has one significant disadvantage: the reliability of reconstructed profiles was unknown. The parameter provides an index for assessment of the crack profile and overcomes this disadvantage. After the parameter is validated by reconstruction of simulated cracks, it is applied to reconstruction of natural cracks that occurred in steam generator tubes of a pressurized water reactor. It is revealed that the parameter is applicable to not only simulated cracks but also natural ones. (author)

  15. Fast Reliability Assessing Method for Distribution Network with Distributed Renewable Energy Generation

    Science.gov (United States)

    Chen, Fan; Huang, Shaoxiong; Ding, Jinjin; Ding, Jinjin; Gao, Bo; Xie, Yuguang; Wang, Xiaoming

    2018-01-01

    This paper proposes a fast reliability assessing method for distribution grid with distributed renewable energy generation. First, the Weibull distribution and the Beta distribution are used to describe the probability distribution characteristics of wind speed and solar irradiance respectively, and the models of wind farm, solar park and local load are built for reliability assessment. Then based on power system production cost simulation probability discretization and linearization power flow, a optimal power flow objected with minimum cost of conventional power generation is to be resolved. Thus a reliability assessment for distribution grid is implemented fast and accurately. The Loss Of Load Probability (LOLP) and Expected Energy Not Supplied (EENS) are selected as the reliability index, a simulation for IEEE RBTS BUS6 system in MATLAB indicates that the fast reliability assessing method calculates the reliability index much faster with the accuracy ensured when compared with Monte Carlo method.

  16. Improvement in the reliability of shells for light water reactors by manufacture from hollow ingots

    International Nuclear Information System (INIS)

    Bocquet, P.; Blondeau, R.; Poitrault, I.; Badeau, J.P.; Dumont, R.

    1991-01-01

    The problems associated to the segregation located at the inner surface and subsurface of heavy shell forgings used in nuclear light water reactors are presented. The effect of A segregation on cold or reheat cracking HAZ has conducted fabricators to use severe welding procedures with high preheat temperature and refining HAZ grain size sequences. For shells and rings, the hollow ingot as developed by CLI, with a good control of the location of the center line of segregation, is an excellent answer to these problems. The use of core shell forgings issued from this type of hollow ingots, free of segregation at the inner surface and subsurface area, contributes, by reducing the irradiation embrittlement effect, to increase the safety factor relatively to the risk of failure of the reactor. (orig.)

  17. Generation reliability assessment in oligopoly power market using Monte Carlo simulation

    International Nuclear Information System (INIS)

    Haroonabadi, H.; Haghifam, M.R.

    2007-01-01

    This paper addressed issues regarding power generation reliability assessment (HLI) in deregulated power pool markets. Most HLI reliability evaluation methods are based on the loss of load (LOLE) approach which is among the most suitable indices to describe the level of generation reliability. LOLE refers to the time in which load is greater than the amount of available generation. While most reliability assessments deal only with power system constraints, this study considered HLI reliability assessment in an oligopoly power market using Monte Carlo simulation (MCS). It evaluated the sensitivity of the reliability index to different reserve margins and future margins. The reliability index was determined by intersecting the offer and demand curves of power plants and comparing them to other parameters. The paper described the fundamentals of an oligopoly power pool market and proposed an algorithm for HLI reliability assessment for such a market. The proposed method was assessed on the IEEE-Reliability Test System with satisfactory results. In all cases, generation reliability indices were evaluated with different reserve margins and various load levels. 19 refs., 7 figs., 1 appendix

  18. Some aspects of statistic evaluation of fast reactor fuel element reliability

    International Nuclear Information System (INIS)

    Proshkin, A.A.; Likhachev, Yu.I.; Tuzov, A.N.; Zabud'ko, L.M.

    1980-01-01

    Certain aspects of application of statistical methods in forecasting operating ability of fuel elements of fast reactors with liquid-metal-heat-carriers are considered. Results of statistical analysis of fuel element operating ability with oxide fuel (U, Pu)O 2 under stationary regime of fast power reactor capacity are given. The analysis carried out permits to single out the main parameters, considerably affecting the calculated determination of fuel element operating ability. It is shown that parameters which introduce the greatest uncertainty are: steel creep rate - up to 30%; steel swelling - up to 20%; fuel ceep rate - up to 30%, fuel swelling - up to 20%, the coating material corrosion - up to 15%; contact conductivity of the fuel-coating gap - up to 10%. Contribution of these parameters in every given case is different depending on the construction, operation conditions and fuel element cross section considered. Contribution of the coating temperature uncertainty to the total dispersion does not exceed several per cent. It is shown that for the given reactor operation conditions the number of fuel elements depressurized increases with the burn out almost exponentially, starting from the burn out higher than 7% of heavy atoms

  19. Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

    2006-07-01

    A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

  20. Self Assessment for the Safety of Research Reactor in Indonesia

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2008-01-01

    At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. National Nuclear Energy Agency (BATAN) has three research reactor; which are: Reactor Triga Mark II at Bandung, Reactor Kartini at Yogyakarta and Reactor Serbaguna (Multi Purpose Reactor) at Serpong. The Code of Conduct on the Safety of Research Reactors establishes 'best practice' guidelines for the licensing, construction and operation of research reactors. In this paper the author use the requirement in code of conduct to review the safety of research reactor in Indonesia

  1. Reliability assessment of the containment of a PWR

    International Nuclear Information System (INIS)

    Schueller, G.I.; Wellein, R.; Wittmann, F.H.; Boulahdour, T.; Mihashi, M.; Zorn, N.F.; Bauer, J.

    1981-09-01

    The aim of this research effort was to contribute to the development of methods to quantify the risk involved with nuclear power plants. Using a large component, i.e. the containment of the reference plant BIBLIS B (PWR) as sample structure a reliability analysis was performed which is based on realistic assumptions of loads and material properties. For this purpose in many fields it was necessary to develop new methods, collect data, and where not available, obtain data in tests. This effort concentrated on partial aspects and on the other hand on the development of a methodology for an overall reliability concept. According to the results of the previous project, the keypoints of this effort are the treatment of loss of coolant accidents (small leak), earthquake loading, the possibly resulting crackpropagation in the steel hull, and the structural mechanics and material strength aspects of the reinforced concrete hull subjected to impact loading (aircraft impact). (orig./HP) [de

  2. Assessment of teacher competence using video portfolios: reliability, construct validity and consequential validity

    NARCIS (Netherlands)

    Admiraal, W.; Hoeksma, M.; van de Kamp, M.-T.; van Duin, G.

    2011-01-01

    The richness and complexity of video portfolios endanger both the reliability and validity of the assessment of teacher competencies. In a post-graduate teacher education program, the assessment of video portfolios was evaluated for its reliability, construct validity, and consequential validity.

  3. Reliability and validity of a tool to assess airway management skills in anesthesia trainees

    Directory of Open Access Journals (Sweden)

    Aliya Ahmed

    2016-01-01

    Conclusion: The tool designed to assess bag-mask ventilation and tracheal intubation skills in anesthesia trainees demonstrated excellent inter-rater reliability, fair test-retest reliability, and good construct validity. The authors recommend its use for formative and summative assessment of junior anesthesia trainees.

  4. Assessment of very high temperature reactors in process applications

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

    1976-01-01

    In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

  5. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  6. Application of probabilistic safety assessment to research reactors

    International Nuclear Information System (INIS)

    1989-07-01

    This document has been prepared to assist in the performance of a research reactor probabilistic safety assessment (PSA). It offers examples of experience gained by a number of Member States in carrying out PSA for research reactors. These examples are illustrative of the types of approach adopted, the problems that arise and the judgements entered into when conducting a PSA. The illustrative examples of experiences gained are discussed in a series of thirteen chapters which address some of the issues that arise in a PSA. The examples are not exhaustive and offer evidence of how other analyses have approached the task of preparing a PSA, for their particular plant. The principles should be capable of being utilised and the various issues which are discussed should be translated into the needs of the analyst. Each PSA will make its own demands on the analyst depending on the reactor and so the illustrations must only be used as guidance and not adopted as published, without critical appreciation. Refs, figs and tabs

  7. Advancing methods for reliably assessing motivational interviewing fidelity using the motivational interviewing skills code.

    Science.gov (United States)

    Lord, Sarah Peregrine; Can, Doğan; Yi, Michael; Marin, Rebeca; Dunn, Christopher W; Imel, Zac E; Georgiou, Panayiotis; Narayanan, Shrikanth; Steyvers, Mark; Atkins, David C

    2015-02-01

    The current paper presents novel methods for collecting MISC data and accurately assessing reliability of behavior codes at the level of the utterance. The MISC 2.1 was used to rate MI interviews from five randomized trials targeting alcohol and drug use. Sessions were coded at the utterance-level. Utterance-based coding reliability was estimated using three methods and compared to traditional reliability estimates of session tallies. Session-level reliability was generally higher compared to reliability using utterance-based codes, suggesting that typical methods for MISC reliability may be biased. These novel methods in MI fidelity data collection and reliability assessment provided rich data for therapist feedback and further analyses. Beyond implications for fidelity coding, utterance-level coding schemes may elucidate important elements in the counselor-client interaction that could inform theories of change and the practice of MI. Copyright © 2015 Elsevier Inc. All rights reserved.

  8. RELIABILITY ASSESSMENT OF ENTROPY METHOD FOR SYSTEM CONSISTED OF IDENTICAL EXPONENTIAL UNITS

    Institute of Scientific and Technical Information of China (English)

    Sun Youchao; Shi Jun

    2004-01-01

    The reliability assessment of unit-system near two levels is the most important content in the reliability multi-level synthesis of complex systems. Introducing the information theory into system reliability assessment, using the addible characteristic of information quantity and the principle of equivalence of information quantity, an entropy method of data information conversion is presented for the system consisted of identical exponential units. The basic conversion formulae of entropy method of unit test data are derived based on the principle of information quantity equivalence. The general models of entropy method synthesis assessment for system reliability approximate lower limits are established according to the fundamental principle of the unit reliability assessment. The applications of the entropy method are discussed by way of practical examples. Compared with the traditional methods, the entropy method is found to be valid and practicable and the assessment results are very satisfactory.

  9. On the use of NDT Data for Reliability-Based Assessment of Existing Timber Structures

    DEFF Research Database (Denmark)

    Sousa, Hélder S.; Sørensen, John Dalsgaard; Kirkegaard, Poul Henning

    2013-01-01

    The objective of this paper is to address the possibilities of using non-destructive testing (NDT) data for updating information and obtaining adequate characterization of the reliability level of existing timber structures and, also, for assessing the evolution in time of performance...... of these structures when exposed to deterioration. By improving the knowledge upon the mechanical properties of timber, better and more substantiated decisions after a reliability safety assessment are aimed at. Bayesian methods are used to update the mechanical properties of timber and reliability assessment......, and information of NDT is also used to calibrate these models. The proposed approach is used for reliability assessment of different structural timber systems. Reliability of the structural system is assessed regarding the failure consequences of individual elements defined as key elements which were determined...

  10. "The Reliability, Validity, and Utility of Self-Assessment"

    Directory of Open Access Journals (Sweden)

    John A. Ross

    2006-11-01

    Full Text Available Despite widespread use of self-assessment, teachers have doubts about the value and accuracy of the technique. This article reviews research evidence on student self-assessment, finding that (1 self-assessment produces consistent results across items, tasks, and short time periods; (2 self-assessment provides information about student achievement that corresponds only in part to the information generated by teacher assessments; (3 self-assessment contributes to higher student achievement and improved behavior. The central finding of this review is that (4 the strengths of self-assessment can be enhanced through training students how to assess their work and each of the weaknesses of the approach (including inflation of grades can be reduced through teacher action.

  11. Quantitative evaluation of the impact of human reliability in risk assessment for nuclear power plants

    International Nuclear Information System (INIS)

    Samanta, P.K.

    1981-01-01

    The role of human beings in the safe operation of a nuclear power plant has been a matter of concern. This study describes methods for the quantitative description of that role and its impact on the risk from nuclear power plants. The impact of human errors was calculated by observing the changes in risk parameters, such as core melt probability, release category probabilities, accident sequence probabilities and system unavailabilities due to changes in the contribution to unavailablity of human errors, within the framework of risk assessment methodology. It was found that for operational pressurized water reactors the opportunity for reduction in core melt probability by reducing the human error rates without simultaneous reduction of hardware failures is limited, but that core melt probability would significantly increase as human error rates increased. More importantly, most of the dominant accident sequences showed a significant increase in their probabilities with an increase in human error rates. Release categories resulting in high consequences showed a much larger sensitivity to human errors than categories resulting in low consequences. A combination of structural importance and reliability importance measure was used to describe the importance of individual errors

  12. An overview of the evolution of human reliability analysis in the context of probabilistic risk assessment

    International Nuclear Information System (INIS)

    Bley, Dennis C.; Lois, Erasmia; Kolaczkowski, Alan M.; Forester, John Alan; Wreathall, John; Cooper, Susan E.

    2009-01-01

    Since the Reactor Safety Study in the early 1970's, human reliability analysis (HRA) has been evolving towards a better ability to account for the factors and conditions that can lead humans to take unsafe actions and thereby provide better estimates of the likelihood of human error for probabilistic risk assessments (PRAs). The purpose of this paper is to provide an overview of recent reviews of operational events and advances in the behavioral sciences that have impacted the evolution of HRA methods and contributed to improvements. The paper discusses the importance of human errors in complex human-technical systems, examines why humans contribute to accidents and unsafe conditions, and discusses how lessons learned over the years have changed the perspective and approach for modeling human behavior in PRAs of complicated domains such as nuclear power plants. It is argued that it has become increasingly more important to understand and model the more cognitive aspects of human performance and to address the broader range of factors that have been shown to influence human performance in complex domains. The paper concludes by addressing the current ability of HRA to adequately predict human failure events and their likelihood

  13. An overview of the evolution of human reliability analysis in the context of probabilistic risk assessment.

    Energy Technology Data Exchange (ETDEWEB)

    Bley, Dennis C. (Buttonwood Consulting Inc., Oakton, VA); Lois, Erasmia (U.S. Nuclear Regulatory Commission, Washington, DC); Kolaczkowski, Alan M. (Science Applications International Corporation, Eugene, OR); Forester, John Alan; Wreathall, John (John Wreathall and Co., Dublin, OH); Cooper, Susan E. (U.S. Nuclear Regulatory Commission, Washington, DC)

    2009-01-01

    Since the Reactor Safety Study in the early 1970's, human reliability analysis (HRA) has been evolving towards a better ability to account for the factors and conditions that can lead humans to take unsafe actions and thereby provide better estimates of the likelihood of human error for probabilistic risk assessments (PRAs). The purpose of this paper is to provide an overview of recent reviews of operational events and advances in the behavioral sciences that have impacted the evolution of HRA methods and contributed to improvements. The paper discusses the importance of human errors in complex human-technical systems, examines why humans contribute to accidents and unsafe conditions, and discusses how lessons learned over the years have changed the perspective and approach for modeling human behavior in PRAs of complicated domains such as nuclear power plants. It is argued that it has become increasingly more important to understand and model the more cognitive aspects of human performance and to address the broader range of factors that have been shown to influence human performance in complex domains. The paper concludes by addressing the current ability of HRA to adequately predict human failure events and their likelihood.

  14. Assessment of nucleonic methods and data for fusion reactors

    International Nuclear Information System (INIS)

    Dudziak, D.J.

    1976-01-01

    An assessment is provided of nucleonic methods, codes, and data necessary for a sound experimental fusion power reactor (EPR) technology base. Gaps in the base are identified and specific development recommendations are made in three areas: computational tools, nuclear data, and integral experiments. The current status of the first two areas is found to be sufficiently inadequate that viable engineering design of an EPR is precluded at this time. However, a program to provide the necessary data and computational capability is judged to be a low-risk effort

  15. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

  16. Using the Hemophilia Joint Health Score for assessment of children: Reliability of the Spanish version.

    Science.gov (United States)

    R, Cuesta-Barriuso; A, Torres-Ortuño; S, Pérez-Alenda; J, Carrasco Juan; F, Querol; J, Nieto-Munuera; Ja, López-Pina

    2018-02-27

    Numerous measuring instruments for the evaluation of hemophilic arthropathy have been developed. One of the most used systems is the Hemophilia Joint Health Score (HJHS) given its sensitivity to clinical changes appearing in the joints because of recurrent hemarthrosis. Assessing the interrater reliability, using the Spanish version of the HJHS (version 2.1) in children with hemophilia. Reliability study to assess the interrater reliability of the Spanish version of HJHS. A sample of 36 children aged 7-13 years diagnosed with hemophilia A or B was used. Two physiotherapists performed physical assessments with the Spanish version of the HJHS. Descriptive statistics (range, mean, standard deviation) and the analysis of interrater reliability were calculated. The interrater reliability was heterogeneous since the Kappa coefficient range (ĸ), although significant (p reliability of the Spanish population version of the HJHS is high. This scale should be used generically in evaluating musculoskeletal pediatric patients with hemophilia.

  17. Reliability of fast reactor mixed-oxide fuel during operational transients

    International Nuclear Information System (INIS)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  18. Reliability assessment based on small samples of normal distribution

    International Nuclear Information System (INIS)

    Ma Zhibo; Zhu Jianshi; Xu Naixin

    2003-01-01

    When the pertinent parameter involved in reliability definition complies with normal distribution, the conjugate prior of its distributing parameters (μ, h) is of normal-gamma distribution. With the help of maximum entropy and the moments-equivalence principles, the subjective information of the parameter and the sampling data of its independent variables are transformed to a Bayesian prior of (μ,h). The desired estimates are obtained from either the prior or the posterior which is formed by combining the prior and sampling data. Computing methods are described and examples are presented to give demonstrations

  19. Integration of the functional reliability of two passive safety systems to mitigate a SBLOCA+BO in a CAREM-like reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    Mezio, Federico, E-mail: federico.mezio@cab.cnea.gov.ar [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Grinberg, Mariela [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina); Lorenzo, Gabriel [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Giménez, Marcelo [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina)

    2014-04-01

    Highlights: • An estimation of the Functional Unreliability was performed using RMPS methodology. • The methodology uses an improved response surface in order to estimate the FU. • The FU may become relevant to be analyzed in the Passive Safety Systems. • There were proposed two ways to incorporate the FU into an APS. - Abstract: This paper describes a case study of a methodological approach for assessing the functional reliability of passive safety systems (PSS) and its treatment within a probabilistic safety assessment (PSA). The functional unreliability (FU) can be understood as the failure probability of PSS to fulfill its mission due to the impairment of the related passive safety function. The safety function accomplishment is characterized and quantified by a performance indicator (PI), which is a measure of how far the system is from verifying its mission. PI uncertainties are estimated from uncertainty propagation of selected parameters. A methodology based on the reliability methodology for passive system (RMPS) one is used to estimate the FU associated to the isolation condensers (ICs) in combination with the accumulators (medium pressure injection system) of a CAREM-like integral advanced reactor. A small break loss of coolant accident with black-out is selected as an evaluation case. This implies success of reactor shut-down (inherent) and failure of residual heat removal by active systems. The safety function to accomplish is to refill the reactor pressure vessel (RPV) in order to avoid core damage. For this case, to allow the discharge of accumulators into RPV, the pressure must be reduced by the IC. The methodology for passive safety function assessment considers uncertainties in code parameters, besides uncertainties in engineering parameters (design, construction, operation and maintenance), in order to perform Monte Carlo simulations based on best estimate (B-E) plant model. Then, response surfaces based on PI are used for improving the

  20. Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

    International Nuclear Information System (INIS)

    Azpitarte, O.; Ramilo, L.

    2013-01-01

    The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)