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

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

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.

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

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

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

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

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.

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.

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

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

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

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.

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

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

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.

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)

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

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

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

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)

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)

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)

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

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

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)

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

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.

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

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)

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.

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

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

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

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

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

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)

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

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)

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

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

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

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

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.)

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

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.)

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

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.

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

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

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

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

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

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.

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.

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.

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

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

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

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

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

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

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

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

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

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)

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

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

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

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

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

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

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

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)

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)

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

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

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)

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

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

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.

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?)

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)

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

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.)

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)

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

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)

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)

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)

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

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

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

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

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)

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)

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.

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

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

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)

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

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)

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

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

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.

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)

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

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)

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

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)

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

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.

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)

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

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

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

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)

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

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)

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

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.

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

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)

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)

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

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.

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

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...

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

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)

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)

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)

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)

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

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

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

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

Development of a methodology for conducting an integrated HRA/PRA --. Task 1, An assessment of human reliability influences during LP&S conditions PWRs

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. [Brookhaven National Lab., Upton, NY (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Cooper, S.E. [Science Applications International Corp., McLean, VA (United States)

1993-06-01

During Low Power and Shutdown (LP&S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant`s systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP&S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP&S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP&S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP&S, (2) identification of potentially important LP&S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP&S conditions for a pressurized water reactor (PWR).

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

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

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

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

Ergonomics design and operator training as contributors to human reliability

International Nuclear Information System (INIS)

Jackson, A.R.G.; Madden, V.J.; Umbers, I.G.; Williams, J.C.

1987-01-01

The safe operation of nuclear reactors depends not only on good physical safety engineering but on the human operators as well. The Central Electricity Generating Board's approach to human reliability includes the following aspects: ergonomics design (task analysis and the development of man-machine interfaces), analysis of human reliability, operational feedback, staff training and assessment, maintenance management, research programmes and management. This paper describes how these combine to achieve the highest practicable level of human reliability, not only for the Sizewell-B pressurized water reactor, but also for the Board's gas-cooled reactors. Examples are used to illustrate the topics considered. (UK)

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.

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...

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

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

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.

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.)

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.)

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

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.

Survey of industry methods for producing highly reliable software

International Nuclear Information System (INIS)

Lawrence, J.D.; Persons, W.L.

1994-11-01

The Nuclear Reactor Regulation Office of the US Nuclear Regulatory Commission is charged with assessing the safety of new instrument and control designs for nuclear power plants which may use computer-based reactor protection systems. Lawrence Livermore National Laboratory has evaluated the latest techniques in software reliability for measurement, estimation, error detection, and prediction that can be used during the software life cycle as a means of risk assessment for reactor protection systems. One aspect of this task has been a survey of the software industry to collect information to help identify the design factors used to improve the reliability and safety of software. The intent was to discover what practices really work in industry and what design factors are used by industry to achieve highly reliable software. The results of the survey are documented in this report. Three companies participated in the survey: Computer Sciences Corporation, International Business Machines (Federal Systems Company), and TRW. Discussions were also held with NASA Software Engineering Lab/University of Maryland/CSC, and the AIAA Software Reliability Project

Proceedings of the international topical meeting on advanced reactors safety: Volume 2

International Nuclear Information System (INIS)

1997-01-01

In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

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

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.

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

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

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)

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

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

Proceedings of the workshop on reliability data collection

International Nuclear Information System (INIS)

1999-01-01

The main purpose of the Workshop was to provide a forum for exchanging information and experience on Reliability Data Collection and analysis to support Living Probabilistic Safety Assessments (LPSA). The Workshop is divided into four sessions which titles are: Session 1: Reliability Data - Database Systems (3 papers), Session 2: Reliability Data Collection for PSA (5 papers), Session 3: NPP Data Collection (3 papers), Session 4: Reliability Data Assessment (Part 1: General - 2 papers; Part 2: CCF - 2 papers; Part 3: Reactor Protection Systems / External Event Data - 2 papers; Part 4: Human Errors - 2 papers)

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.

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).

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

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.

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

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

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

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

Fundamentals and applications of systems reliability analysis

International Nuclear Information System (INIS)

Boesebeck, K.; Heuser, F.W.; Kotthoff, K.

1976-01-01

The lecture gives a survey on the application of methods of reliability analysis to assess the safety of nuclear power plants. Possible statements of reliability analysis in connection with specifications of the atomic licensing procedure are especially dealt with. Existing specifications of safety criteria are additionally discussed with the help of reliability analysis by the example of the reliability analysis of a reactor protection system. Beyond the limited application to single safety systems, the significance of reliability analysis for a closed risk concept is explained in the last part of the lecture. (orig./LH) [de

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.

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

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.

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

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.

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...

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)

Reliability and safety of nuclear power stations

International Nuclear Information System (INIS)

Stepanek, S.

1979-01-01

The main problems are briefly discussed associated with the assessment of the safety and reliability of reactor pressure vessels. Two approaches are being applied to the assessment: one is based on the crack arrest temperature, the other on the determination of conditions corresponding to brittle fracture formation and on the determination of the critical defect size. The importance is stressed of continuous in-service inspection which may increase the factor of reliability by up to 10 4 times. (Z.M.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1990-02-01

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

Local AREA networks in advanced nuclear reactors

International Nuclear Information System (INIS)

Bicknell, J.; Keats, A.B.

1984-01-01

The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

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

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

The development on the methodology of the initiating event frequencies for liquid metal reactor KALIMER

International Nuclear Information System (INIS)

Jeong, K. S.; Yang, Z. A.; Ah, Y. B.; Jang, W. P.; Jeong, H. Y.; Ha, K. S.; Han, D. H.

2002-01-01

In this paper, the PSA methodology of PRISM,Light Water Reactor, Pressurized Heavy Water Reactor are analyzed and the methodology of Initiating Events for KALIMER are suggested. Also,the reliability assessment of assumptions for Pipes Corrosion Frequency is set up. The reliability assessment of Passive Safety System, one of Main Safety System of KALIMER, are discussed and analyzed

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

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.

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

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.

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)

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)

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)

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

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

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.

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

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

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.

Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

1999-05-15

The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

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

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

Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

Science.gov (United States)

Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

2017-01-01

An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are 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. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

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

Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

International Nuclear Information System (INIS)

Jordanov, M.

1999-01-01

The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

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.

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

Safety of RBMK reactors: Major results and prospects

International Nuclear Information System (INIS)

Sidorenko, V.A.

1996-01-01

The paper considers the following issues: basic reasons for the advent of NPPs with RBMK reactors; the logic of identifying top-priority measures immediately after the accident; top-priority measures for improving the safety and reliability of NPPs with RBMK reactors; upgrading NPPs with RBMK reactors in compliance with the Norms; programmes for retrofitting and upgrading of NPPs of the ''Rosnergoatom'' Concern and progress with their implementation as of April 1996; the safety of RBMK plants and the programmes of its enhancement with regard to modern requirements in the light of national and international assessment; objective indicators of safety, reliability, and economic efficiency of NPPs with RBMK reactors; economics: rationale for continuing plants operation till the end of their design lifetime. 8 refs, 3 figs

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)

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

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.)

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)

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)

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

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.)

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

Research and development into power reactor fuel performance

International Nuclear Information System (INIS)

Notley, M.J.F.

1983-07-01

The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

Thermodynamic characterization of salt components for Molten Salt Reactor fuel

NARCIS (Netherlands)

Capelli, E.

2016-01-01

The Molten Salt Reactor (MSR) is a promising future nuclear fission reactor technology with excellent performance in terms of safety and reliability, sustainability, proliferation resistance and economics. For the design and safety assessment of this concept, it is extremely important to have a

Reliability assessment of shut-off rod drive mechanism for TAPP - 3 and 4 and critical facility through life cycle testing

International Nuclear Information System (INIS)

Singh, Manjit; Badodkar, D.N.; Singh, N.K.; Dalal, N.S.; Mishra, M.K.; Veda Vyas, G.; Kothari, C.B.; Rao, V.V.S.S.; Saraf, R.K.

2006-01-01

Shut-off rod drive mechanism forms a safety critical system of a nuclear reactor. It is the space constraints for the given reactor layout, which makes design of shut-off rod drive mechanism (SRDM) a custom built design. Design of SRDM adopts fail-safe, replaceability and the simplicity criterion ensuring very high reliability of its operation. Shut-off rod drive mechanism for TAPP-3 and 4 and 'Critical Facility' have been recently designed and developed at Division of Remote Handling and Robotics (DRHR), BARC. These are designed with a number of advanced features and these are significantly different than those used in Dhruva and 220 MWe PHWRs. Design of SRDM is qualified through proto typing and life cycle testing on a full-scale test set-up. This paper gives details of qualification and life cycle test data for prototype SRDM for TAPP-3 and 4 and 'Critical Facility' and reliability assessment. (author)

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

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

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

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]....

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

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

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

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...

Risk-based management system development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Davis, M.L.; Eide, S.A.

1990-01-01

A Risk-Based Management System (RBMS) is being developed to facilitate the use of the Advanced Test Reactor (ATR) probabilistic risk assessment to support ATR operation. Most ATR RBMS questions can best be answered using the System Analysis and Risk Assessment System (SARA) developed at the Idaho National Engineering Laboratory. However, some applications may require employment of the other four codes used to develop and report the PRA. These four codes include the Integrated Reliability and Risk Analysis System (IRRAS), SETS, ETA-II, and the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The ATR RBMS will evolve over three years, and will include the results of the Level 3 and external events analysis

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)

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.

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

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

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

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

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

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

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

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

Design and reliability assessment of control systems for a nuclear-based hydrogen production plant with copper-chlorine thermochemical cycle

Energy Technology Data Exchange (ETDEWEB)

Al-Dabbagh, Ahmad W. [Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4 (Canada); Lu, Lixuan [Faculty of Energy Systems and Nuclear Science, Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4 (Canada)

2010-02-15

The thermochemical Copper-Chlorine (Cu-Cl) cycle is an emerging new method of nuclear-based hydrogen production. In the process, water is decomposed into hydrogen and oxygen through several physical and chemical processes. In this paper, a Distributed Control System (DCS) is designed for the thermochemical Cu-Cl cycle. The architecture and the communication networks of the DCS are discussed. Reliability of the DCS is assessed using fault trees. In the assessment, the impact of the malfunction of the actuators, sensors, controllers and communication networks on the overall system reliability is investigated. This provides key information for the selection of control system components, and determination of their inspection frequency and maintenance strategy. The hydrogen reactor unit, which is one of the major components in the thermochemical Cu-Cl cycle, is used to demonstrate the detailed design and analysis. (author)

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.

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

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

Issues in cognitive reliability

International Nuclear Information System (INIS)

Woods, D.D.; Hitchler, M.J.; Rumancik, J.A.

1984-01-01

This chapter examines some problems in current methods to assess reactor operator reliability at cognitive tasks and discusses new approaches to solve these problems. The two types of human failures are errors in the execution of an intention and errors in the formation/selection of an intention. Topics considered include the types of description, error correction, cognitive performance and response time, the speed-accuracy tradeoff function, function based task analysis, and cognitive task analysis. One problem of human reliability analysis (HRA) techniques in general is the question of what are the units of behavior whose reliability are to be determined. A second problem for HRA is that people often detect and correct their errors. The use of function based analysis, which maps the problem space for plant control, is recommended

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)

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...

Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

International Nuclear Information System (INIS)

Tong Jiejuan; Zhao Jun; Liu Tao; Xue Dazhi

2012-01-01

Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited

Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

International Nuclear Information System (INIS)

Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

1980-01-01

The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

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

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.

Conceptual design of high resolution and reliable density measurement system on helical reactor FFHR-d1 and demonstration on LHD

International Nuclear Information System (INIS)

Akiyama, T.; Yasuhara, R.; Isobe, M.; Sakamoto, R.; Goto, T.; Kawahata, K.; Sagara, A.; Nakayama, K.; Okajima, S.

2014-10-01

This paper describes a conceptual design of the density measurement system on the helical reactor FFHR-d1 based on its quantitative operation scenario. The density measurement is required to meet the reactor design, and to have a high density resolution of the order of 10 17 m -3 with a time resolution of 10 ms and high reliability (no fringe jump). “A dispersion interferometer” is designed and a prototype is tested and installed on LHD, which can realize a demo relevant density plasma. The prototype demonstrates the feasibility on a demo reactor. (author)

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)

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

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)

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

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.

Safety regulations concerning instrumentation and control systems for research reactors

International Nuclear Information System (INIS)

El-Shanshoury, A.I.

2009-01-01

A brief study on the safety and reliability issues related to instrumentation and control systems in nuclear reactor plants is performed. In response, technical and strategic issues are used to accomplish instrumentation and control systems safety. For technical issues there are ; systems aspects of digital I and C technology, software quality assurance, common-mode software, failure potential, safety and reliability assessment methods, and human factors and human machine interfaces. The strategic issues are the case-by-case licensing process and the adequacy of the technical infrastructure. The purpose of this work was to review the reliability of the safety systems related to these technical issues for research reactors

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

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

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

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.

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.

Experimental research of fuel element reliability

International Nuclear Information System (INIS)

Cech, B.; Novak, J.; Chamrad, B.

1980-01-01

The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

International Nuclear Information System (INIS)

Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

1980-06-01

The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and 233 U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles

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.

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…

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.

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

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

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.

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.

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.

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 ...

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.

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)

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

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

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.

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

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.

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).

Thermochemical data for reactor materials and fission products

International Nuclear Information System (INIS)

Cordfunke, E.H.P.; Konings, R.J.M.

1990-01-01

This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

Reliability of the ECHOWS Tool for Assessment of Patient Interviewing Skills.

Science.gov (United States)

Boissonnault, Jill S; Evans, Kerrie; Tuttle, Neil; Hetzel, Scott J; Boissonnault, William G

2016-04-01

History taking is an important component of patient/client management. Assessment of student history-taking competency can be achieved via a standardized tool. The ECHOWS tool has been shown to be valid with modest intrarater reliability in a previous study but did not demonstrate sufficient power to definitively prove its stability. The purposes of this study were: (1) to assess the reliability of the ECHOWS tool for student assessment of patient interviewing skills and (2) to determine whether the tool discerns between novice and experienced skill levels. A reliability and construct validity assessment was conducted. Three faculty members from the United States and Australia scored videotaped histories from standardized patients taken by students and experienced clinicians from each of these countries. The tapes were scored twice, 3 to 6 weeks apart. Reliability was assessed using interclass correlation coefficients (ICCs) and repeated measures. Analysis of variance models assessed the ability of the tool to discern between novice and experienced skill levels. The ECHOWS tool showed excellent intrarater reliability (ICC [3,1]=.74-.89) and good interrater reliability (ICC [2,1]=.55) as a whole. The summary of performance (S) section showed poor interrater reliability (ICC [2,1]=.27). There was no statistical difference in performance on the tool between novice and experienced clinicians. A possible ceiling effect may occur when standardized patients are not coached to provide complex and obtuse responses to interviewer questions. Variation in familiarity with the ECHOWS tool and in use of the online training may have influenced scoring of the S section. The ECHOWS tool demonstrates excellent intrarater reliability and moderate interrater reliability. Sufficient training with the tool prior to student assessment is recommended. The S section must evolve in order to provide a more discerning measure of interviewing skills. © 2016 American Physical Therapy

Workshop on PSA for New and Advanced Reactors

International Nuclear Information System (INIS)

2012-01-01

This workshop was organized by the NEA Working Group on Risk Assessment (WGRISK). The key objective of the workshop was to share the current state-of-the art on the PSA (Probabilistic Safety Assessment) applied for new reactors and advanced reactors. Fifty experts from 13 countries and one international organization (IAEA) participated in the present workshop, and 35 technical papers were presented. The main topics of interest, discussed during the workshop, included the followings: regulatory aspects, risk-informed methods, technical aspects of the PSA for new and advanced reactors, hazards of PSA (internal and external), severe accident/source term/Level 2 PSA, and consequence analysis/Level 3 PSA. Among the technical aspects of the PSA, the assessment of the reliability of passive safety systems appears to be a recurrent issue

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)

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

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...

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2012-02-15

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

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.

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.)

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

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

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

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.)

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...

Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

INPRO economic assessment of the IRIS nuclear reactor for deployment in Brazil

Energy Technology Data Exchange (ETDEWEB)

Goncalves Filho, Orlando Joao Agostinho, E-mail: orlando@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN - RJ), Rua Helio de Almeida, 75, Cidade Universitaria, Ilha do Fundao, 21941-906 Rio de Janeiro, RJ (Brazil)

2011-06-15

Highlights: > First INPRO evaluation of IRIS economic competitiveness for deployment in Brazil. > Plant arrangement of three independent IRIS single units constructed in series. > Angra 3 reactor used as reference design for judgment of IRIS economic potential. > IRIS economically competes with 2nd generation nuclear power plants in Brazil - Abstract: This paper presents the results of the economic assessment of the International Reactor Innovative and Secure (IRIS) for deployment in Brazil using the assessment methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO initiated in 2001 and has the main objective of helping to ensure that nuclear energy will be available to contribute in a sustainable manner to the energy needs of the 21st century. Among its missions is the development of a methodology to assess innovative nuclear energy systems (INSs) on a global, regional and national basis. In 2005, Brazil submitted a proposal for the assessment of two small-size reactors as components of an INS, completed with a conventional open nuclear fuel cycle based on enriched uranium. One of the reactors assessed was IRIS, a small-size, modular, integral-type PWR reactor. IRIS was evaluated with regard to the areas of reactor safety and economics only. This paper outlines the rationale for the study and summarizes the results of the economic assessment. The study concluded that the reference design of IRIS complies with most of INPRO economics criteria and has potential to comply with the remaining ones.

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

Life cycle reliability assessment of new products—A Bayesian model updating approach

International Nuclear Information System (INIS)

Peng, Weiwen; Huang, Hong-Zhong; Li, Yanfeng; Zuo, Ming J.; Xie, Min

2013-01-01

The rapidly increasing pace and continuously evolving reliability requirements of new products have made life cycle reliability assessment of new products an imperative yet difficult work. While much work has been done to separately estimate reliability of new products in specific stages, a gap exists in carrying out life cycle reliability assessment throughout all life cycle stages. We present a Bayesian model updating approach (BMUA) for life cycle reliability assessment of new products. Novel features of this approach are the development of Bayesian information toolkits by separately including “reliability improvement factor” and “information fusion factor”, which allow the integration of subjective information in a specific life cycle stage and the transition of integrated information between adjacent life cycle stages. They lead to the unique characteristics of the BMUA in which information generated throughout life cycle stages are integrated coherently. To illustrate the approach, an application to the life cycle reliability assessment of a newly developed Gantry Machining Center is shown

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)

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

Remote assembly and maintenance of fusion reactors

International Nuclear Information System (INIS)

Becquet, M.C.; Farfaletti-Casali, F.

1991-01-01

This paper intend to present the state of the art in the field of remote assembly and maintenance, including system analysis design and operation for controlled fusion device such as JET, and the next NET and ITER reactors. The operational constraints of fusion reactors with respect to temperature, radiations dose rates and cumulated doses are considered with the resulting design requirements. Concepts like articulated boom, in-vessel vehicle and blanket handling device are presented. The close relations between computer simulations and experimental validation of those concepts are emphasized to ensure reliability of the operational behavior. Mockups and prototypes in reduced and full scale, as operating machines are described to illustrate the progress in remote operations for fusion reactors. The developments achieved at the Institute for System Engineering and Informatics of the Joint Research Center, in the field of remote blanket maintenance, reliability assessment of RH systems and remote cut and welding of lips joints are considered. (author)

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.

Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

1997-01-01

In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

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

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

Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

International Nuclear Information System (INIS)

Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

2012-01-01

Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

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

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

State of the art of Monte Carlo technics for reliable activated waste evaluations

International Nuclear Information System (INIS)

Culioli, Matthieu; Chapoutier, Nicolas; Barbier, Samuel; Janski, Sylvain

2016-01-01

This paper presents the calculation scheme used for many studies to assess the activities inventory of French shutdown reactors (including Pressurized Water Reactor, Heavy Water Reactor, Sodium-Cooled Fast Reactor and Natural Uranium Gas Cooled or UNGG). This calculation scheme is based on Monte Carlo calculations (MCNP) and involves advanced technique for source modeling, geometry modeling (with Computer-Aided Design integration), acceleration methods and depletion calculations coupling on 3D meshes. All these techniques offer efficient and reliable evaluations on large scale model with a high level of details reducing the risks of underestimation or conservatisms. (authors)

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

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.

Nuclear reactor kinetics and control

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

Cooling System Design Options for a Fusion Reactor

Science.gov (United States)

Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

1997-06-01

The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

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)

A model for reliability avaliation of the electrical supply source of the 1A3 and 1A4 control rods assemblies of Angra I reactor

International Nuclear Information System (INIS)

Yang, T.

1978-01-01

The reliability of the electrical power supply to the 4.16KV buses for the safety system operation of a nuclear power plant was studied. Particularly, Angra Unit I system was focused. Initially, reliability of each electrical supply source was estimated. Using a probabilistic approach based on the Markov processes, the system reliability was evaluated in terms of frequency and duration of loss of power supply and of the system failure probability evolution when one or more sources remained unavailable. Based on these results, certain reactor operating rules were proposed concerning later shutdown of the plant without compromising the nuclear reactor safety. A sensitivity analysis was also performed to show the different reliability parameter influences on final results. This analysis showed that the diesel system performs an important role in the power supply for a nuclear power plant [pt

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

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

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...

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.

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

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

Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

2014-10-15

The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

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.

A fuzzy-based reliability approach to evaluate basic events of fault tree analysis for nuclear power plant probabilistic safety assessment

International Nuclear Information System (INIS)

Purba, Julwan Hendry

2014-01-01

Highlights: • We propose a fuzzy-based reliability approach to evaluate basic event reliabilities. • It implements the concepts of failure possibilities and fuzzy sets. • Experts evaluate basic event failure possibilities using qualitative words. • Triangular fuzzy numbers mathematically represent qualitative failure possibilities. • It is a very good alternative for conventional reliability approach. - Abstract: Fault tree analysis has been widely utilized as a tool for nuclear power plant probabilistic safety assessment. This analysis can be completed only if all basic events of the system fault tree have their quantitative failure rates or failure probabilities. However, it is difficult to obtain those failure data due to insufficient data, environment changing or new components. This study proposes a fuzzy-based reliability approach to evaluate basic events of system fault trees whose failure precise probability distributions of their lifetime to failures are not available. It applies the concept of failure possibilities to qualitatively evaluate basic events and the concept of fuzzy sets to quantitatively represent the corresponding failure possibilities. To demonstrate the feasibility and the effectiveness of the proposed approach, the actual basic event failure probabilities collected from the operational experiences of the David–Besse design of the Babcock and Wilcox reactor protection system fault tree are used to benchmark the failure probabilities generated by the proposed approach. The results confirm that the proposed fuzzy-based reliability approach arises as a suitable alternative for the conventional probabilistic reliability approach when basic events do not have the corresponding quantitative historical failure data for determining their reliability characteristics. Hence, it overcomes the limitation of the conventional fault tree analysis for nuclear power plant probabilistic safety assessment

Methodology for allocating reliability and risk

International Nuclear Information System (INIS)

Cho, N.Z.; Papazoglou, I.A.; Bari, R.A.

1986-05-01

This report describes a methodology for reliability and risk allocation in nuclear power plants. The work investigates the technical feasibility of allocating reliability and risk, which are expressed in a set of global safety criteria and which may not necessarily be rigid, to various reactor systems, subsystems, components, operations, and structures in a consistent manner. The report also provides general discussions on the problem of reliability and risk allocation. The problem is formulated as a multiattribute decision analysis paradigm. The work mainly addresses the first two steps of a typical decision analysis, i.e., (1) identifying alternatives, and (2) generating information on outcomes of the alternatives, by performing a multiobjective optimization on a PRA model and reliability cost functions. The multiobjective optimization serves as the guiding principle to reliability and risk allocation. The concept of ''noninferiority'' is used in the multiobjective optimization problem. Finding the noninferior solution set is the main theme of the current approach. The final step of decision analysis, i.e., assessment of the decision maker's preferences could then be performed more easily on the noninferior solution set. Some results of the methodology applications to a nontrivial risk model are provided, and several outstanding issues such as generic allocation, preference assessment, and uncertainty are discussed. 29 refs., 44 figs., 39 tabs

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.

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

Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

International Nuclear Information System (INIS)

1993-09-01

The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses

Systems reliability analyses and risk analyses for the licencing procedure under atomic law

International Nuclear Information System (INIS)

Berning, A.; Spindler, H.

1983-01-01

For the licencing procedure under atomic law in accordance with Article 7 AtG, the nuclear power plant as a whole needs to be assessed, plus the reliability of systems and plant components that are essential to safety are to be determined with probabilistic methods. This requirement is the consequence of safety criteria for nuclear power plants issued by the Home Department (BMI). Systems reliability studies and risk analyses used in licencing procedures under atomic law are identified. The stress is on licencing decisions, mainly for PWR-type reactors. Reactor Safety Commission (RSK) guidelines, examples of reasoning in legal proceedings and arguments put forth by objectors are also dealt with. Correlations are shown between reliability analyses made by experts and licencing decisions by means of examples. (orig./HP) [de

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

Reactor vessel assessment and the development of a reactor vessel life extension program for Calvert Cliffs Units One and Two

International Nuclear Information System (INIS)

Montgomery, B.; Hijeck, P.J.

1988-01-01

A study has been undertaken to provide a general assessment of the life extension capabilities for the Calvert Cliffs Units One and Two reactor pressure vessels. The purpose of the study is to assess the general life extension capabilities for the Calvert Cliffs reactor pressure vessels based upon an extension and variation of the Surry pilot plant life extension study. This assessment provided a detailed reactor vessel surveillance program for plant life extension along with a hierarchy of specific tasks necessary for attaining maximum useful life. The assessment identified a number of critical issues which may impact life attainment and extension along with potential solutions to address these issues to ensure the life extension option is not precluded

Reliability centred maintenance of nuclear power plant facilities

International Nuclear Information System (INIS)

Kovacs, Zoltan; Novakova, Helena; Hlavac, Pavol; Janicek, Frantisek

2011-01-01

A method for the optimization of preventive maintenance nuclear power plant equipment, i.e. reliability centred maintenance, is described. The method enables procedures and procedure schedules to be defined such as allow the maintenance cost to be minimized without compromising operational safety or reliability. Also, combinations of facilities which remain available and ensure reliable operation of the reactor unit during the maintenance of other pieces of equipment are identified. The condition-based maintenance concept is used in this process, thereby preventing unnecessary operator interventions into the equipment, which are often associated with human errors. Where probabilistic safety assessment is available, the most important structures, systems and components with the highest maintenance priority can be identified. (orig.)

Assessment of Current Inservice Inspection and Leak Monitoring Practices for Detecting Materials Degradation in Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Anderson, Michael T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Simonen, Fredric A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Muscara, Joseph [US Nuclear Regulatory Commission (NRC), Rockville, MD (United States); Doctor, Steven R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kupperman, David S. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-01

An assessment was performed to determine the effectiveness of existing inservice inspection (ISI) and leak monitoring techniques, and recommend improvements, as necessary, to the programs as currently performed for light water reactor (LWR) components. Information from nuclear power plant (NPP) aging studies and from the U. S. Nuclear Regulatory Commission’s Generic Aging Lessons Learned (GALL) report (NUREG-1801) was used to identify components that have already experienced, or are expected to experience, degradation. This report provides a discussion of the key aspects and parameters that constitute an effective ISI program and a discussion of the basis and background against which the effectiveness of the ISI and leak monitoring programs for timely detection of degradation was evaluated. Tables based on the GALL components were used to systematically guide the process, and table columns were included that contained the ISI requirements and effectiveness assessment. The information in the tables was analyzed using histograms to reduce the data and help identify any trends. The analysis shows that the overall effectiveness of the ISI programs is very similar for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). The evaluations conducted as part of this research showed that many ISI programs are not effective at detecting degradation before its extent reached 75% of the component wall thickness. This work should be considered as an assessment of NDE practices at this time; however, industry and regulatory activities are currently underway that will impact future effectiveness assessments. A number of actions have been identified to improve the current ISI programs so that degradation can be more reliably detected.

Decommissioning planning and the assessment of alternatives for the Hanford production reactors

International Nuclear Information System (INIS)

Miller, C.E. Jr.; Potter, R.F.

1985-01-01

Several years ago, the US Department of Energy began assessing alternatives and planning the decommissioning of eight shut-down plutonium production reactors located on the DOE Hanford Site in Washington State. The first of these graphite-moderated, water-cooled, reactors was built and started up in 1944 as part of the World War II Manhattan Project. The last of them started up in 1955. The eight reactors each operated for 12 to 24 years, with all eight operating simultaneously for about 10 years. In the 1960's, production needs declined and the reactors were one-by-one permanently shut down, the last of them in 1971. (A ninth Hanford production reactor, N Reactor, was started up in 1963; it is still operating and is not within the scope of the decommissioning planning and alternatives assessment work reported in this paper). This paper provides an overview description of the decommissioning plan for the eight shut-down Hanford production reactors and their associated fuel storage basins. Included are descriptions of the decommissioning alternatives considered for the facilities, along with discussions of National Environmental Policy Act (NEPA) process activities applicable to the Hanford decommissioning work. The criteria used in assessing decommissioning alternatives and the assumptions used in the decommissioning planning are identified. 4 refs., 8 figs., 3 tabs

Uncertainty propagation and sensitivity analysis in system reliability assessment via unscented transformation

International Nuclear Information System (INIS)

Rocco Sanseverino, Claudio M.; Ramirez-Marquez, José Emmanuel

2014-01-01

The reliability of a system, notwithstanding it intended function, can be significantly affected by the uncertainty in the reliability estimate of the components that define the system. This paper implements the Unscented Transformation to quantify the effects of the uncertainty of component reliability through two approaches. The first approach is based on the concept of uncertainty propagation, which is the assessment of the effect that the variability of the component reliabilities produces on the variance of the system reliability. This assessment based on UT has been previously considered in the literature but only for system represented through series/parallel configuration. In this paper the assessment is extended to systems whose reliability cannot be represented through analytical expressions and require, for example, Monte Carlo Simulation. The second approach consists on the evaluation of the importance of components, i.e., the evaluation of the components that most contribute to the variance of the system reliability. An extension of the UT is proposed to evaluate the so called “main effects” of each component, as well to assess high order component interaction. Several examples with excellent results illustrate the proposed approach. - Highlights: • Simulation based approach for computing reliability estimates. • Computation of reliability variance via 2n+1 points. • Immediate computation of component importance. • Application to network systems

Application of probabilistic risk assessment to advanced liquid metal reactor designs

International Nuclear Information System (INIS)

Carroll, W.P.; Temme, M.I.

1987-01-01

The United States Department of Energy (US DOE) has been active in the development and application of probabilistic risk assessment methods within its liquid metal breeder reactor development program for the past eleven years. These methods have been applied to comparative risk evaluations, the selection of design features for reactor concepts, the selection and emphasis of research and development programs, and regulatory discussions. The application of probabilistic methods to reactors which are in the conceptual design stage presents unique data base, modeling, and timing challenges, and excellent opportunities to improve the final design. We provide here the background and insights on the experience which the US DOE liquid metal breeder reactor program has had in its application of probabilistic methods to the Clinch River Breeder Reactor Plant project, the Conceptual Design State of the Large Development Plant, and updates on this design. Plans for future applications of probabilistic risk assessment methods are also discussed. The US DOE is embarking on an innovative design program for liquid metal reactors. (author)

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

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)

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

Assessment and status report High-Temperature Gas-Cooled Reactor gas-turbine technology

International Nuclear Information System (INIS)

1981-01-01

Purpose of this report is to present a brief summary assessment of the High Temperature Gas-Cooled Reactor - Gas Turbine (HTGR-GT) technology. The focal point for the study was a potential 2000 MW(t)/800 MW(e) HTGR-GT commercial plant. Principal findings of the study were that: the HTGR-GT is feasible, but with significantly greater development risk than the HTGR-SC (Steam Cycle). At the level of performance corresponding to the reference design, no incremental economic incentive can be identified for the HTGR-GT to offset the increased development costs and risk relative to the HTGR-SC. The relative economics of the HTGR-GT and HTGR-SC are not significantly impacted by dry cooling considerations. While reduced cycel complexity may ultimately result in a reliability advantage for the HTGR-GT, the value of that potential advantage was not quantified

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)

Review of Reliability Assessment of Westinghouse SSPS Using SPC by WEC

International Nuclear Information System (INIS)

Kang, H. T.; Chung, H. Y.

2007-01-01

Westinghouse Electric Company (WEC) has accomplished the reliability assessment of Westinghouse Solid State Protection System (SSPS) in KORI no. 2, 3, 4, and YGN no. 1, 2. In their studies, it is reported that creating a cost-effective plan for improving the reliability of the SSPS and at KORI no. 2, 3 and 4, and YGN no. 1, 2 should be needed while reducing their maintenance cost. In this paper, we reviewed the reliability assessment of Westinghouse SSPS analyzed in two performance standards, availability, and the maintenance expense using Statistic Process Control (SPC). As a result, it is concluded all plants have several failures reported but no effect on the system's availability, and the maintenance expense analysis did not reduce the current maintenance expense by 30%. Therefore, overall review for the reliability assessment is that a new strategy for cost-effective plan and/or upgrade approach for improving the reliability of the aging Westinghouse SSPS should be needed

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.

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.

NUPEC proves reliability of LWR fuel assemblies

International Nuclear Information System (INIS)

Anon.

1987-01-01

It is very important in assuring the safety of nuclear reactors to confirm the reliability of fuel assemblies. The test program of the Nuclear Power Engineering Center on the reliability of fuel assemblies has verified the high performance and reliability of Japanese LWR fuels, and confirmed the propriety of their design and fabrication. This claim is based on the data obtained from the fuel assemblies irradiated in commercial reactors. The NUPEC program includes irradiation test which has been conducted for 11 years since fiscal 1976, and the maximum thermal loading test using the out of pile test facilities simulating a real reactor which has been continued since fiscal 1978. The irradiation test on BWR fuel assemblies in No.3 reactor in Fukushima No.1 Nuclear Power Station, Tokyo Electric Power Co., Inc., and on PWR fuel assemblies in No.3 reactor in Mihama Power Station, Kansai Electric Power Co., Inc., and the maximum thermal loading test on BWR and PWR fuel assemblies are reported. The series of postirradiation examination of the fuel assemblies used for commercial reactors was conducted for the first time in Japan, and the highly systematic data on 27 items were obtained. (Kako, I.)

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

OpenAIRE

KO, DO-YOUNG; KIM, KYU-HYUNG

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

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....

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

Clinical assessment of scapular positioning in musicians: an intertester reliability study.

Science.gov (United States)

Struyf, Filip; Nijs, Jo; De Coninck, Kris; Giunta, Marco; Mottram, Sarah; Meeusen, Romain

2009-01-01

The reliability of the measurement of the distance between the posterior border of the acromion and the wall and the reliability of the modified lateral scapular slide test have not been studied. Overall, the reliability of the clinical tools used to assess scapular positioning has not been studied in musicians. To examine the intertester reliability of scapular observation and 2 clinical tests for the assessment of scapular positioning in musicians. Intertester reliability study. University research laboratory. Thirty healthy student musicians at a single university. Two assessors performed a standardized observation protocol, the measurement of the distance between the posterior border of the acromion and the wall, and the modified lateral scapular slide test. Each assessor was blinded to the other's findings. The intertester reliability coefficients (kappa) for the observation in relaxed position, during unloaded movement, and during loaded movement were 0.41, 0.63, and 0.36, respectively. The kappa values for the observation of tilting and winging at rest were 0.48 and 0.42, respectively; during unloaded movement, the kappa values were 0.52 and 0.78, respectively; and with a 1-kg load, the kappa values were 0.24 and 0.50, respectively. The intraclass correlation coefficient (ICC) of the measurement of the acromial distance was 0.72 in relaxed position and 0.75 with the participant actively retracting both shoulders. The ICCs for the modified lateral scapular slide test varied between 0.63 and 0.58. Our results demonstrated that the modified lateral scapular slide test was not a reliable tool to assess scapular positioning in these participants. Our data indicated that scapular observation in the relaxed position and during unloaded abduction in the frontal plane was a reliable assessment tool. The reliability of the measurement of the distance between the posterior border of the acromion and the wall in healthy musicians was moderate.

Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

2002-01-01

The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

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...

Safety and reliability criteria

International Nuclear Information System (INIS)

O'Neil, R.

1978-01-01

Nuclear power plants and, in particular, reactor pressure boundary components have unique reliability requirements, in that usually no significant redundancy is possible, and a single failure can give rise to possible widespread core damage and fission product release. Reliability may be required for availability or safety reasons, but in the case of the pressure boundary and certain other systems safety may dominate. Possible Safety and Reliability (S and R) criteria are proposed which would produce acceptable reactor design. Without some S and R requirement the designer has no way of knowing how far he must go in analysing his system or component, or whether his proposed solution is likely to gain acceptance. The paper shows how reliability targets for given components and systems can be individually considered against the derived S and R criteria at the design and construction stage. Since in the case of nuclear pressure boundary components there is often very little direct experience on which to base reliability studies, relevant non-nuclear experience is examined. (author)

Regulatory analysis for the resolution of Generic Issue 115, enhancement of the reliability of the Westinghouse Solid State Protection System

International Nuclear Information System (INIS)

Basdekas, D.L.

1989-05-01

Generic Issue 115 addresses a concern related to the reliability of the Westinghouse reactor protection system for plants using the Westinghouse Solid State Protection System (SSPS). Several options for improving the reliability of the Westinghouse reactor trip function for these plants and their effect on core damage frequency (CDF) and overall risk were evaluated. This regulatory analysis includes a quantitative assessment of the costs and benefits associated with the various options for enhancing the reliability of the Westinghouse SSPS and provides insights for consideration and industry initiatives. No new regulatory requirements are proposed. 25 refs., 11 tabs

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.)

Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

2013-01-01

The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

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....

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

Fuel R and D to improve fuel reliability

International Nuclear Information System (INIS)

Yang, Rosa; Cheng, Bo; Deshon, Jeff; Edsinger, Kurt; Ozer, Odelli

2006-01-01

Light water reactor fuel is operating in an increasingly challenging environment. Fuel burnup extension and cycle length increase both can increase the local duty. Reactor water chemistry modifications for the purpose of protection the plant system materials have the potential of increasing fuel surface deposition and cladding corrosion and hydriding. The status of fuel performance in US reactors is summarized and an update of the Fuel Reliability Program' established by the utility industry to ensure reliability is provided. (author)

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

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

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

Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

2015-10-15

Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

International Nuclear Information System (INIS)

Utoh, Hiroyasu; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

2015-01-01

Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

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...

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

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)

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)

A Bayesian belief nets based quantitative software reliability assessment for PSA: COTS case study

International Nuclear Information System (INIS)

Eom, H. S.; Sung, T. Y.; Jeong, H. S.; Park, J. H.; Kang, H. G.; Lee, K. Y.; Park, J. K

2002-03-01

Current reliability assessments of safety critical software embedded in the digital systems in nuclear power plants are based on the rule-based qualitative assessment methods. Then recently practical needs require the quantitative features of software reliability for Probabilistic Safety Assessment (PSA) that is one of important methods being used in assessing the whole safety of nuclear power plant. But conventional quantitative software reliability assessment methods are not enough to get the necessary results in assessing the safety critical software used in nuclear power plants. Thus, current reliability assessment methods for these digital systems exclude the software part or use arbitrary values for the software reliability in the assessment. This reports discusses a Bayesian Belief Nets (BBN) based quantification method that models current qualitative software assessment in formal way and produces quantitative results required for PSA. Commercial Off-The-Shelf (COTS) software dedication process that KAERI developed was applied to the discussed BBN based method for evaluating the plausibility of the proposed method in PSA

Business Opportunities for Small Reactors

International Nuclear Information System (INIS)

Minato, Akio; Nishimura, Satoshi; Brown, Neil W.

2007-01-01

This report assesses the market potential and identifies a number of potential paths for developing the small nuclear reactor business. There are several potential opportunities identified and evaluated. Selecting a specific approach for the business development requires additional information related to a specific market and sources of capital to support the investment. If and how a market for small nuclear plants may develop is difficult to predict because of the complexity of the economic and institutional factors that will influence such development. Key factors are; economics, safety, proliferation resistance and investment risk. The economic and political interest of any of the identified markets is also dependent on successful demonstration of the safety and reliability of small nuclear reactor. Obtaining a US-NRC Standard Design approval would be an important development step toward establishing a market for small reactors. (authors)

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)

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)

Assessment of computational fluid dynamics (CFD) for nuclear reactor safety problems

International Nuclear Information System (INIS)

Smith, B. L.; Andreani, M.; Bieder, U.; Bestion, D.; Ducros, F.; Graffard, E.; Heitsch, M.; Scheuerer, M.; Henriksson, M.; Hoehne, T.; Rohde, U.; Lucas, D.; Komen, E.; Houkema, M.; Mahaffy, J.; Moretti, F.; Morii, T.; Muehlbauer, P.; Song, C.H.; Zigh, G.; Menter, F.; Watanabe, T.

2008-01-01

The basic objective of the present work was to provide documented evidence of the need to perform CFD simulations in Nuclear Reactor Safety (NRS), concentrating on single-phase applications, and to assess the competence of the present generation of CFD codes to perform these simulations reliably. The fulfilling of this objective involves multiple tasks, summarized as: to provide a classification of NRS problems requiring CFD analysis, to identify and catalogue existing CFD assessment bases, to identify shortcomings in CFD approaches, to put into place a means for extending the CFD assessment database, with an emphasis on NRS applications. The resulting document is presented here. After some introductory remarks, chapter 3 lists twenty-two NRS issues for which it is considered that the application of CFD would bring real benefits in terms of better predictive capability. This classification is followed by a short description of the safety issue, a state-of-the-art summary of what has been attempted, and what is still needed to be done to improve reliability. Chapter 4 details the assessment bases that have already been established in both the nuclear and non-nuclear domains, and discusses the usefulness and relevance of the work to NRS applications, where appropriate. This information is augmented in Chapter 5 by descriptions of the existing CFD assessment bases that have been established around specific, NRS problems. Typical examples are experiments devoted to the boron dilution issue, pressurised thermal shock, and thermal fatigue in pipes. Chapter 6 is devoted to identifying the technology gaps which need to be closed to make CFD a more trustworthy analytical tool. Some deficiencies identified are lack of a Phenomenon Identification and Ranking Table (PIRT), limitations in the range of application of turbulence models, coupling of CFD with neutronics and system codes, and computer power limitations. Most CFD codes currently being used have their own, custom

A Valid and Reliable Tool to Assess Nursing Students` Clinical Performance

OpenAIRE

Mehrnoosh Pazargadi; Tahereh Ashktorab; Sharareh Khosravi; Hamid Alavi majd

2013-01-01

Background: The necessity of a valid and reliable assessment tool is one of the most repeated issues in nursing students` clinical evaluation. But it is believed that present tools are not mostly valid and can not assess students` performance properly.Objectives: This study was conducted to design a valid and reliable assessment tool for evaluating nursing students` performance in clinical education.Methods: In this methodological study considering nursing students` performance definition; th...

Reliability and Validity of the Greek Migraine Disability Assessment (MIDAS) Questionnaire.

Science.gov (United States)

Oikonomidi, Theodora; Vikelis, Michail; Artemiadis, Artemios; Chrousos, George P; Darviri, Christina

2018-03-01

The Migraine Disability Assessment (MIDAS) Questionnaire is a reliable and valid instrument for migraine-related disability. Such a tool is needed to quantify migraine-related disability in the Greek population. This validation study aims to assess the test-retest reliability, internal consistency, item discriminant and convergent validity of the Greek translation of the MIDAS. Adults diagnosed with migraine completed the MIDAS Questionnaire on two occasions 3 weeks apart to assess reliability, and completed the RAND-36 to assess validity. Participants (n = 152) had a median MIDAS score of 24 and mostly severe disability (58% were grade IV). The test-retest reliability analysis (N = 59) revealed excellent reliability for the total score. Internal consistency was α = 0.71 for initial and α = 0.82 for retest completion. For item discriminant validity, the correlations between each question and the total score were significant, with high correlations for questions 2-5 (range 0.67 ≤ r ≤ 0.79; p MIDAS score tended to have better wellbeing. Psychometric properties are comparable with those of other published validation studies of the MIDAS and the original. Findings on question 1 show that missing work/school days may be closely related with increased affect issues. The Greek version of the MIDAS Questionnaire has good reliability and validity. This study allowed for cross-cultural comparability of research findings.

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

Aging assessment of the boiling-water reactor (BWR) standby liquid control system

International Nuclear Information System (INIS)

Orton, R.D.; Johnson, A.B.; Buckley, G.D.; Larson, L.L.

1992-10-01

Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious

A heuristic-based approach for reliability importance assessment of energy producers

International Nuclear Information System (INIS)

Akhavein, A.; Fotuhi Firuzabad, M.

2011-01-01

Reliability of energy supply is one of the most important issues of service quality. On one hand, customers usually have different expectations for service reliability and price. On the other hand, providing different level of reliability at load points is a challenge for system operators. In order to take reasonable decisions and obviate reliability implementation difficulties, market players need to know impacts of their assets on system and load-point reliabilities. One tool to specify reliability impacts of assets is the criticality or reliability importance measure by which system components can be ranked based on their effect on reliability. Conventional methods for determination of reliability importance are essentially on the basis of risk sensitivity analysis and hence, impose prohibitive calculation burden in large power systems. An approach is proposed in this paper to determine reliability importance of energy producers from perspective of consumers or distribution companies in a composite generation and transmission system. In the presented method, while avoiding immense computational burden, the energy producers are ranked based on their rating, unavailability and impact on power flows in the lines connecting to the considered load points. Study results on the IEEE reliability test system show successful application of the proposed method. - Research highlights: → Required reliability level at load points is a concern in modern power systems. → It is important to assess reliability importance of energy producers or generators. → Generators can be ranked based on their impacts on power flow to a selected area. → Ranking of generators is an efficient tool to assess their reliability importance.

Reliability of Lactation Assessment Tools Applied to Overweight and Obese Women.

Science.gov (United States)

Chapman, Donna J; Doughty, Katherine; Mullin, Elizabeth M; Pérez-Escamilla, Rafael

2016-05-01

The interrater reliability of lactation assessment tools has not been evaluated in overweight/obese women. This study aimed to compare the interrater reliability of 4 lactation assessment tools in this population. A convenience sample of 45 women (body mass index > 27.0) was videotaped while breastfeeding (twice daily on days 2, 4, and 7 postpartum). Three International Board Certified Lactation Consultants independently rated each videotaped session using 4 tools (Infant Breastfeeding Assessment Tool [IBFAT], modified LATCH [mLATCH], modified Via Christi [mVC], and Riordan's Tool [RT]). For each day and tool, we evaluated interrater reliability with 1-way repeated-measures analyses of variance, intraclass correlation coefficients (ICCs), and percentage absolute agreement between raters. Analyses of variance showed significant differences between raters' scores on day 2 (all scales) and day 7 (RT). Intraclass correlation coefficient values reflected good (mLATCH) to excellent reliability (IBFAT, mVC, and RT) on days 2 and 7. All day 4 ICCs reflected good reliability. The ICC for mLATCH was significantly lower than all others on day 2 and was significantly lower than IBFAT (day 7). Percentage absolute interrater agreement for scale components ranged from 31% (day 2: observable swallowing, RT) to 92% (day 7: IBFAT, fixing; and mVC, latch time). Swallowing scores on all scales had the lowest levels of interrater agreement (31%-64%). We demonstrated differences in the interrater reliability of 4 lactation assessment tools when applied to overweight/obese women, with the lowest values observed on day 4. Swallowing assessment was particularly unreliable. Researchers and clinicians using these scales should be aware of the differences in their psychometric behavior. © The Author(s) 2015.

Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

International Nuclear Information System (INIS)

Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

2014-01-01

The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

An effective surveillance strategy for reactor pressure vessel assessment in the long term operation perspective

International Nuclear Information System (INIS)

Chaouadi, R.; Gerard, R.

2015-01-01

The reactor pressure vessel (RPV) irradiation embrittlement is monitored by means of surveillance capsules containing the RPV belt-line materials, inserted inside the reactor pressure vessel (RPV) before the start of operation. These capsules are placed at location where they receive a higher neutron flux than the vessel wall, by a factor of the order of 2 to 3. They are regularly retrieved and tested to evaluate the RPV irradiation embrittlement according to specific regulatory procedures and standards, in order to guarantee the safe operation of the RPV throughout its lifetime. These procedures are often relying on empirical but conservative concepts. In parallel, material research reactor (MTR) irradiations are often used to support the surveillance data and to develop a better understanding of irradiation effects, not only qualitatively but also quantitatively. Taking advantage of the increased understanding of irradiation effects, analytical tools were developed to improve the evaluation embrittlement and quality assurance of the RPV embrittlement assessment. In this framework, an alternative but complementary surveillance program assessment was developed in Belgium, the so-called enhanced surveillance, in order to benefit from the latest developments in the area of materials science and irradiation effects. The neutron flux and fracture properties of the surveillance materials can be reliably characterized and correlated to each other using physically-based rather than empirical concepts. The enhanced surveillance approach is complementary to the mandatory regulatory procedure and allows quantifying the conservatism of the regulatory approach. The enhanced surveillance approach that uses the reconstitution technology to fabricate additional small size specimens, appropriate modeling tools and microstructural examination when required, makes it possible to rationalize all available information in a physically-based way