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)

A hybrid approach to quantify software reliability in nuclear safety systems

International Nuclear Information System (INIS)

Arun Babu, P.; Senthil Kumar, C.; Murali, N.

2012-01-01

Highlights: ► A novel method to quantify software reliability using software verification and mutation testing in nuclear safety systems. ► Contributing factors that influence software reliability estimate. ► Approach to help regulators verify the reliability of safety critical software system during software licensing process. -- Abstract: Technological advancements have led to the use of computer based systems in safety critical applications. As computer based systems are being introduced in nuclear power plants, effective and efficient methods are needed to ensure dependability and compliance to high reliability requirements of systems important to safety. Even after several years of research, quantification of software reliability remains controversial and unresolved issue. Also, existing approaches have assumptions and limitations, which are not acceptable for safety applications. This paper proposes a theoretical approach combining software verification and mutation testing to quantify the software reliability in nuclear safety systems. The theoretical results obtained suggest that the software reliability depends on three factors: the test adequacy, the amount of software verification carried out and the reusability of verified code in the software. The proposed approach may help regulators in licensing computer based safety systems in nuclear reactors.

Addressing Uniqueness and Unison of Reliability and Safety for a Better Integration

Science.gov (United States)

Huang, Zhaofeng; Safie, Fayssal

2016-01-01

Over time, it has been observed that Safety and Reliability have not been clearly differentiated, which leads to confusion, inefficiency, and, sometimes, counter-productive practices in executing each of these two disciplines. It is imperative to address this situation to help Reliability and Safety disciplines improve their effectiveness and efficiency. The paper poses an important question to address, "Safety and Reliability - Are they unique or unisonous?" To answer the question, the paper reviewed several most commonly used analyses from each of the disciplines, namely, FMEA, reliability allocation and prediction, reliability design involvement, system safety hazard analysis, Fault Tree Analysis, and Probabilistic Risk Assessment. The paper pointed out uniqueness and unison of Safety and Reliability in their respective roles, requirements, approaches, and tools, and presented some suggestions for enhancing and improving the individual disciplines, as well as promoting the integration of the two. The paper concludes that Safety and Reliability are unique, but compensating each other in many aspects, and need to be integrated. Particularly, the individual roles of Safety and Reliability need to be differentiated, that is, Safety is to ensure and assure the product meets safety requirements, goals, or desires, and Reliability is to ensure and assure maximum achievability of intended design functions. With the integration of Safety and Reliability, personnel can be shared, tools and analyses have to be integrated, and skill sets can be possessed by the same person with the purpose of providing the best value to a product development.

Operational safety reliability research

International Nuclear Information System (INIS)

Hall, R.E.; Boccio, J.L.

1986-01-01

Operating reactor events such as the TMI accident and the Salem automatic-trip failures raised the concern that during a plant's operating lifetime the reliability of systems could degrade from the design level that was considered in the licensing process. To address this concern, NRC is sponsoring the Operational Safety Reliability Research project. The objectives of this project are to identify the essential tasks of a reliability program and to evaluate the effectiveness and attributes of such a reliability program applicable to maintaining an acceptable level of safety during the operating lifetime at the plant

Standards in reliability and safety engineering

International Nuclear Information System (INIS)

O'Connor, Patrick

1998-01-01

This article explains how the highest 'world class' levels of reliability and safety are achieved, by adherence to the basic principles of excellence in design, production, support and maintenance, by continuous improvement, and by understanding that excellence and improvement lead to reduced costs. These principles are contrasted with the methods that have been developed and standardised, particularly military standards for reliability, ISO9000, and safety case regulations. The article concludes that the formal, standardised approaches are misleading and counterproductive, and recommends that they be replaced by a philosophy based on the realities of human performance

A reliability program approach to operational safety

International Nuclear Information System (INIS)

Mueller, C.J.; Bezella, W.A.

1985-01-01

A Reliability Program (RP) model based on proven reliability techniques is being formulated for potential application in the nuclear power industry. Methods employed under NASA and military direction, commercial airline and related FAA programs were surveyed and a review of current nuclear risk-dominant issues conducted. The need for a reliability approach to address dependent system failures, operating and emergency procedures and human performance, and develop a plant-specific performance data base for safety decision making is demonstrated. Current research has concentrated on developing a Reliability Program approach for the operating phase of a nuclear plant's lifecycle. The approach incorporates performance monitoring and evaluation activities with dedicated tasks that integrate these activities with operation, surveillance, and maintenance of the plant. The detection, root-cause evaluation and before-the-fact correction of incipient or actual systems failures as a mechanism for maintaining plant safety is a major objective of the Reliability Program. (orig./HP)

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)

Towards higher safety and reliability

Energy Technology Data Exchange (ETDEWEB)

Takekuro, I. [Tokyo Electric Power Company, Tokyo (Japan)

2001-06-01

Japanese electric power companies are now positioning themselves to gain a stronger position in the liberalised electricity market. Nuclear power in particular plays an important role in satisfying a large part of domestic electricity demand and its performance has continued to improve as a result of enhanced safety operation and tough maintenance programmes. Although the criticality accident which occurred in 1999 shocked not only the public but also the nuclear industry itself, the accident provided an opportunity for the industry and the regulators to learn lessons and look again at safety issues. Japanese electric power companies are now eager to be seen as front-runners in the safe, reliable, and efficient generation of nuclear power for the twenty-first century. (author)

A reliability evaluation method for NPP safety DCS application software

International Nuclear Information System (INIS)

Li Yunjian; Zhang Lei; Liu Yuan

2014-01-01

In the field of nuclear power plant (NPP) digital i and c application, reliability evaluation for safety DCS application software is a key obstacle to be removed. In order to quantitatively evaluate reliability of NPP safety DCS application software, this paper propose a reliability evaluating method based on software development life cycle every stage's v and v defects density characteristics, by which the operating reliability level of the software can be predicted before its delivery, and helps to improve the reliability of NPP safety important software. (authors)

Improving the safety and reliability of Monju

International Nuclear Information System (INIS)

Itou, Kazumoto; Maeda, Hiroshi; Moriyama, Masatoshi

1998-01-01

Comprehensive safety review has been performed at Monju to determine why the Monju secondary sodium leakage accident occurred. We investigated how to improve the situation based on the results of the safety review. The safety review focused on five aspects of whether the facilities for dealing with the sodium leakage accident were adequate: the reliability of the detection method, the reliability of the method for preventing the spread of the sodium leakage accident, whether the documented operating procedures are adequate, whether the quality assurance system, program, and actions were properly performed and so on. As a result, we established for Monju a better method of dealing with sodium leakage accidents, rapid detection of sodium leakage, improvement of sodium drain facilities, and way to reduce damage to Monju systems after an accident. We also improve the operation procedures and quality assurance actions to increase the safety and reliability of Monju. (author)

Software diversity: way to enhance safety?

International Nuclear Information System (INIS)

Dahll, G.; Bishop, P.

1990-01-01

The topic of the paper is the use of diversely produced programs to enhance the safety of computer-based systems applied in safety-critical areas. The paper starts with a survey of scientific investigations on the impact of software redundancy made at various institutions around the world. Main emphasis will, however, be put on the PODS/STEM projects, which have been performed at the OECD Halden Project in cooperation with the Technical Research Center of Finland, the Safety and Reliability Directorate, AEA Technology, UK, and Central Electricity Research Laboratory (now National Power Technology and Environment Centre), UK. In these projects, three program versions were made independently by three different teams, all based on the same specification. The three programs were tested back-to-back with a large amount of test data. The experience and results from this process were carefully logged and used for further analysis. Various strategies for test data selection were compared, with respect to fault finding strategies, as well as to branch and statement coverages of the tested programs. The assumption of independence of failures in diversely produced programs was investigated. A particularly interesting effect, namely failure masking due to program structure, was revealed. Static analysis techniques, software measures, and software reliability estimates were also studied. (author)

STARS software tool for analysis of reliability and safety

International Nuclear Information System (INIS)

Poucet, A.; Guagnini, E.

1989-01-01

This paper reports on the STARS (Software Tool for the Analysis of Reliability and Safety) project aims at developing an integrated set of Computer Aided Reliability Analysis tools for the various tasks involved in systems safety and reliability analysis including hazard identification, qualitative analysis, logic model construction and evaluation. The expert system technology offers the most promising perspective for developing a Computer Aided Reliability Analysis tool. Combined with graphics and analysis capabilities, it can provide a natural engineering oriented environment for computer assisted reliability and safety modelling and analysis. For hazard identification and fault tree construction, a frame/rule based expert system is used, in which the deductive (goal driven) reasoning and the heuristic, applied during manual fault tree construction, is modelled. Expert system can explain their reasoning so that the analyst can become aware of the why and the how results are being obtained. Hence, the learning aspect involved in manual reliability and safety analysis can be maintained and improved

Enhancement of the reliability of automated ultrasonic inspections using tools of quantitative NDT

International Nuclear Information System (INIS)

Kappes, W.; Baehr, W.; Kroening, M.; Schmitz, V.

1994-01-01

To achieve reliable test results from automated ultrasonic inspection of safety related components, optimization and integral consideration of the various inspection stages - inspection planning, inspection performance and evaluation of results - are indispensable. For this purpose, a large potential of methods is available: advanced measurement techniques, mathematical-numerical modelling processes, artificial intelligence tools, data bases and CAD systems. The potential inherent in these methods to enhance inspection reliability is outlined by way of different applications. (orig.) [de

Partial Safety Factors and Target Reliability Level in Danish Structural Codes

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Hansen, J. O.; Nielsen, T. A.

2001-01-01

The partial safety factors in the newly revised Danish structural codes have been derived using a reliability-based calibration. The calibrated partial safety factors result in the same average reliability level as in the previous codes, but a much more uniform reliability level has been obtained....... The paper describes the code format, the stochastic models and the resulting optimised partial safety factors....

Optimal, Reliability-Based Code Calibration

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard

2002-01-01

Reliability based code calibration is considered in this paper. It is described how the results of FORM based reliability analysis may be related to the partial safety factors and characteristic values. The code calibration problem is presented in a decision theoretical form and it is discussed how...... of reliability based code calibration of LRFD based design codes....

Performance and Reliability of DSRC Vehicular Safety Communication: A Formal Analysis

Directory of Open Access Journals (Sweden)

2009-02-01

Full Text Available IEEE- and ASTM-adopted dedicated short range communications (DSRC standard toward 802.11p is a key enabling technology for the next generation of vehicular safety communication. Broadcasting of safety messages is one of the fundamental services in DSRC. There have been numerous publications addressing design and analysis of such broadcast ad hoc system based on the simulations. For the first time, an analytical model is proposed in this paper to evaluate performance and reliability of IEEE 802.11a-based vehicle-to-vehicle (V2V safety-related broadcast services in DSRC system on highway. The proposed model takes two safety services with different priorities, nonsaturated message arrival, hidden terminal problem, fading transmission channel, transmission range, IEEE 802.11 backoff counter process, and highly mobile vehicles on highway into account. Based on the solutions to the proposed analytic model, closed-form expressions of channel throughput, transmission delay, and packet reception rates are derived. From the obtained numerical results under various offered traffic and network parameters, new insights and enhancement suggestions are given.

Reliability of computerized safety systems at nuclear power plants. Report of a technical committee meeting held in Vienna, 21-25 June 1993

International Nuclear Information System (INIS)

1995-03-01

Computer based technology is increasingly used in order to perform safety functions. In some recently designed nuclear power plants the whole safety system is computerized. In older plants replacement of conventional technology based system is seen to be of benefit. If the new technology is to be used, it must meet at least the same level of quality and reliability requirements as specified for conventional technology. However, there is a potential for enhancing the safety of nuclear power plants if the full power of computer technology is applied correctly through well designed, engineered and tested systems which are properly installed and maintained. It is essential that areas where reliability and quality can be improved are identified and that methods for assessing and assuring reliability are developed. The results of the Technical Committee Meeting on Reliability of Computerized Safety Systems at Nuclear Power Plants presented in this report are a step on the road to this goal of improved nuclear safety. Refs, figs and tabs

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

Safety and reliability of automatization software

Energy Technology Data Exchange (ETDEWEB)

Kapp, K; Daum, R [Karlsruhe Univ. (TH) (Germany, F.R.). Lehrstuhl fuer Angewandte Informatik, Transport- und Verkehrssysteme

1979-02-01

Automated technical systems have to meet very high requirements concerning safety, security and reliability. Today, modern computers, especially microcomputers, are used as integral parts of those systems. In consequence computer programs must work in a safe and reliable mannter. Methods are discussed which allow to construct safe and reliable software for automatic systems such as reactor protection systems and to prove that the safety requirements are met. As a result it is shown that only the method of total software diversification can satisfy all safety requirements at tolerable cost. In order to achieve a high degree of reliability, structured and modular programming in context with high level programming languages are recommended.

Reliability of containment and safety-related structures

International Nuclear Information System (INIS)

Nessim, M.A.

1995-09-01

A research program on Reliability of Containment and Safety-related Structures has been developed and is described in this document. This program is designed to support AECB's regulatory activities aimed at ensuring the safety of these structures. These activities include evaluating submissions by operators and requesting special assessments when necessary. The results of the proposed research will also be useful in revising and enhancing the CSA design standards for containment and safety-related structures. The process of developing the research program started with an information collection and review phase. The sources of information included C-FER's previous work in the area, various recent research publications, regulatory documents and relevant design standards, and a detailed discussion with AECB staff. The second step was to outline the process of reliability evaluation, and identify the required models and parameters. Comparison between the required and available information was used to identify gaps in the state-of-the-art, and the research program was designed to fill these gaps. The program is organized in four major topics, namely: development of an approach for reliability analysis; compilation and development of the required analysis tools; application to specific problems related to design, assessment, maintenance and testing of structures; and testing and validation. It is suggested that the program should be supported by an on-going process of communication and consultation between AECB staff and industry experts. This will lend credibility to the results and facilitate their future application. (author). 1 fig

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

Safety and reliability. V. 1. Proceedings

International Nuclear Information System (INIS)

Soares, C.G.

1997-01-01

Proceedings of a 1997 conference on industrial safety and reliability are reported. The first volume looks at risk management, probabilistic safety assessment and management styles in various industrial settings, including nuclear power plants. The second volume addresses safety and reliability in the offshore and transport industries, focusing on the role of staff training and appropriate maintenance routines to effectively reduce accidents and outages. (UK)

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

Global optimization of maintenance and surveillance testing based on reliability and probabilistic safety assessment. Research project

International Nuclear Information System (INIS)

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

1997-01-01

Background, objective, scope, detailed working plan and follow-up and final product of the project ''Global optimization of maintenance and surveillance testing based on reliability and probabilistic safety assessment'' are described

Reliability-based assessment of polyethylene pipe creep lifetime

International Nuclear Information System (INIS)

Khelif, Rabia; Chateauneuf, Alaa; Chaoui, Kamel

2007-01-01

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

Reliability-based assessment of polyethylene pipe creep lifetime

Energy Technology Data Exchange (ETDEWEB)

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

2007-12-15

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

Contribution to a quantitative assessment model for reliability-based metrics of electronic and programmable safety-related functions

International Nuclear Information System (INIS)

Hamidi, K.

2005-10-01

The use of fault-tolerant EP architectures has induced growing constraints, whose influence on reliability-based performance metrics is no more negligible. To face up the growing influence of simultaneous failure, this thesis proposes, for safety-related functions, a new-trend assessment method of reliability, based on a better taking into account of time-aspect. This report introduces the concept of information and uses it to interpret the failure modes of safety-related function as the direct result of the initiation and propagation of erroneous information until the actuator-level. The main idea is to distinguish the apparition and disappearance of erroneous states, which could be defined as intrinsically dependent of HW-characteristic and maintenance policies, and their possible activation, constrained through architectural choices, leading to the failure of safety-related function. This approach is based on a low level on deterministic SED models of the architecture and use non homogeneous Markov chains to depict the time-evolution of probabilities of errors. (author)

Addressing the fundamental issues in reliability evaluation of passive safety of AP1000 for a comparison with active safety of PWR

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Yang Ming

2013-01-01

Passive safety systems adopted in advanced Pressurized Water Reactor (PWR), such as AP1000 and EPR, should attain higher reliability than the existing active safety systems of the conventional PWR. The objective of this study is to discuss the fundamental issues relating to the reliability evaluation of AP1000 passive safety systems for a comparison with the active safety systems of conventional PWR, based on several aspects. First, comparisons between conventional PWR and AP1000 are made from the both aspects of safety design and cost reduction. The main differences between these PWR plants exist in the configurations of safety systems: AP1000 employs the passive safety system while reducing the number of active systems. Second, the safety of AP1000 is discussed from the aspect of severe accident prevention in the event of large break loss of coolant accidents (LOCA). Third, detailed fundamental issues on reliability evaluation of AP1000 passive safety systems are discussed qualitatively by using single loop models of safety systems of both PWRs plants. Lastly, methodology to conduct quantitative estimation of dynamic reliability for AP1000 passive safety systems in LOCA condition is discussed, in order to evaluate the reliability of AP1000 in future by a success-path-based reliability analysis method (i.e., GO-FLOW). (author)

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.)

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.)

On safety enhancements for medical robots

International Nuclear Information System (INIS)

Ng, W.S.; Tan, C.K.

1996-01-01

Both software and hardware methods to enhance safety are discussed for active medical robots applied to, among others, neurosurgery, orthopaedic surgery and prostatectomy. This paper advocates that while it is practically difficult, if not impossible, for software reliability to be 100%, there are positive measures by which a medical robot system can be made adequately or inherently safe. Such measures avoid the problems of software reliability but turn to mathematical logic directly to build a safer system. Examples in a newly developed prototype, known as surgeon assistant robot for selected urological disorders (SARUD), are given to illustrate the concept. Although software measures to promote reliability of a system is less preferred compared to hardware measures, as it can never escape from operating on a hardware platform, it is suggested that a complementary/ hybrid approach can be a good solution for achieving a safe and flexible (by being reprogrammable) system. A totally independent safety monitor is being built. It can arrest a servo runaway and detect out-of-safe-boundary conditions, using encoder pulses as input. This dedicated system can resolve some major safety concerns of a medical robot such as SARUD

Space transportation main engine reliability and safety

Science.gov (United States)

Monk, Jan C.

1991-01-01

Viewgraphs are used to illustrate the reliability engineering and aerospace safety of the Space Transportation Main Engine (STME). A technology developed is called Total Quality Management (TQM). The goal is to develop a robust design. Reducing process variability produces a product with improved reliability and safety. Some engine system design characteristics are identified which improves reliability.

Development of reliability database for safety-related I and C component based on operating experience of KSNP

International Nuclear Information System (INIS)

Jang, S. C.; Han, S. H.; Min, K. R.

2001-01-01

Reliability database for safety-related I and C components has been developed, based on domestic operating experience of total 8.63 years from four units-Yonggwang Units 3 and 4, and Ulchin Units 3 and 4. This plant-specific data of safety-related I and C components has compared with operating experience for CE-supplied plants in U.S.A. As a results, we found that on the whole the domestic reliability data was similar to CE-supplied plants in USA, through lots of failures occurred early in the commercial operation were included in our analyses without percolation

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-05-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

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.)

Human reliability analysis of performing tasks in plants based on fuzzy integral

International Nuclear Information System (INIS)

Washio, Takashi; Kitamura, Yutaka; Takahashi, Hideaki

1991-01-01

The effective improvement of the human working conditions in nuclear power plants might be a solution for the enhancement of the operation safety. The human reliability analysis (HRA) gives a methodological basis of the improvement based on the evaluation of human reliability under various working conditions. This study investigates some difficulties of the human reliability analysis using conventional linear models and recent fuzzy integral models, and provides some solutions to the difficulties. The following practical features of the provided methods are confirmed in comparison with the conventional methods: (1) Applicability to various types of tasks (2) Capability of evaluating complicated dependencies among working condition factors (3) A priori human reliability evaluation based on a systematic task analysis of human action processes (4) A conversion scheme to probability from indices representing human reliability. (author)

Reliability Based Ship Structural Design

DEFF Research Database (Denmark)

Dogliani, M.; Østergaard, C.; Parmentier, G.

1996-01-01

This paper deals with the development of different methods that allow the reliability-based design of ship structures to be transferred from the area of research to the systematic application in current design. It summarises the achievements of a three-year collaborative research project dealing...... with developments of models of load effects and of structural collapse adopted in reliability formulations which aim at calibrating partial safety factors for ship structural design. New probabilistic models of still-water load effects are developed both for tankers and for containerships. New results are presented...... structure of several tankers and containerships. The results of the reliability analysis were the basis for the definition of a target safety level which was used to asses the partial safety factors suitable for in a new design rules format to be adopted in modern ship structural design. Finally...

Possibilities and Limitations of Applying Software Reliability Growth Models to Safety- Critical Software

International Nuclear Information System (INIS)

Kim, Man Cheol; Jang, Seung Cheol; Ha, Jae Joo

2006-01-01

As digital systems are gradually introduced to nuclear power plants (NPPs), the need of quantitatively analyzing the reliability of the digital systems is also increasing. Kang and Sung identified (1) software reliability, (2) common-cause failures (CCFs), and (3) fault coverage as the three most critical factors in the reliability analysis of digital systems. For the estimation of the safety-critical software (the software that is used in safety-critical digital systems), the use of Bayesian Belief Networks (BBNs) seems to be most widely used. The use of BBNs in reliability estimation of safety-critical software is basically a process of indirectly assigning a reliability based on various observed information and experts' opinions. When software testing results or software failure histories are available, we can use a process of directly estimating the reliability of the software using various software reliability growth models such as Jelinski- Moranda model and Goel-Okumoto's nonhomogeneous Poisson process (NHPP) model. Even though it is generally known that software reliability growth models cannot be applied to safety-critical software due to small number of expected failure data from the testing of safety-critical software, we try to find possibilities and corresponding limitations of applying software reliability growth models to safety critical software

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-01-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of performance, safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith S.G.H.W.R. that the necessary bias toward safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications and has to be paid for. By a detailed evaluation of S.G.H.W.R. fuel defects it is shown that very few defects can be shown to be related to design, rating or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that must be met to achieve the desirable reliability level. It is possible that large-scale experience with S.G.H.W.R. fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to S.G.H.W.R. fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. The philosophy and supporting experimental work programme are outlined and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to new 60 pin fuel element to be used in the commercial S.G.H.W.R.'s and how it compares in design and performance aspects with the 36 pin element that has been used to date in the Winfrith S.G.H.W.R

Development of reliability-based safety enhancement technology; development of organization concept model in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Hyun; Kim, Ju Youl; Kim, Yoon Ik; Yang, Hui Chang; Lee, Yong Sik; Kim, Se Hyung [Seoul National University, Seoul (Korea)

2002-03-01

The influences of organizational factors on safety of nuclear power plants are mentioned in the early 1970s and noticed after being focused on in the accident report of TMI in 1979. These needs let us implement this research and the purposes of this research are to assess the organizational influences and to develop the organizational conceptual model to establish the basis of identifying the organizational factors, using this model to contribute to enhance safety and economics in nuclear power plants. Eventually research on the organizational influences is expected to have two effects, which are to improve safety through identifying potential causes of accidents and to elevate economics as a new approach to more efficient operation of nuclear power plants. In this study, recent studies were surveyed on the organizational conceptual model, the identification of organizational factors, assessment of organizational influences and evaluation methods of organizational factors and organizational influences among the overseas and domestic researches. In addition specific characteristics of domestic nuclear power plants were tried to identify through plant visit and an evaluation method of organizational influences on component maintenance and human performance were developed and presented. 71 refs., 40 figs., 18 tabs. (Author)

Development of a Reliability Program approach to assuring operational nuclear safety

International Nuclear Information System (INIS)

Mueller, C.J.; Bezella, W.A.

1985-01-01

A Reliability Program (RP) model based on proven reliability techniques used in other high technology industries is being formulated for potential application in the nuclear power industry. Research findings are discussed. The reliability methods employed under NASA and military direction, commercial airline and related FAA programs were surveyed with several reliability concepts (e.g., quantitative reliability goals, reliability centered maintenance) appearing to be directly transferable. Other tasks in the RP development effort involved the benchmarking and evaluation of the existing nuclear regulations and practices relevant to safety/reliability integration. A review of current risk-dominant issues was also conducted using results from existing probabilistic risk assessment studies. The ongoing RP development tasks have concentrated on defining a RP for the operating phase of a nuclear plant's lifecycle. The RP approach incorporates safety systems risk/reliability analysis and performance monitoring activities with dedicated tasks that integrate these activities with operating, surveillance, and maintenance of the plant. The detection, root-cause evaluation and before-the-fact correction of incipient or actual systems failures as a mechanism for maintaining plant safety is a major objective of the RP

High level issues in reliability quantification of safety-critical software

International Nuclear Information System (INIS)

Kim, Man Cheol

2012-01-01

For the purpose of developing a consensus method for the reliability assessment of safety-critical digital instrumentation and control systems in nuclear power plants, several high level issues in reliability assessment of the safety-critical software based on Bayesian belief network modeling and statistical testing are discussed. Related to the Bayesian belief network modeling, the relation between the assessment approach and the sources of evidence, the relation between qualitative evidence and quantitative evidence, how to consider qualitative evidence, and the cause-consequence relation are discussed. Related to the statistical testing, the need of the consideration of context-specific software failure probabilities and the inability to perform a huge number of tests in the real world are discussed. The discussions in this paper are expected to provide a common basis for future discussions on the reliability assessment of safety-critical software. (author)

Efforts to improve safety and reliability of nuclear power plants in Kyushu Electric Power

International Nuclear Information System (INIS)

Yamamoto, Satoshi

2014-01-01

After the Fukushima accident, Kyushu Electric Power Co. took emergency safety measures requested by government to ensure power supply, coolant supply pumps and cooling water so as to keep cooling fuels in the reactor and spent fuel storage pool in case of losses of ordinary cooling capability caused by earthquake and tsunami. In order to improve safety and reliability of nuclear power plants, further efforts based on lessons learned from the Fukushima accident had been made to diversify corresponding equipment of safety measures in terms of prevention of core damage, prevention of containment failure, mitigation of radioactive materials release, cooling of spent fuel pit and ensurance of power supply, and to enhance emergency response capability so as to make operational management more complete. Additional safety measures applicable to new regulatory requirements against severe accidents were in progress. This article introduced details of such activities. (T. Tanaka)

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

Trends in Control Area of PLC Reliability and Safety Parameters

Directory of Open Access Journals (Sweden)

Juraj Zdansky

2008-01-01

Full Text Available Extension of the PLC application possibilities is closely related to increase of reliability and safety parameters. If the requirement of reliability and safety parameters will be suitable, the PLC could by implemented to specific applications such the safety-related processes control. The goal of this article is to show the way which producers are approaching to increase PLC`s reliability and safety parameters. The second goal is to analyze these parameters for range of present choice and describe the possibility how the reliability and safety parameters can be affected.

Reliability analysis and computation of computer-based safety instrumentation and control used in German nuclear power plant. Final report

International Nuclear Information System (INIS)

Ding, Yongjian; Krause, Ulrich; Gu, Chunlei

2014-01-01

The trend of technological advancement in the field of safety instrumentation and control (I and C) leads to increasingly frequent use of computer-based (digital) control systems which consisting of distributed, connected bus communications computers and their functionalities are freely programmable by qualified software. The advantages of the new I and C system over the old I and C system with hard-wired technology are e.g. in the higher flexibility, cost-effective procurement of spare parts, higher hardware reliability (through higher integration density, intelligent self-monitoring mechanisms, etc.). On the other hand, skeptics see the new technology with the computer-based I and C a higher potential by influences of common cause failures (CCF), and the easier manipulation by sabotage (IT Security). In this joint research project funded by the Federal Ministry for Economical Affaires and Energy (BMWi) (2011-2014, FJZ 1501405) the Otto-von-Guericke-University Magdeburg and Magdeburg-Stendal University of Applied Sciences are therefore trying to develop suitable methods for the demonstration of the reliability of the new instrumentation and control systems with the focus on the investigation of CCF. This expertise of both houses shall be extended to this area and a scientific contribution to the sound reliability judgments of the digital safety I and C in domestic and foreign nuclear power plants. First, the state of science and technology will be worked out through the study of national and international standards in the field of functional safety of electrical and I and C systems and accompanying literature. On the basis of the existing nuclear Standards the deterministic requirements on the structure of the new digital I and C system will be determined. The possible methods of reliability modeling will be analyzed and compared. A suitable method called multi class binomial failure rate (MCFBR) which was successfully used in safety valve applications will be

Aviation Fuel System Reliability and Fail-Safety Analysis. Promising Alternative Ways for Improving the Fuel System Reliability

Directory of Open Access Journals (Sweden)

I. S. Shumilov

2017-01-01

Full Text Available The paper deals with design requirements for an aviation fuel system (AFS, AFS basic design requirements, reliability, and design precautions to avoid AFS failure. Compares the reliability and fail-safety of AFS and aircraft hydraulic system (AHS, considers the promising alternative ways to raise reliability of fuel systems, as well as elaborates recommendations to improve reliability of the pipeline system components and pipeline systems, in general, based on the selection of design solutions.It is extremely advisable to design the AFS and AHS in accordance with Aviation Regulations АП25 and Accident Prevention Guidelines, ICAO (International Civil Aviation Association, which will reduce risk of emergency situations, and in some cases even avoid heavy disasters.ATS and AHS designs should be based on the uniform principles to ensure the highest reliability and safety. However, currently, this principle is not enough kept, and AFS looses in reliability and fail-safety as compared with AHS. When there are the examined failures (single and their combinations the guidelines to ensure the AFS efficiency should be the same as those of norm-adopted in the Regulations АП25 for AHS. This will significantly increase reliability and fail-safety of the fuel systems and aircraft flights, in general, despite a slight increase in AFS mass.The proposed improvements through the use of components redundancy of the fuel system will greatly raise reliability of the fuel system of a passenger aircraft, which will, without serious consequences for the flight, withstand up to 2 failures, its reliability and fail-safety design will be similar to those of the AHS, however, above improvement measures will lead to a slightly increasing total mass of the fuel system.It is advisable to set a second pump on the engine in parallel with the first one. It will run in case the first one fails for some reasons. The second pump, like the first pump, can be driven from the

Post Fukushima safety enhancements in Indian PHWRS

International Nuclear Information System (INIS)

Ramasomayajulu, M.; Khot, Pankaj; Chauhan, Ashok

2016-01-01

Fukushima event was reviewed in Nuclear Power Corporation of India (NPCIL) and based on these reviews, safety enhancements were identified for Indian PHWRs. Safety enhancements such as additional emergency power sources, enhanced onsite water inventories, external water injection arrangements (Hook up points), measures related to hydrogen management, containment venting provision, seismic trip, mobile pumps, onsite emergency support Centre. These safety enhancements were reviewed by the regulatory body (Atomic Energy Regulatory Board, AERB) and were approved for implementation. Most of these are either implemented or in the advance stage of implementation. The paper elaborates above safety enhancements implemented post Fukushima accident; and preparedness to use these provisions. (author)

Analysis and recommendations for a reliable programming of software based safety systems

International Nuclear Information System (INIS)

Nunez McLeod, J.; Nunez McLeod, J.E.; Rivera, S.S.

1997-01-01

The present paper summarizes the results of several studies performed for the development of high software on i486 microprocessors, towards its utilization for control and safety systems for nuclear power plants. The work is based on software programmed in C language. Several recommendations oriented to high reliability software are analyzed, relating the requirements on high level language to its influence on assembler level. Several metrics are implemented, that allow for the quantification of the results achieved. New metrics were developed and other were adapted, in order to obtain more efficient indexes for the software description. Such metrics are helpful to visualize the adaptation of the software under development to the quality rules under use. A specific program developed to assist the reliability analyst on this quantification is also present in the paper. It performs the analysis of an executable program written in C language, disassembling it and evaluating its inter al structures. (author)

New enhancements to SCALE for criticality safety analysis

International Nuclear Information System (INIS)

Hollenbach, D.F.; Bowman, S.M.; Petrie, L.M.; Parks, C.V.

1995-01-01

As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portability of the SCALE code package, and (9) other minor enhancements, modifications, and corrections to SCALE. Each of these additions and enhancements to the criticality safety analysis capabilities of the SCALE code system are discussed below

A SOFTWARE RELIABILITY ESTIMATION METHOD TO NUCLEAR SAFETY SOFTWARE

Directory of Open Access Journals (Sweden)

GEE-YONG PARK

2014-02-01

Full Text Available A method for estimating software reliability for nuclear safety software is proposed in this paper. This method is based on the software reliability growth model (SRGM, where the behavior of software failure is assumed to follow a non-homogeneous Poisson process. Two types of modeling schemes based on a particular underlying method are proposed in order to more precisely estimate and predict the number of software defects based on very rare software failure data. The Bayesian statistical inference is employed to estimate the model parameters by incorporating software test cases as a covariate into the model. It was identified that these models are capable of reasonably estimating the remaining number of software defects which directly affects the reactor trip functions. The software reliability might be estimated from these modeling equations, and one approach of obtaining software reliability value is proposed in this paper.

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

Human reliability in high dose rate afterloading radiotherapy based on FMECA

International Nuclear Information System (INIS)

Deng Jun; Fan Yaohua; Yue Baorong; Wei Kedao; Ren Fuli

2012-01-01

Objective: To put forward reasonable and feasible recommendations against the procedure with relative high risk during the high dose rate (HDR) afterloading radiotherapy, so as to enhance its clinical application safety, through studying the human reliability in the process of carrying out the HDR afterloading radiotherapy. Methods: Basic data were collected by on-site investigation and process analysis as well as expert evaluation. Failure mode, effect and criticality analysis (FMECA) employed to study the human reliability in the execution of HDR afterloading radiotherapy. Results: The FMECA model of human reliability for HDR afterloading radiotherapy was established, through which 25 procedures with relative high risk index were found,accounting for 14.1% of total 177 procedures. Conclusions: FMECA method in human reliability study for HDR afterloading radiotherapy is feasible. The countermeasures are put forward to reduce the human error, so as to provide important basis for enhancing clinical application safety of HDR afterloading radiotherapy. (authors)

Reliability enhancement of portal frame structure by finite element synthesis

International Nuclear Information System (INIS)

Nakagiri, S.

1989-01-01

The stochastic finite element methods have been applied to the evaluation of structural response and reliability of uncertain structural systems. The structural reliability index of the advanced first-order second moment (AFOSM) method is a candidate of the measure of assessing structural safety and reliability. The reliability index can be evaluated when a baseline design of structures under interest is proposed and the covariance matrix of the probabilistic variables is acquired to represent uncertainties involved in the structure systems. The reliability index thus evaluated is not assured the largest one for the structure. There is left a possibility to enhance the structural reliability for the given covariance matrix by changing the baseline design. From such a viewpoint of structural optimization, some ideas have been proposed to maximize the reliability or to minimize the failure probability of uncertain structural systems. A method of changing the design is proposed to increase the reliability index from its baseline value to another desired value. The reliability index in this paper is calculated mainly by the method of Lagrange multiplier

A Research Roadmap for Computation-Based Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-08-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

A Research Roadmap for Computation-Based Human Reliability Analysis

International Nuclear Information System (INIS)

Boring, Ronald; Mandelli, Diego; Joe, Jeffrey; Smith, Curtis; Groth, Katrina

2015-01-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

Columbus safety and reliability

Science.gov (United States)

Longhurst, F.; Wessels, H.

1988-10-01

Analyses carried out to ensure Columbus reliability, availability, and maintainability, and operational and design safety are summarized. Failure modes/effects/criticality is the main qualitative tool used. The main aspects studied are fault tolerance, hazard consequence control, risk minimization, human error effects, restorability, and safe-life design.

Reliability analysis of Angra I safety systems

International Nuclear Information System (INIS)

Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

1980-07-01

An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

Reliability on the move: safety and reliability in transportation

International Nuclear Information System (INIS)

Guy, G.B.

1989-01-01

The development of transportation has been a significant factor in the development of civilisation as a whole. Our technical ability to move people and goods now seems virtually limitless when one considers for example the achievements of the various space programmes. Yet our current achievements rely heavily on high standards of safety and reliability from equipment and the human component of transportation systems. Recent failures have highlighted our dependence on equipment and human reliability. This book represents the proceedings of the 1989 Safety and Reliability Society symposium held at Bath on 11-12 October 1989. The structure of the book follows the structure of the symposium itself and the papers selected represent current thinking the the wide field of transportation, and the areas of rail (6 papers, three on railway signalling), air including space (two papers), road (one paper), road and rail (two papers) and sea (three papers) are covered. There are four papers concerned with general transport issues. Three papers concerned with the transport of radioactive materials are indexed separately. (author)

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

International Nuclear Information System (INIS)

2014-09-01

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

Reliability model for common mode failures in redundant safety systems

International Nuclear Information System (INIS)

Fleming, K.N.

1974-12-01

A method is presented for computing the reliability of redundant safety systems, considering both independent and common mode type failures. The model developed for the computation is a simple extension of classical reliability theory. The feasibility of the method is demonstrated with the use of an example. The probability of failure of a typical diesel-generator emergency power system is computed based on data obtained from U. S. diesel-generator operating experience. The results are compared with reliability predictions based on the assumption that all failures are independent. The comparison shows a significant increase in the probability of redundant system failure, when common failure modes are considered. (U.S.)

Reliability Analysis for Safety Grade PLC(POSAFE-Q)

International Nuclear Information System (INIS)

Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

2012-01-01

Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

LOFT pressurizer safety: relief valve reliability

Energy Technology Data Exchange (ETDEWEB)

Brown, E.S.

1978-01-18

The LOFT pressurizer self-actuating safety-relief valves are constructed to the present state-of-the-art and should have reliability equivalent to the valves in use on PWR plants in the U.S. There have been no NRC incident reports on valve failures to lift that would challenge the Technical Specification Safety Limit. Fourteen valves have been reported as lifting a few percentage points outside the +-1% Tech. Spec. surveillance tolerance (9 valves tested over and 5 valves tested under specification). There have been no incident reports on failures to reseat. The LOFT surveillance program for assuring reliability is equivalent to nuclear industry practice.

LOFT pressurizer safety: relief valve reliability

International Nuclear Information System (INIS)

Brown, E.S.

1978-01-01

The LOFT pressurizer self-actuating safety-relief valves are constructed to the present state-of-the-art and should have reliability equivalent to the valves in use on PWR plants in the U.S. There have been no NRC incident reports on valve failures to lift that would challenge the Technical Specification Safety Limit. Fourteen valves have been reported as lifting a few percentage points outside the +-1% Tech. Spec. surveillance tolerance (9 valves tested over and 5 valves tested under specification). There have been no incident reports on failures to reseat. The LOFT surveillance program for assuring reliability is equivalent to nuclear industry practice

NPP Mochovce nuclear safety enhancement program

International Nuclear Information System (INIS)

Cech, J.; Baumester, P.

1997-01-01

Nuclear power plant Mochovce is currently under construction and an extensive nuclear safety enhancement programme is under way. The upgrading and modifications are based on IAEA documents and on those of the Nuclear Regulatory Authority of the Slovak Republic. Based on a contract concluded with Riskaudit from the CEC, safety examinations of the Mochovce design were performed. An extensive list of technical specifications of safety measures is given. (M.D.)

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 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.

Reliability Analysis of Public Survey in Satisfaction with Nuclear Safety

International Nuclear Information System (INIS)

Park, Moon Soo; Moon, Joo Hyun; Kang, Chang Sun

2005-01-01

Korea Institute of Nuclear Safety (KINS) carried out a questionnaire survey on public's understanding nuclear safety and regulation in order to grasp public acceptance for nuclear energy. The survey was planned to help to analyze public opinion on nuclear energy and provide basic data for advertising strategy and policy development. In this study, based on results of the survey, the reliability of the survey was evaluated according to each nuclear site

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

49 CFR Appendix E to Part 238 - General Principles of Reliability-Based Maintenance Programs

Science.gov (United States)

2010-10-01

... STANDARDS Pt. 238, App. E Appendix E to Part 238—General Principles of Reliability-Based Maintenance... 49 Transportation 4 2010-10-01 2010-10-01 false General Principles of Reliability-Based... the design level of safety and reliability of the equipment; (2) To restore safety and reliability to...

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

Evaluation of reliability assurance approaches to operational nuclear safety

International Nuclear Information System (INIS)

Mueller, C.J.; Bezella, W.A.

1984-01-01

This report discusses the results of research to evaluate existing and/or recommended safety/reliability assurance activities among nuclear and other high technology industries for potential nuclear industry implementation. Since the Three Mile Island (TMI) accident, there has been increased interest in the use of reliability programs (RP) to assure the performance of nuclear safety systems throughout the plant's lifetime. Recently, several Nuclear Regulatory Commission (NRC) task forces or safety issue review groups have recommended RPs for assuring the continuing safety of nuclear reactor plants. 18 references

Reliability Analysis and Calibration of Partial Safety Factors for Redundant Structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard

1998-01-01

Redundancy is important to include in the design and analysis of structural systems. In most codes of practice redundancy is not directly taken into account. In the paper various definitions of a deterministic and reliability based redundancy measure are reviewed. It is described how reundancy can...... be included in the safety system and how partial safety factors can be calibrated. An example is presented illustrating how redundancy is taken into account in the safety system in e.g. the Danish codes. The example shows how partial safety factors can be calibrated to comply with the safety level...

Reliability Analysis of Public Survey in Satisfaction with Nuclear Safety

Energy Technology Data Exchange (ETDEWEB)

Park, Moon Soo; Moon, Joo Hyun; Kang, Chang Sun [Seoul National Univ., Seoul (Korea, Republic of)

2005-07-01

Korea Institute of Nuclear Safety (KINS) carried out a questionnaire survey on public's understanding nuclear safety and regulation in order to grasp public acceptance for nuclear energy. The survey was planned to help to analyze public opinion on nuclear energy and provide basic data for advertising strategy and policy development. In this study, based on results of the survey, the reliability of the survey was evaluated according to each nuclear site.

Automatic creation of Markov models for reliability assessment of safety instrumented systems

International Nuclear Information System (INIS)

Guo Haitao; Yang Xianhui

2008-01-01

After the release of new international functional safety standards like IEC 61508, people care more for the safety and availability of safety instrumented systems. Markov analysis is a powerful and flexible technique to assess the reliability measurements of safety instrumented systems, but it is fallible and time-consuming to create Markov models manually. This paper presents a new technique to automatically create Markov models for reliability assessment of safety instrumented systems. Many safety related factors, such as failure modes, self-diagnostic, restorations, common cause and voting, are included in Markov models. A framework is generated first based on voting, failure modes and self-diagnostic. Then, repairs and common-cause failures are incorporated into the framework to build a complete Markov model. Eventual simplification of Markov models can be done by state merging. Examples given in this paper show how explosively the size of Markov model increases as the system becomes a little more complicated as well as the advancement of automatic creation of Markov models

Safety and reliability in superconducting MHD magnets

International Nuclear Information System (INIS)

Laverick, C.; Powell, J.; Hsieh, S.; Reich, M.; Botts, T.; Prodell, A.

1979-07-01

This compilation adapts studies on safety and reliability in fusion magnets to similar problems in superconducting MHD magnets. MHD base load magnet requirements have been identified from recent Francis Bitter National Laboratory reports and that of other contracts. Information relevant to this subject in recent base load magnet design reports for AVCO - Everett Research Laboratories and Magnetic Corporation of America is included together with some viewpoints from a BNL workshop on structural analysis needed for superconducting coils in magnetic fusion energy. A summary of design codes used in large bubble chamber magnet design is also included

Improving patient safety: patient-focused, high-reliability team training.

Science.gov (United States)

McKeon, Leslie M; Cunningham, Patricia D; Oswaks, Jill S Detty

2009-01-01

Healthcare systems are recognizing "human factor" flaws that result in adverse outcomes. Nurses work around system failures, although increasing healthcare complexity makes this harder to do without risk of error. Aviation and military organizations achieve ultrasafe outcomes through high-reliability practice. We describe how reliability principles were used to teach nurses to improve patient safety at the front line of care. Outcomes include safety-oriented, teamwork communication competency; reflections on safety culture and clinical leadership are discussed.

Evaluation for nuclear safety-critical software reliability of DCS

International Nuclear Information System (INIS)

Liu Ying

2015-01-01

With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I and C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make besement for evaluating the reliability and safety of DCS. (author)

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.

A Regulatory Perspective on the Performance and Reliability of Nuclear Passive Safety Systems

International Nuclear Information System (INIS)

Quan, Pham Trung; Lee, Sukho

2016-01-01

Passive safety systems have been proven to enhance the safety of NPPs. When an accident such as station blackout occurs, these systems can perform the following functions: the decay heat removal, passive safety injection, containment cooling, and the retention of radioactive materials. Following the IAEA definitions, using passive safety systems reduces reliance on active components to achieve proper actuation and not requiring operator intervention in accident conditions. That leads to the deviations in boundary conditions of the critical process or geometric parameters, which activate and operate the system to perform accident prevention and mitigation functions. The main difficulties in evaluation of functional failure of passive systems arise because of (a) lack of plant operational experience; (b) scarcity of adequate experimental data from integral test facilities or from separate effect tests in order to understand the performance characteristics of these passive systems, not only at normal operation but also during accidents and transients; (c) lack of accepted definitions of failure modes for these systems; and (d) difficulty in modeling certain physical behavior of these systems. Reliability assessment of the PSS is still one of the important issues. Several reliability methodologies such as REPAS, RMPS and ASPRA have been applied to the reliability assessments. However, some issues are remained unresolved due to lack of understanding of the treatment of dynamic failure characteristics of components of the PSS, the treatment of dynamic variation of independence process parameters such as ambient temperature and the functional failure criteria of the PSS. Dynamic reliability methodologies should be integrated in the PSS reliability analysis to have a true estimate of system failure probability. The methodology should estimate the physical variation of the parameters and the frequency of the accident sequences when the dynamic effects are considered

Transparent reliability model for fault-tolerant safety systems

International Nuclear Information System (INIS)

Bodsberg, Lars; Hokstad, Per

1997-01-01

A reliability model is presented which may serve as a tool for identification of cost-effective configurations and operating philosophies of computer-based process safety systems. The main merit of the model is the explicit relationship in the mathematical formulas between failure cause and the means used to improve system reliability such as self-test, redundancy, preventive maintenance and corrective maintenance. A component failure taxonomy has been developed which allows the analyst to treat hardware failures, human failures, and software failures of automatic systems in an integrated manner. Furthermore, the taxonomy distinguishes between failures due to excessive environmental stresses and failures initiated by humans during engineering and operation. Attention has been given to develop a transparent model which provides predictions which are in good agreement with observed system performance, and which is applicable for non-experts in the field of reliability

Engineering systems reliability, safety, and maintenance an integrated approach

CERN Document Server

Dhillon, B S

2017-01-01

Today, engineering systems are an important element of the world economy and each year billions of dollars are spent to develop, manufacture, operate, and maintain various types of engineering systems around the globe. Many of these systems are highly sophisticated and contain millions of parts. For example, a Boeing jumbo 747 is made up of approximately 4.5 million parts including fasteners. Needless to say, reliability, safety, and maintenance of systems such as this have become more important than ever before.　 Global competition and other factors are forcing manufacturers to produce highly reliable, safe, and maintainable engineering products. Therefore, there is a definite need for the reliability, safety, and maintenance professionals to work closely during design and other phases. Engineering Systems Reliability, Safety, and Maintenance: An Integrated Approach eliminates the need to consult many different and diverse sources in the hunt for the information required to design better engineering syste...

Use of PRA methodology for enhancing operational safety and reliability

International Nuclear Information System (INIS)

Chu, B.; Rumble, E.; Najafi, B.; Putney, B.; Young, J.

1985-01-01

This paper describes a broad scope, on-going R and D study, sponsored by the Electric Power Research Institute (EPRI) to utilize key features of the state-of-the-art plant information management and system analysis techniques to develop and demonstrate a practical engineering tool for assisting plant engineering and operational staff to perform their activities more effectively. The study is foreseen to consist of two major activities: to develop a user-friendly, integrated software system; and to demonstrate the applications of this software on-site. This integrated software, Reliability Analysis Program with In-Plant Data (RAPID), will consist of three types of interrelated elements: an Executive Controller which will provide engineering and operations staff users with interface and control of the other two software elements, a Data Base Manager which can acquire, store, select, and transfer data, and Applications Modules which will perform the specific reliability-oriented functions. A broad range of these functions has been envisaged. The immediate emphasis will be focused on four application modules: a Plant Status Module, a Technical Specification Optimization Module, a Reliability Assessment Module, and a Utility Module for acquiring plant data

Safety and reliability in the 90s: will past experience or prediction meet our needs?

International Nuclear Information System (INIS)

Walter, M.H.; Cox, R.F.

1990-01-01

Twenty-six papers are presented in the proceedings of the 1990 Safety and Reliability Society Symposium. The papers selected provide current thinking on improved methods for identification, quantification and management of risks based on the safety culture developed across a range of industries during the last decade. In particular organizational and management factors feature in a large number of the papers. Two papers on the safety of all the operating plants at Sellafield's irradiated nuclear fuel handling and reprocessing site and the selection of field component reliability data for use in nuclear safety studies are selected and indexed separately. (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

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

Application of safety and reliability approaches in the power sector: Inside-sectoral overview

DEFF Research Database (Denmark)

Kozine, Igor

2010-01-01

This chapter summarizes the state-of-the-art and state-of-practice on the applications of safety and reliability approaches in the Power Sector. The nature and composition of this industrial sector including the characteristics of major hazards are summarized. The present situation with regard...... to a number of key technical aspects involved in the use of safety and reliability approaches in the power sector is discussed. Based on this review a Technology Maturity Matrix is synthesized. Barriers to the wider use of risk and reliability methods in the design and operation of power installations...... are identified and possible ways of overcoming these barriers are suggested. Key issues and priorities for research are identified....

Optimal Reliability-Based Code Calibration

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Kroon, I. B.; Faber, Michael Havbro

1994-01-01

Calibration of partial safety factors is considered in general, including classes of structures where no code exists beforehand. The partial safety factors are determined such that the difference between the reliability for the different structures in the class considered and a target reliability...... level is minimized. Code calibration on a decision theoretical basis is also considered and it is shown how target reliability indices can be calibrated. Results from code calibration for rubble mound breakwater designs are shown....

Infusing Reliability Techniques into Software Safety Analysis

Science.gov (United States)

Shi, Ying

2015-01-01

Software safety analysis for a large software intensive system is always a challenge. Software safety practitioners need to ensure that software related hazards are completely identified, controlled, and tracked. This paper discusses in detail how to incorporate the traditional reliability techniques into the entire software safety analysis process. In addition, this paper addresses how information can be effectively shared between the various practitioners involved in the software safety analyses. The author has successfully applied the approach to several aerospace applications. Examples are provided to illustrate the key steps of the proposed approach.

Advances in methods and applications of reliability and safety analysis

International Nuclear Information System (INIS)

Fieandt, J.; Hossi, H.; Laakso, K.; Lyytikaeinen, A.; Niemelae, I.; Pulkkinen, U.; Pulli, T.

1986-01-01

The know-how of the reliability and safety design and analysis techniques of Vtt has been established over several years in analyzing the reliability in the Finnish nuclear power plants Loviisa and Olkiluoto. This experience has been later on applied and developed to be used in the process industry, conventional power industry, automation and electronics. VTT develops and transfers methods and tools for reliability and safety analysis to the private and public sectors. The technology transfer takes place in joint development projects with potential users. Several computer-aided methods, such as RELVEC for reliability modelling and analysis, have been developed. The tool developed are today used by major Finnish companies in the fields of automation, nuclear power, shipbuilding and electronics. Development of computer-aided and other methods needed in analysis of operating experience, reliability or safety is further going on in a number of research and development projects

Requirements of safety and reliability

International Nuclear Information System (INIS)

Franzen, L.F.

1977-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the findings derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essential for accident analyses, and the determination of the loads occuring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig./HP) [de

78 FR 53790 - Public Forum-Safety Culture: Enhancing Transportation Safety

Science.gov (United States)

2013-08-30

... NATIONAL TRANSPORTATION SAFETY BOARD Public Forum--Safety Culture: Enhancing Transportation Safety On Tuesday and Wednesday, September 10-11, 2013, the National Transportation Safety Board (NTSB) will convene a forum titled, ``Safety Culture: Enhancing Transportation Safety.'' The forum will begin at 9:00...

Modular reliability modeling of the TJNAF personnel safety system

International Nuclear Information System (INIS)

Cinnamon, J.; Mahoney, K.

1997-01-01

A reliability model for the Thomas Jefferson National Accelerator Facility (formerly CEBAF) personnel safety system has been developed. The model, which was implemented using an Excel spreadsheet, allows simulation of all or parts of the system. Modularity os the model's implementation allows rapid open-quotes what if open-quotes case studies to simulate change in safety system parameters such as redundancy, diversity, and failure rates. Particular emphasis is given to the prediction of failure modes which would result in the failure of both of the redundant safety interlock systems. In addition to the calculation of the predicted reliability of the safety system, the model also calculates availability of the same system. Such calculations allow the user to make tradeoff studies between reliability and availability, and to target resources to improving those parts of the system which would most benefit from redesign or upgrade. The model includes calculated, manufacturer's data, and Jefferson Lab field data. This paper describes the model, methods used, and comparison of calculated to actual data for the Jefferson Lab personnel safety system. Examples are given to illustrate the model's utility and ease of use

Diversity requirements for safety critical software-based automation systems

International Nuclear Information System (INIS)

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

1998-03-01

System vendors nowadays propose software-based systems even for the most critical safety functions in nuclear power plants. Due to the nature and mechanisms of influence of software faults new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)' various safety assessment methods and tools for software based systems are developed and evaluated. This report first discusses the (common cause) failure mechanisms in software-based systems, then defines fault-tolerant system architectures to avoid common cause failures, then studies the various alternatives to apply diversity and their influence on system reliability. Finally, a method for the assessment of diversity is described. Other recently published reports in OHA-report series handles the statistical reliability assessment of software based (STUK-YTO-TR 119), usage models in reliability assessment of software-based systems (STUK-YTO-TR 128) and handling of programmable automation in plant PSA-studies (STUK-YTO-TR 129)

Conceptual Software Reliability Prediction Models for Nuclear Power Plant Safety Systems

International Nuclear Information System (INIS)

Johnson, G.; Lawrence, D.; Yu, H.

2000-01-01

The objective of this project is to develop a method to predict the potential reliability of software to be used in a digital system instrumentation and control system. The reliability prediction is to make use of existing measures of software reliability such as those described in IEEE Std 982 and 982.2. This prediction must be of sufficient accuracy to provide a value for uncertainty that could be used in a nuclear power plant probabilistic risk assessment (PRA). For the purposes of the project, reliability was defined to be the probability that the digital system will successfully perform its intended safety function (for the distribution of conditions under which it is expected to respond) upon demand with no unintended functions that might affect system safety. The ultimate objective is to use the identified measures to develop a method for predicting the potential quantitative reliability of a digital system. The reliability prediction models proposed in this report are conceptual in nature. That is, possible prediction techniques are proposed and trial models are built, but in order to become a useful tool for predicting reliability, the models must be tested, modified according to the results, and validated. Using methods outlined by this project, models could be constructed to develop reliability estimates for elements of software systems. This would require careful review and refinement of the models, development of model parameters from actual experience data or expert elicitation, and careful validation. By combining these reliability estimates (generated from the validated models for the constituent parts) in structural software models, the reliability of the software system could then be predicted. Modeling digital system reliability will also require that methods be developed for combining reliability estimates for hardware and software. System structural models must also be developed in order to predict system reliability based upon the reliability

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

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

Reliability and safety analyses under fuzziness

International Nuclear Information System (INIS)

Onisawa, T.; Kacprzyk, J.

1995-01-01

Fuzzy theory, for example possibility theory, is compatible with probability theory. What is shown so far is that probability theory needs not be replaced by fuzzy theory, but rather that the former works much better in applications if it is combined with the latter. In fact, it is said that there are two essential uncertainties in the field of reliability and safety analyses: One is a probabilistic uncertainty which is more relevant for mechanical systems and the natural environment, and the other is fuzziness (imprecision) caused by the existence of human beings in systems. The classical probability theory alone is therefore not sufficient to deal with uncertainties in humanistic system. In such a context this collection of works will put a milestone in the arguments of probability theory and fuzzy theory. This volume covers fault analysis, life time analysis, reliability, quality control, safety analysis and risk analysis. (orig./DG). 106 figs

How to interpret safety critical failures in risk and reliability assessments

International Nuclear Information System (INIS)

Selvik, Jon Tømmerås; Signoret, Jean-Pierre

2017-01-01

Management of safety systems often receives high attention due to the potential for industrial accidents. In risk and reliability literature concerning such systems, and particularly concerning safety-instrumented systems, one frequently comes across the term ‘safety critical failure’. It is a term associated with the term ‘critical failure’, and it is often deduced that a safety critical failure refers to a failure occurring in a safety critical system. Although this is correct in some situations, it is not matching with for example the mathematical definition given in ISO/TR 12489:2013 on reliability modeling, where a clear distinction is made between ‘safe failures’ and ‘dangerous failures’. In this article, we show that different interpretations of the term ‘safety critical failure’ exist, and there is room for misinterpretations and misunderstandings regarding risk and reliability assessments where failure information linked to safety systems are used, and which could influence decision-making. The article gives some examples from the oil and gas industry, showing different possible interpretations of the term. In particular we discuss the link between criticality and failure. The article points in general to the importance of adequate risk communication when using the term, and gives some clarification on interpretation in risk and reliability assessments.

Reliability assessment for safety critical systems by statistical random testing

International Nuclear Information System (INIS)

Mills, S.E.

1995-11-01

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

Reliability assessment for safety critical systems by statistical random testing

Energy Technology Data Exchange (ETDEWEB)

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

1995-11-01

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

Component reliability data for use in probabilistic safety assessment

International Nuclear Information System (INIS)

1988-10-01

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

Software reliability for safety-critical applications

International Nuclear Information System (INIS)

Everett, B.; Musa, J.

1994-01-01

In this talk, the authors address the question open-quotes Can Software Reliability Engineering measurement and modeling techniques be applied to safety-critical applications?close quotes Quantitative techniques have long been applied in engineering hardware components of safety-critical applications. The authors have seen a growing acceptance and use of quantitative techniques in engineering software systems but a continuing reluctance in using such techniques in safety-critical applications. The general case posed against using quantitative techniques for software components runs along the following lines: safety-critical applications should be engineered such that catastrophic failures occur less frequently than one in a billion hours of operation; current software measurement/modeling techniques rely on using failure history data collected during testing; one would have to accumulate over a billion operational hours to verify failure rate objectives of about one per billion hours

The establish and application of equipment reliability database in Nuclear Power Plant

International Nuclear Information System (INIS)

Zheng Wei; Li He

2006-03-01

Take the case of Daya Bay Nuclear Power Plant, the collecting and handling of equipment reliability data, the calculation method of reliability parameters and the establish and application of reliability databases, etc. are discussed. The data source involved the design information of the equipment, the operation information, the maintenance information and periodically test record, etc. Equipment reliability database built on a base of the operation experience. It provided the valid tool for thoroughly and objectively recording the operation history and the present condition of various equipment of the plant; supervising the appearance of the equipment, especially the safety-related equipment, provided the very practical worth information for enhancing the safety and availability management of the equipment and insuring the safety and economic operation of the plant; and provided the essential data for the research and applications in safety management, reliability analysis, probabilistic safety assessment, reliability centered maintenance and economic management in nuclear power plant. (authors)

Reliability Analysis Multiple Redundancy Controller for Nuclear Safety Systems

International Nuclear Information System (INIS)

Son, Gwangseop; Kim, Donghoon; Son, Choulwoong

2013-01-01

This controller is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). The architecture of MRC is briefly described, and the Markov model is developed. Based on the model, the reliability and Mean Time To Failure (MTTF) are analyzed. In this paper, the architecture of MRC for nuclear safety systems is described. The MRC is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). Markov models for MRC architecture was developed, and then the reliability was analyzed by using the model. From the reliability analyses for the MRC, it is obtained that the failure rate of each module in the MRC should be less than 2 Χ 10 -4 /hour and the MTTF average increase rate depending on FCF increment, i. e. ΔMTTF/ΔFCF, is 4 months/0.1

Quantitative software-reliability analysis of computer codes relevant to nuclear safety

International Nuclear Information System (INIS)

Mueller, C.J.

1981-12-01

This report presents the results of the first year of an ongoing research program to determine the probability of failure characteristics of computer codes relevant to nuclear safety. An introduction to both qualitative and quantitative aspects of nuclear software is given. A mathematical framework is presented which will enable the a priori prediction of the probability of failure characteristics of a code given the proper specification of its properties. The framework consists of four parts: (1) a classification system for software errors and code failures; (2) probabilistic modeling for selected reliability characteristics; (3) multivariate regression analyses to establish predictive relationships among reliability characteristics and generic code property and development parameters; and (4) the associated information base. Preliminary data of the type needed to support the modeling and the predictions of this program are described. Illustrations of the use of the modeling are given but the results so obtained, as well as all results of code failure probabilities presented herein, are based on data which at this point are preliminary, incomplete, and possibly non-representative of codes relevant to nuclear safety

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

International Nuclear Information System (INIS)

Tomic, B.; Lederman, L.

1991-01-01

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

25. MPA-seminar: safety and reliability of plant technology with special emphasis on safety and reliability - integrity proofs, qualification of components, damage prevention. Vol. 1. Papers 1-29

International Nuclear Information System (INIS)

1999-01-01

The proceedings of the 25th MPA Seminar on 'Safety and Reliability of Plant Technology' were issued in two volumes. The main topics of the first volume are: 1. Structural and safety analysis, 2. Reliability analysis, 3. Fracture mechanics, and 4. Nondestructive Testing. s

Possibilities and limitations of applying software reliability growth models to safety-critical software

International Nuclear Information System (INIS)

Kim, Man Cheol; Jang, Seung Cheol; Ha, Jae Joo

2007-01-01

It is generally known that software reliability growth models such as the Jelinski-Moranda model and the Goel-Okumoto's Non-Homogeneous Poisson Process (NHPP) model cannot be applied to safety-critical software due to a lack of software failure data. In this paper, by applying two of the most widely known software reliability growth models to sample software failure data, we demonstrate the possibility of using the software reliability growth models to prove the high reliability of safety-critical software. The high sensitivity of a piece of software's reliability to software failure data, as well as a lack of sufficient software failure data, is also identified as a possible limitation when applying the software reliability growth models to safety-critical software

Journey Toward High Reliability: A Comprehensive Safety Program to Improve Quality of Care and Safety Culture in a Large, Multisite Radiation Oncology Department.

Science.gov (United States)

Woodhouse, Kristina Demas; Volz, Edna; Maity, Amit; Gabriel, Peter E; Solberg, Timothy D; Bergendahl, Howard W; Hahn, Stephen M

2016-05-01

High-reliability organizations (HROs) focus on continuous identification and improvement of safety issues. We sought to advance a large, multisite radiation oncology department toward high reliability through the implementation of a comprehensive safety culture (SC) program at the University of Pennsylvania Department of Radiation Oncology. In 2011, with guidance from safety literature and experts in HROs, we designed an SC framework to reduce radiation errors. All state-reported medical events (SRMEs) from 2009 to 2016 were retrospectively reviewed and plotted on a control chart. Changes in SC grade were assessed using the Agency for Healthcare Research and Quality Hospital Survey. Outcomes measured included the number of radiation treatment fractions and days between SRMEs, as well as SC grade. Multifaceted safety initiatives were implemented at our main academic center and across all network sites. Postintervention results demonstrate increased staff fundamental safety knowledge, enhanced peer review with an electronic system, and special cause variation of SRMEs on control chart analysis. From 2009 to 2016, the number of days and fractions between SRMEs significantly increased, from a mean of 174 to 541 days (P safety framework. Our multifaceted initiatives, focusing on culture and system changes, can be successfully implemented in a large academic radiation oncology department to yield measurable improvements in SC and outcomes. Copyright © 2016 by American Society of Clinical Oncology.

Reliability Quantification Method for Safety Critical Software Based on a Finite Test Set

International Nuclear Information System (INIS)

Shin, Sung Min; Kim, Hee Eun; Kang, Hyun Gook; Lee, Seung Jun

2014-01-01

Software inside of digitalized system have very important role because it may cause irreversible consequence and affect the whole system as common cause failure. However, test-based reliability quantification method for some safety critical software has limitations caused by difficulties in developing input sets as a form of trajectory which is series of successive values of variables. To address these limitations, this study proposed another method which conduct the test using combination of single values of variables. To substitute the trajectory form of input using combination of variables, the possible range of each variable should be identified. For this purpose, assigned range of each variable, logical relations between variables, plant dynamics under certain situation, and characteristics of obtaining information of digital device are considered. A feasibility of the proposed method was confirmed through an application to the Reactor Protection System (RPS) software trip logic

A Secure ECC-based RFID Mutual Authentication Protocol to Enhance Patient Medication Safety.

Science.gov (United States)

Jin, Chunhua; Xu, Chunxiang; Zhang, Xiaojun; Li, Fagen

2016-01-01

Patient medication safety is an important issue in patient medication systems. In order to prevent medication errors, integrating Radio Frequency Identification (RFID) technology into automated patient medication systems is required in hospitals. Based on RFID technology, such systems can provide medical evidence for patients' prescriptions and medicine doses, etc. Due to the mutual authentication between the medication server and the tag, RFID authentication scheme is the best choice for automated patient medication systems. In this paper, we present a RFID mutual authentication scheme based on elliptic curve cryptography (ECC) to enhance patient medication safety. Our scheme can achieve security requirements and overcome various attacks existing in other schemes. In addition, our scheme has better performance in terms of computational cost and communication overhead. Therefore, the proposed scheme is well suitable for patient medication systems.

Development of web-based reliability data base platform

International Nuclear Information System (INIS)

Hwang, Seok Won; Lee, Chang Ju; Sung, Key Yong

2004-01-01

Probabilistic safety assessment (PSA) is a systematic technique which estimates the degree of risk impacts to the public due to an accident scenario. Estimating the occurrence frequencies and consequences of potential scenarios requires a thorough analysis of the accident details and all fundamental parameters. The robustness of PSA to check weaknesses in a design and operation will allow a better informed and balanced decision to be reached. The fundamental parameters for PSA, such as the component failure rates, should be estimated under the condition of steady collection of the evidence throughout the operational period. However, since any single plant data does not sufficiently enough to provide an adequate PSA result, in actual, the whole operating data was commonly used to estimate the reliability parameters for the same type of components. The reliability data of any component type consists of two categories; the generic that is based on the operating experiences of whole plants, and the plant-specific that is based on the operation of a specific plant of interest. The generic data is highly essential for new or recently-built nuclear power plants (NPPs). Generally, the reliability data base may be categorized into the component reliability, initiating event frequencies, human performance, and so on. Among these data, the component reliability seems a key element because it has the most abundant population. Therefore, the component reliability data is essential for taking a part in the quantification of accident sequences because it becomes an input of various basic events which consists of the fault tree

How to use an optimization-based method capable of balancing safety, reliability, and weight in an aircraft design process

International Nuclear Information System (INIS)

Johansson, Cristina; Derelov, Micael; Olvander, Johan

2017-01-01

In order to help decision-makers in the early design phase to improve and make more cost-efficient system safety and reliability baselines of aircraft design concepts, a method (Multi-objective Optimization for Safety and Reliability Trade-off) that is able to handle trade-offs such as system safety, system reliability, and other characteristics, for instance weight and cost, is used. Multi-objective Optimization for Safety and Reliability Trade-off has been developed and implemented at SAAB Aeronautics. The aim of this paper is to demonstrate how the implemented method might work to aid the selection of optimal design alternatives. The method is a three-step method: step 1 involves the modelling of each considered target, step 2 is optimization, and step 3 is the visualization and selection of results (results processing). The analysis is performed within Architecture Design and Preliminary Design steps, according to the company's Product Development Process. The lessons learned regarding the use of the implemented trade-off method in the three cases are presented. The results are a handful of solutions, a basis to aid in the selection of a design alternative. While the implementation of the trade-off method is performed for companies, there is nothing to prevent adapting this method, with minimal modifications, for use in other industrial applications

How to use an optimization-based method capable of balancing safety, reliability, and weight in an aircraft design process

Energy Technology Data Exchange (ETDEWEB)

Johansson, Cristina [Mendeley, Broderna Ugglasgatan, Linkoping (Sweden); Derelov, Micael; Olvander, Johan [Linkoping University, IEI, Dept. of Machine Design, Linkoping (Sweden)

2017-03-15

In order to help decision-makers in the early design phase to improve and make more cost-efficient system safety and reliability baselines of aircraft design concepts, a method (Multi-objective Optimization for Safety and Reliability Trade-off) that is able to handle trade-offs such as system safety, system reliability, and other characteristics, for instance weight and cost, is used. Multi-objective Optimization for Safety and Reliability Trade-off has been developed and implemented at SAAB Aeronautics. The aim of this paper is to demonstrate how the implemented method might work to aid the selection of optimal design alternatives. The method is a three-step method: step 1 involves the modelling of each considered target, step 2 is optimization, and step 3 is the visualization and selection of results (results processing). The analysis is performed within Architecture Design and Preliminary Design steps, according to the company's Product Development Process. The lessons learned regarding the use of the implemented trade-off method in the three cases are presented. The results are a handful of solutions, a basis to aid in the selection of a design alternative. While the implementation of the trade-off method is performed for companies, there is nothing to prevent adapting this method, with minimal modifications, for use in other industrial applications.

Prediction of safety critical software operational reliability from test reliability using testing environment factors

International Nuclear Information System (INIS)

Jung, Hoan Sung; Seong, Poong Hyun

1999-01-01

It has been a critical issue to predict the safety critical software reliability in nuclear engineering area. For many years, many researches have focused on the quantification of software reliability and there have been many models developed to quantify software reliability. Most software reliability models estimate the reliability with the failure data collected during the test assuming that the test environments well represent the operation profile. User's interest is however on the operational reliability rather than on the test reliability. The experiences show that the operational reliability is higher than the test reliability. With the assumption that the difference in reliability results from the change of environment, from testing to operation, testing environment factors comprising the aging factor and the coverage factor are developed in this paper and used to predict the ultimate operational reliability with the failure data in testing phase. It is by incorporating test environments applied beyond the operational profile into testing environment factors. The application results show that the proposed method can estimate the operational reliability accurately. (Author). 14 refs., 1 tab., 1 fig

Passive safety systems reliability and integration of these systems in nuclear power plant PSA

International Nuclear Information System (INIS)

La Lumia, V.; Mercier, S.; Marques, M.; Pignatel, J.F.

2004-01-01

Innovative nuclear reactor concepts could lead to use passive safety features in combination with active safety systems. A passive system does not need active component, external energy, signal or human interaction to operate. These are attractive advantages for safety nuclear plant improvements and economic competitiveness. But specific reliability problems, linked to physical phenomena, can conduct to stop the physical process. In this context, the European Commission (EC) starts the RMPS (Reliability Methods for Passive Safety functions) program. In this RMPS program, a quantitative reliability evaluation of the RP2 system (Residual Passive heat Removal system on the Primary circuit) has been realised, and the results introduced in a simplified PSA (Probabilistic Safety Assessment). The scope is to get out experience of definition of characteristic parameters for reliability evaluation and PSA including passive systems. The simplified PSA, using event tree method, is carried out for the total loss of power supplies initiating event leading to a severe core damage. Are taken into account: failures of components but also failures of the physical process involved (e.g. natural convection) by a specific method. The physical process failure probabilities are assessed through uncertainty analyses based on supposed probability density functions for the characteristic parameters of the RP2 system. The probabilities are calculated by MONTE CARLO simulation coupled to the CATHARE thermalhydraulic code. The yearly frequency of the severe core damage is evaluated for each accident sequence. This analysis has identified the influence of the passive system RP2 and propose a re-dimensioning of the RP2 system in order to satisfy the safety probabilistic objectives for reactor core severe damage. (authors)

Microprocessor hardware reliability

Energy Technology Data Exchange (ETDEWEB)

Wright, R I

1982-01-01

Microprocessor-based technology has had an impact in nearly every area of industrial electronics and many applications have important safety implications. Microprocessors are being used for the monitoring and control of hazardous processes in the chemical, oil and power generation industries, for the control and instrumentation of aircraft and other transport systems and for the control of industrial machinery. Even in the field of nuclear reactor protection, where designers are particularly conservative, microprocessors are used to implement certain safety functions and may play increasingly important roles in protection systems in the future. Where microprocessors are simply replacing conventional hard-wired control and instrumentation systems no new hazards are created by their use. In the field of robotics, however, the microprocessor has opened up a totally new technology and with it has created possible new and as yet unknown hazards. The paper discusses some of the design and manufacturing techniques which may be used to enhance the reliability of microprocessor based systems and examines the available reliability data on lsi/vlsi microcircuits. 12 references.

Design and reliability, availability, maintainability, and safety analysis of a high availability quadruple vital computer system

Institute of Scientific and Technical Information of China (English)

Ping TAN; Wei-ting HE; Jia LIN; Hong-ming ZHAO; Jian CHU

2011-01-01

With the development of high-speed railways in China,more than 2000 high-speed trains will be put into use.Safety and efficiency of railway transportation is increasingly important.We have designed a high availability quadruple vital computer (HAQVC) system based on the analysis of the architecture of the traditional double 2-out-of-2 system and 2-out-of-3 system.The HAQVC system is a system with high availability and safety,with prominent characteristics such as fire-new internal architecture,high efficiency,reliable data interaction mechanism,and operation state change mechanism.The hardware of the vital CPU is based on ARM7 with the real-time embedded safe operation system (ES-OS).The Markov modeling method is designed to evaluate the reliability,availability,maintainability,and safety (RAMS) of the system.In this paper,we demonstrate that the HAQVC system is more reliable than the all voting triple modular redundancy (AVTMR) system and double 2-out-of-2 system.Thus,the design can be used for a specific application system,such as an airplane or high-speed railway system.

Britain's Nuclear Electric pursues a major programme to enhance safety

International Nuclear Information System (INIS)

Anon.

1991-01-01

The UK state-owned nuclear utility, Nuclear Electric, is actively pursuing a major initiative inherited from its predecessor the Central Electricity Generating Board, known as the Station Enhancement Programme. The Enhancement Programme is a multi-million pound project and covers all Nuclear Electric's stations (including relevant aspects of the de-commissioning site at Berkeley). Its fundamental objective is to enhance the arrangements, practices and attitudes on the power stations to create an overall environment likely to promote ever safer and more reliable operation, and to take full account of the best international practices. The four year programme started in the spring of 1989, and is expected to be completed in 1993. Key elements of the programme include quality assurance based management systems; improvements to site and plant material conditions; a more formal approach to training; additional resources and controls to aid the management of maintenance defects; higher profile for the feedback of operational experience; management targets; a station evaluation programme; and formal feedback of lessons learned from the IAEA operational and safety review team visit to the Oldbury-on-Severn station in 1989. (author)

A survey on reliability and safety analysis techniques of robot systems in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

The reliability and safety analysis techniques was surveyed for the purpose of overall quality improvement of reactor inspection system which is under development in our current project. The contents of this report are : 1. Reliability and safety analysis techniques suvey - Reviewed reliability and safety analysis techniques are generally accepted techniques in many industries including nuclear industry. And we selected a few techniques which are suitable for our robot system. They are falut tree analysis, failure mode and effect analysis, reliability block diagram, markov model, combinational method, and simulation method. 2. Survey on the characteristics of robot systems which are distinguished from other systems and which are important to the analysis. 3. Survey on the nuclear environmental factors which affect the reliability and safety analysis of robot system 4. Collection of the case studies of robot reliability and safety analysis which are performed in foreign countries. The analysis results of this survey will be applied to the improvement of reliability and safety of our robot system and also will be used for the formal qualification and certification of our reactor inspection system.

A survey on reliability and safety analysis techniques of robot systems in nuclear power plants

International Nuclear Information System (INIS)

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

2000-12-01

The reliability and safety analysis techniques was surveyed for the purpose of overall quality improvement of reactor inspection system which is under development in our current project. The contents of this report are : 1. Reliability and safety analysis techniques suvey - Reviewed reliability and safety analysis techniques are generally accepted techniques in many industries including nuclear industry. And we selected a few techniques which are suitable for our robot system. They are falut tree analysis, failure mode and effect analysis, reliability block diagram, markov model, combinational method, and simulation method. 2. Survey on the characteristics of robot systems which are distinguished from other systems and which are important to the analysis. 3. Survey on the nuclear environmental factors which affect the reliability and safety analysis of robot system 4. Collection of the case studies of robot reliability and safety analysis which are performed in foreign countries. The analysis results of this survey will be applied to the improvement of reliability and safety of our robot system and also will be used for the formal qualification and certification of our reactor inspection system

Component fragility data base for reliability and probability studies

International Nuclear Information System (INIS)

Bandyopadhyay, K.; Hofmayer, C.; Kassier, M.; Pepper, S.

1989-01-01

Safety-related equipment in a nuclear plant plays a vital role in its proper operation and control, and failure of such equipment due to an earthquake may pose a risk to the safe operation of the plant. Therefore, in order to assess the overall reliability of a plant, the reliability of performance of the equipment should be studied first. The success of a reliability or a probability study depends to a great extent on the data base. To meet this demand, Brookhaven National Laboratory (BNL) has formed a test data base relating the seismic capacity of equipment specimens to the earthquake levels. Subsequently, the test data have been analyzed for use in reliability and probability studies. This paper describes the data base and discusses the analysis methods. The final results that can be directly used in plant reliability and probability studies are also presented in this paper

Technology development of maintenance optimization and reliability analysis for safety features in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Woon; Choi, Seong Soo; Lee, Dong Gue; Kim, Young Il

1999-12-01

The reliability data management system (RDMS) for safety systems of PHWR type plants has been developed and utilized in the reliability analysis of the special safety systems of Wolsong Unit 1,2 with plant overhaul period lengthened. The RDMS is developed for the periodic efficient reliability analysis of the safety systems of Wolsong Unit 1,2. In addition, this system provides the function of analyzing the effects on safety system unavailability if the test period of a test procedure changes as well as the function of optimizing the test periods of safety-related test procedures. The RDMS can be utilized in handling the requests of the regulatory institute actively with regard to the reliability validation of safety systems. (author)

Optimized work control process to improve safety and reliability in a risk-based and deregulated environment

International Nuclear Information System (INIS)

Anderson, Jon G.; Jeffries, Jeffrey D. E.; Mairs, Todd P.; Rahn, Frank J.

1999-01-01

This paper provides an overview of strategic models to assist power generating plants to improve their work control processes. These models include mechanisms to continually keep the process up to date. Included in the work control process are elements for system cost/performance analysis, life-cycle maintenance planning, on-line scheduling and look-ahead techniques, and schedule implementation to conduct work on the asset. The paper also discusses how risk management associated with work control issues that effect the safety and reliability, as well as O and M costs, is integrated into this strategy. The work control process is a pervasive and critical element in the successful implementation of operations and work management programs. While providing a method to implement maintenance activities in a cost-effective manner, the work control process improves plant safety and system reliability

Enhancing NPP Safety Through an Effective Dependability Management

Energy Technology Data Exchange (ETDEWEB)

Vieru, G., E-mail: g_vieru@yahoo.com [AREN, Bucharest (Romania)

2014-10-15

Taking into account the importance of the continuous improvement of the performance and reliability of a NPP and practical measures to strengthen nuclear safety and security, it is to be noted that a good management for a nuclear power reactor involves a ''good dependability management'' of the activities, such as: Reliability, Availability, Maintainability (RAM) and maintenance support. In order to evaluate certain safety assessment criteria intended to be applied at the level of the nuclear reactor unit management, equipment dependability indicators and their impact over the availability and reactor safety have to be evaluated. Reactor equipment dependability indicators provide a quantitative indication of equipment RAM performances (Reliability, Availability and Maintenance). One of the important benefits of maintenance and failure data gathering is that it can be used as a support of probabilistic safety assessment (PSA). Also, a good dependability management implementation may be used to complement reactor level unit performance indicators in the field of safe operation, maintenance and improving operating parameters, as well as for Strengthening Safety and Improving Reliability of a NPP. This paper underlines the importance of nuclear safety and security as prerequisites for nuclear power. In addition, it demonstrates how different technical aspects, through implementation of a good dependability management, contribute to a strengthened safety and an improvement of availability of the NPP through dependability indicators determination and evaluation. (author)

Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos

2009-01-01

Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: vasconv@cdtn.br, e-mail: silvaem@cdtn.br, e-mail: aclc@cdtn.br, e-mail: reissc@cdtn.br

2009-07-01

Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

How could intelligent safety transport systems enhance safety ?

NARCIS (Netherlands)

Wiethoff, M. Heijer, T. & Bekiaris, E.

2017-01-01

In Europe, many deaths and injured each years are the cost of today's road traffic. Therefore, it is wise to look for possible solutions for enhancing traffic safety. Some Advanced Driver Assistance Systems (ADAS) are expected to increase safety, but they may also evoke new safety hazards. Only

Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2000-01-01

Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

Development of reliability-based design and assessment standards for onshore gas transmission pipelines

Energy Technology Data Exchange (ETDEWEB)

Zhou, Joe; Rothwell, Brian [TransCanada PipeLines Ltd., Calgary, AB (Canada); Nessim, Maher; Zhou, Wenxing [C-FER Technologies, Edmonton, AB (Canada)

2005-07-01

Onshore pipelines have traditionally been designed with a deterministic stress based methodology. The changing operating environment has however imposed many challenges to the pipeline industry, including heightened public awareness of risk, more challenging natural hazards and increased economic competitiveness. To meet the societal expectation of pipeline safety and enhance the competitiveness of the pipeline industry, significant efforts have been spent for the development of reliability-based design and assessment (RBDA) methodology. This paper will briefly review the technology development in the RBDA area and the focus will be on the progresses in the past years in standard development within the American Society of Mechanical Engineers (ASME) and the Canadian Standard Association (CSA) organizations. (author)

Reliability and Maintainability Engineering - A Major Driver for Safety and Affordability

Science.gov (United States)

Safie, Fayssal M.

2011-01-01

The United States National Aeronautics and Space Administration (NASA) is in the midst of an effort to design and build a safe and affordable heavy lift vehicle to go to the moon and beyond. To achieve that, NASA is seeking more innovative and efficient approaches to reduce cost while maintaining an acceptable level of safety and mission success. One area that has the potential to contribute significantly to achieving NASA safety and affordability goals is Reliability and Maintainability (R&M) engineering. Inadequate reliability or failure of critical safety items may directly jeopardize the safety of the user(s) and result in a loss of life. Inadequate reliability of equipment may directly jeopardize mission success. Systems designed to be more reliable (fewer failures) and maintainable (fewer resources needed) can lower the total life cycle cost. The Department of Defense (DOD) and industry experience has shown that optimized and adequate levels of R&M are critical for achieving a high level of safety and mission success, and low sustainment cost. Also, lessons learned from the Space Shuttle program clearly demonstrated the importance of R&M engineering in designing and operating safe and affordable launch systems. The Challenger and Columbia accidents are examples of the severe impact of design unreliability and process induced failures on system safety and mission success. These accidents demonstrated the criticality of reliability engineering in understanding component failure mechanisms and integrated system failures across the system elements interfaces. Experience from the shuttle program also shows that insufficient Reliability, Maintainability, and Supportability (RMS) engineering analyses upfront in the design phase can significantly increase the sustainment cost and, thereby, the total life cycle cost. Emphasis on RMS during the design phase is critical for identifying the design features and characteristics needed for time efficient processing

Fusion reactor passive safety and ignitor risk-based regulation

International Nuclear Information System (INIS)

Zucchetti, M.

1995-01-01

Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine

Enhancing operational nuclear safety

International Nuclear Information System (INIS)

Sengoku, Katsuhisa

2008-01-01

Since Chernobyl, the dictum A n accident anywhere is an accident everywhere i s a globally shared perception. The paper presents challenges to the international nuclear community: globalization, sustainable and dynamic development, secure, safe and clean energy supply, nuclear r enaissance , public concern for nuclear safety, nuclear security, and technology and management. Strong national safety infrastructures and international cooperation are required to maintain a high level of nuclear safety and security worldwide. There is an increasing number of countries thinking of going nuclear: Morocco, Indonesia, Iran, Poland, Turkey, Bangladesh, Egypt, Vietnam, Chile, Nigeria, Malaysia, Thailand, Uruguay, Tunisia, Algeria. Another serious incident will jeopardize the prospect of nuclear renaissance. Safety and security are preconditions for countries newly introducing NPP as well as for those with mature nuclear programmes. The Global Nuclear Safety Regime (GNSR) is referred to as the institutional, legal and technical framework to achieve worldwide implementation of the safety of nuclear installations. At the top of the framework is the Convention on Nuclear Safety which covers the nuclear power plants. The convention has 56 contracting parties which meet triennially where national reports are presented and subject to the review of peers. The International Atomic Energy Agency (IAEA) undertakes a programme to foster the GNSR through the establishment of IAEA safety standards and related publications. The programme provides for the application of standards for the (1) safety of nuclear installations, (2) safety of radioactive sources, (3) safe transport of radioactive material and (4) management of radioactive waste. It also provides for the security of nuclear installations, nuclear material and radioactive material. The safety standards hierarchy is as follows: safety fundamental, safety requirements and safety guides. The safety fundamentals are the bases for IAEA

Reliability-based evaluation of bridge components for consistent safety margins.

Science.gov (United States)

2010-10-01

The Load and Resistant Factor Design (LRFD) approach is based on the concept of structural reliability. The approach is more : rational than the former design approaches such as Load Factor Design or Allowable Stress Design. The LRFD Specification fo...

Enhancing operational safety

Energy Technology Data Exchange (ETDEWEB)

Wiebe, J S

1997-09-01

The presentation briefly considers the following aspects concerning enhancing operational safety of NPP: licensed control room supervision, reactivity changes, personnel access to control room, simulator training.

Reliability and safety of functional capacity evaluation in patients with whiplash associated disorders.

Science.gov (United States)

Trippolini, M A; Reneman, M F; Jansen, B; Dijkstra, P U; Geertzen, J H B

2013-09-01

Whiplash-associated disorders (WAD) are a burden for both individuals and society. It is recommended to evaluate patients with WAD at risk of chronification to enhance rehabilitation and promote an early return to work. In patients with low back pain (LBP), functional capacity evaluation (FCE) contributes to clinical decisions regarding fitness-for-work. FCE should have demonstrated sufficient clinimetric properties. Reliability and safety of FCE for patients with WAD is unknown. Thirty-two participants (11 females and 21 males; mean age 39.6 years) with WAD (Grade I or II) were included. The FCE consisted of 12 tests, including material handling, hand grip strength, repetitive arm movements, static arm activities, walking speed, and a 3 min step test. Overall the FCE duration was 60 min. The test-retest interval was 7 days. Interclass correlations (model 1) (ICCs) and limits of agreement (LoA) were calculated. Safety was assessed by a Pain Response Questionnaire, observation criteria and heart rate monitoring. ICCs ranged between 0.57 (3 min step test) and 0.96 (short two-handed carry). LoA relative to mean performance ranged between 15 % (50 m walking test) and 57 % (lifting waist to overhead). Pain reactions after WAD FCE decreased within days. Observations and heart rate measurements fell within the safety criteria. The reliability of the WAD FCE was moderate in two tests, good in five tests and excellent in five tests. Safety-criteria were fulfilled. Interpretation at the patient level should be performed with care because LoA were substantial.

Development of seismic technology and reliability based on vibration tests

International Nuclear Information System (INIS)

Sasaki, Youichi

1997-01-01

This paper deals with some of the vibration tests and investigations on the seismic safety of nuclear power plants (NPPs) in Japan. To ensure the reliability of the seismic safety of nuclear power plants, nuclear power plants in Japan have been designed according to the Technical Guidelines for Aseismic Design of Nuclear Power Plants. This guideline has been developed based on technical date base and findings which were obtained from many vibration tests and investigations. Besides the tests for the guideline, proving tests on seismic reliability of operating nuclear power plants equipment and systems have been carried out. In this paper some vibration tests and their evaluation results are presented. They have crucially contributed to develop the guideline. (J.P.N.)

Development of seismic technology and reliability based on vibration tests

Energy Technology Data Exchange (ETDEWEB)

Sasaki, Youichi [Nuclear Power Engineering Corp., Tokyo (Japan)

1997-03-01

This paper deals with some of the vibration tests and investigations on the seismic safety of nuclear power plants (NPPs) in Japan. To ensure the reliability of the seismic safety of nuclear power plants, nuclear power plants in Japan have been designed according to the Technical Guidelines for Aseismic Design of Nuclear Power Plants. This guideline has been developed based on technical date base and findings which were obtained from many vibration tests and investigations. Besides the tests for the guideline, proving tests on seismic reliability of operating nuclear power plants equipment and systems have been carried out. In this paper some vibration tests and their evaluation results are presented. They have crucially contributed to develop the guideline. (J.P.N.)

Reliability of thermal-hydraulic passive safety systems

International Nuclear Information System (INIS)

D'Auria, F.; Araneo, D.; Pierro, F.; Galassi, G.

2014-01-01

The scholar will be informed of reliability concepts applied to passive system adopted for nuclear reactors. Namely, for classical components and systems the failure concept is associated with malfunction of breaking of hardware. In the case of passive systems the failure is associated with phenomena. A method for studying the reliability of passive systems is discussed and is applied. The paper deals with the description of the REPAS (Reliability Evaluation of Passive Safety System) methodology developed by University of Pisa (UNIPI) and with results from its application. The general objective of the REPAS methodology is to characterize the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

2014-01-01

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Pump performance and reliability follow-up by the French Safety Authorities

International Nuclear Information System (INIS)

Clausner, J.P.; De La Ronciere, X.; Scott de Martinville, E.; Courbiere, P.

1990-12-01

This paper will present, through actual examples, the methodology of the performance and reliability safety-related pumps evaluation applied by the French Safety Authorities and the lessons drawn from this evaluation

A knowledge-based operator advisor system for integration of fault detection, control, and diagnosis to enhance the safe and reliable operation of nuclear power plants

International Nuclear Information System (INIS)

Bhatnagar, R.

1989-01-01

A Knowledged-Based Operator Advisor System has been developed for enhancing the complex task of maintaining safe and reliable operation of nuclear power plants. The operator's activities have been organized into the four tasks of data interpretation for abstracting high level information from sensor data, plant state monitoring for identification of faults, plan execution for controlling the faults, and diagnosis for determination of root causes of faults. The Operator Advisor System is capable of identifying the abnormal functioning of the plant in terms of: (1) deviations from normality, (2) pre-enumerated abnormal events, and (3) safety threats. The classification of abnormal functioning into the three categories of deviations from normality, abnormal events, and safety threats allows the detection of faults at three levels of: (1) developing faults, (2) developed faults, and (3) safety threatening faults. After the identification of abnormal functioning the system will identify the procedures to be executed to mitigate the consequences of abnormal functioning and will help the operator by displaying the procedure steps and monitoring the success of actions taken. The system also is capable of diagnosing the root causes of abnormal functioning. The identification, and diagnosis of root causes of abnormal functioning are done in parallel to the task of procedure execution, allowing the detection of more critical safety threats while executing procedures to control abnormal events

Comparative analysis of different configurations of PLC-based safety systems from reliability point of view

Science.gov (United States)

Tapia, Moiez A.

1993-01-01

The study of a comparative analysis of distinct multiplex and fault-tolerant configurations for a PLC-based safety system from a reliability point of view is presented. It considers simplex, duplex and fault-tolerant triple redundancy configurations. The standby unit in case of a duplex configuration has a failure rate which is k times the failure rate of the standby unit, the value of k varying from 0 to 1. For distinct values of MTTR and MTTF of the main unit, MTBF and availability for these configurations are calculated. The effect of duplexing only the PLC module or only the sensors and the actuators module, on the MTBF of the configuration, is also presented. The results are summarized and merits and demerits of various configurations under distinct environments are discussed.

The safety monitor and RCM workstation as complementary tools in risk based maintenance optimization

International Nuclear Information System (INIS)

Rawson, P.D.

2000-01-01

Reliability Centred Maintenance (RCM) represents a proven technique for rendering maintenance activities safer, more effective, and less expensive, in terms of systems unavailability and resource management. However, it is believed that RCM can be enhanced by the additional consideration of operational plant risk. This paper discusses how two computer-based tools, i.e., the RCM Workstation and the Safety Monitor, can complement each other in helping to create a living preventive maintenance strategy. (author)

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 Calculations

DEFF Research Database (Denmark)

Petersen, Kurt Erling

1986-01-01

Risk and reliability analysis is increasingly being used in evaluations of plant safety and plant reliability. The analysis can be performed either during the design process or during the operation time, with the purpose to improve the safety or the reliability. Due to plant complexity and safety...... and availability requirements, sophisticated tools, which are flexible and efficient, are needed. Such tools have been developed in the last 20 years and they have to be continuously refined to meet the growing requirements. Two different areas of application were analysed. In structural reliability probabilistic...... approaches have been introduced in some cases for the calculation of the reliability of structures or components. A new computer program has been developed based upon numerical integration in several variables. In systems reliability Monte Carlo simulation programs are used especially in analysis of very...

Jefferson Lab IEC 61508/61511 Safety PLC Based Safety System

International Nuclear Information System (INIS)

Mahoney, Kelly; Robertson, Henry

2009-01-01

This paper describes the design of the new 12 GeV Upgrade Personnel Safety System (PSS) at the Thomas Jefferson National Accelerator Facility (TJNAF). The new PSS design is based on the implementation of systems designed to meet international standards IEC61508 and IEC 61511 for programmable safety systems. In order to meet the IEC standards, TJNAF engineers evaluated several SIL 3 Safety PLCs before deciding on an optimal architecture. In addition to hardware considerations, software quality standards and practices must also be considered. Finally, we will discuss R and D that may lead to both high safety reliability and high machine availability that may be applicable to future accelerators such as the ILC.

The importance of the reliability study for the safety operation of chemical plants. Application in heavy water plants

International Nuclear Information System (INIS)

Dumitrescu, Maria; Lazar, Roxana Elena; Preda, Irina Aida; Stefanescu, Ioan

1999-01-01

Heavy water production in Romania is based on H 2 O-H 2 S isotopic exchange process followed by vacuum isotopic distillation. The heavy water plant are complex chemical systems, characterized by an ensemble of static and dynamic equipment, AMC components, enclosures. Such equipment must have a high degree of reliability, a maximum safety in technological operation and a high availability index. Safety, reliable and economical operation heavy water plants need to maintain the systems and the components at adequate levels of reliability. The paper is a synthesis of the qualitative and quantitative assessment reliability studies for heavy water plants. The operation analysis on subsystems, each subsystems being a well-defined unit, is required by the plant complexity. For each component the reliability indicators were estimated by parametric and non-parametric methods based on the plant operation data. Also, the reliability qualitative and quantitative assessment was done using the fault tree technique. For the dual temperature isotopic exchange plants the results indicate an increase of the MTBF after the first years of operation, illustrating both the operation experience increasing and maintenance improvement. Also a high degree of availability was illustrated by the reliability studies of the vacuum distillation plant. The establishment of the reliability characteristics for heavy water plant represents an important step, a guide for highlighting the elements and process liable to failure being at the same time a planning modality to correlate the control times with the maintenance operations. This is the way to minimise maintenance, control and costs. The main purpose of the reliability study was the safety increase of the plant operation and the support for decision making. (authors)

Safety Enhancements for PHWRs Based on Macroscopic Losses of the Fukushima Accident

Directory of Open Access Journals (Sweden)

Sang Ho Kim

2015-01-01

Full Text Available The role of nuclear energy is to supply electric power on a stable basis to meet increasing demands, reduce carbon dioxide emissions, and maintain stable electric power costs while ensuring safety. The Fukushima accident taught us many lessons for creating safer nuclear power plants. Considering the design of systems, the areas of weakness at the Fukushima nuclear power plants can be divided into three categories: plant protection, electricity supply, and cooling of the nuclear fuel. In this paper, focusing on these three areas, the lessons learned are proposed and applied for pressurized heavy water reactors. Firstly, hard protection against external risks ensures the integrity of components and systems such that they can perform their original functions. Secondly, additional emergency power supply systems for electrical redundancy and diversity can improve the response capabilities for an accident by increasing the availability of active components. Thirdly, cooling for removing decay heat can be augmented by adopting diverse safety systems derived from other types of reactors. This study is expected to contribute to the safety enhancement of pressurized heavy water reactors by applying design changes based on the lessons learned from the Fukushima accident.

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

Condition-based fault tree analysis (CBFTA): A new method for improved fault tree analysis (FTA), reliability and safety calculations

International Nuclear Information System (INIS)

Shalev, Dan M.; Tiran, Joseph

2007-01-01

Condition-based maintenance methods have changed systems reliability in general and individual systems in particular. Yet, this change does not affect system reliability analysis. System fault tree analysis (FTA) is performed during the design phase. It uses components failure rates derived from available sources as handbooks, etc. Condition-based fault tree analysis (CBFTA) starts with the known FTA. Condition monitoring (CM) methods applied to systems (e.g. vibration analysis, oil analysis, electric current analysis, bearing CM, electric motor CM, and so forth) are used to determine updated failure rate values of sensitive components. The CBFTA method accepts updated failure rates and applies them to the FTA. The CBFTA recalculates periodically the top event (TE) failure rate (λ TE ) thus determining the probability of system failure and the probability of successful system operation-i.e. the system's reliability. FTA is a tool for enhancing system reliability during the design stages. But, it has disadvantages, mainly it does not relate to a specific system undergoing maintenance. CBFTA is tool for updating reliability values of a specific system and for calculating the residual life according to the system's monitored conditions. Using CBFTA, the original FTA is ameliorated to a practical tool for use during the system's field life phase, not just during system design phase. This paper describes the CBFTA method and its advantages are demonstrated by an example

Maintenance management of railway infrastructures based on reliability analysis

International Nuclear Information System (INIS)

Macchi, Marco; Garetti, Marco; Centrone, Domenico; Fumagalli, Luca; Piero Pavirani, Gian

2012-01-01

Railway infrastructure maintenance plays a crucial role for rail transport. It aims at guaranteeing safety of operations and availability of railway tracks and related equipment for traffic regulation. Moreover, it is one major cost for rail transport operations. Thus, the increased competition in traffic market is asking for maintenance improvement, aiming at the reduction of maintenance expenditures while keeping the safety of operations. This issue is addressed by the methodology presented in the paper. The first step of the methodology consists of a family-based approach for the equipment reliability analysis; its purpose is the identification of families of railway items which can be given the same reliability targets. The second step builds the reliability model of the railway system for identifying the most critical items, given a required service level for the transportation system. The two methods have been implemented and tested in practical case studies, in the context of Rete Ferroviaria Italiana, the Italian public limited company for railway transportation.

Safety, reliability and worker satisfaction during organizational change

NARCIS (Netherlands)

Zwetsloot, G.I.J.M.; Drupsteen, L.; Vroome, E.M.M. de

2014-01-01

The research presented in this paper was carried out in four process industry plants in the Netherlands, to identify factors that have the potential to increase safety and reliability while maintaining or improving job satisfaction. The data used were gathered as part of broader trajectories in

Increased nuclear safety and reliability through power beaming

International Nuclear Information System (INIS)

Coomes, E.P.; Widrig, R.D.

1989-01-01

Space satellites and platforms currently include self-contained power systems to supply the energy necessary to accomplish mission objectives. With power beaming, the power system is separate from the satellite and the two are connected by an energy beam. This approach is analogous to earth-based central station power generation and distribution over transmission lines to various customers. In space, power is produced by power satellites (central power generating stations) and transmitted via energy beams to individual users. Power beaming has the ability to provide an order of magnitude increase in power availability over solar-based power systems with less mass on orbit. The technologies needed for power beaming are being developed today under existing programs directed by the Strategic Defense Initiative Office, the National Aeronautics and Space Administration, and the US Department of Energy. A space power architecture based on power beaming would greatly increase the safety and reliability of employing nuclear power in space

Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

Energy Technology Data Exchange (ETDEWEB)

Lowe, Andrew; Hayward, Brent (Dedale Asia Pacific, Albert Park VIC 3206 (Australia))

2006-08-15

This report documents a study with the objective of enhancing safety culture in the Swedish nuclear power industry. A primary objective of this study was to ensure that the latest thinking on human factors principles was being recognised and applied by nuclear power operators as a means of ensuring optimal safety performance. The initial phase of the project was conducted as a pilot study, involving the senior management group at one Swedish nuclear power-producing site. The pilot study enabled the project methodology to be validated after which it was repeated at other Swedish nuclear power industry sites, providing a broad-ranging analysis of opportunities across the industry to enhance safety culture. The introduction to this report contains an overview of safety culture, explains the background to the project and sets out the project rationale and objectives. The methodology used for understanding and analysing the important safety culture issues at each nuclear power site is then described. This section begins with a summary of the processes used in the information gathering and data analysis stage. The six components of the Management Workshops conducted at each site are then described. These workshops used a series of presentations, interactive events and group exercises to: (a) provide feedback to site managers on the safety culture and safety leadership issues identified at their site, and (b) stimulate further safety thinking and provide 'take-away' information and leadership strategies that could be applied to promote safety culture improvements. Section 3, project Findings, contains the main observations and output from the project. These include: - a brief overview of aspects of the local industry operating context that impinge on safety culture; - a summary of strengths or positive attributes observed within the safety culture of the Swedish nuclear industry; - a set of identified opportunities for further improvement; - the aggregated

Mochovce NPP safety measures evaluation from point of view of operational safety enhancement

International Nuclear Information System (INIS)

Cillik, I.; Vrtik, L.

2000-01-01

Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing. (author)

Reliability-based performance simulation for optimized pavement maintenance

International Nuclear Information System (INIS)

Chou, Jui-Sheng; Le, Thanh-Son

2011-01-01

Roadway pavement maintenance is essential for driver safety and highway infrastructure efficiency. However, regular preventive maintenance and rehabilitation (M and R) activities are extremely costly. Unfortunately, the funds available for the M and R of highway pavement are often given lower priority compared to other national development policies, therefore, available funds must be allocated wisely. Maintenance strategies are typically implemented by optimizing only the cost whilst the reliability of facility performance is neglected. This study proposes a novel algorithm using multi-objective particle swarm optimization (MOPSO) technique to evaluate the cost-reliability tradeoff in a flexible maintenance strategy based on non-dominant solutions. Moreover, a probabilistic model for regression parameters is employed to assess reliability-based performance. A numerical example of a highway pavement project is illustrated to demonstrate the efficacy of the proposed MOPSO algorithms. The analytical results show that the proposed approach can help decision makers to optimize roadway maintenance plans. - Highlights: →A novel algorithm using multi-objective particle swarm optimization technique. → Evaluation of the cost-reliability tradeoff in a flexible maintenance strategy. → A probabilistic model for regression parameters is employed to assess reliability-based performance. → The proposed approach can help decision makers to optimize roadway maintenance plans.

Reliability-based performance simulation for optimized pavement maintenance

Energy Technology Data Exchange (ETDEWEB)

Chou, Jui-Sheng, E-mail: jschou@mail.ntust.edu.tw [Department of Construction Engineering, National Taiwan University of Science and Technology (Taiwan Tech), 43 Sec. 4, Keelung Rd., Taipei 106, Taiwan (China); Le, Thanh-Son [Department of Construction Engineering, National Taiwan University of Science and Technology (Taiwan Tech), 43 Sec. 4, Keelung Rd., Taipei 106, Taiwan (China)

2011-10-15

Roadway pavement maintenance is essential for driver safety and highway infrastructure efficiency. However, regular preventive maintenance and rehabilitation (M and R) activities are extremely costly. Unfortunately, the funds available for the M and R of highway pavement are often given lower priority compared to other national development policies, therefore, available funds must be allocated wisely. Maintenance strategies are typically implemented by optimizing only the cost whilst the reliability of facility performance is neglected. This study proposes a novel algorithm using multi-objective particle swarm optimization (MOPSO) technique to evaluate the cost-reliability tradeoff in a flexible maintenance strategy based on non-dominant solutions. Moreover, a probabilistic model for regression parameters is employed to assess reliability-based performance. A numerical example of a highway pavement project is illustrated to demonstrate the efficacy of the proposed MOPSO algorithms. The analytical results show that the proposed approach can help decision makers to optimize roadway maintenance plans. - Highlights: > A novel algorithm using multi-objective particle swarm optimization technique. > Evaluation of the cost-reliability tradeoff in a flexible maintenance strategy. > A probabilistic model for regression parameters is employed to assess reliability-based performance. > The proposed approach can help decision makers to optimize roadway maintenance plans.

Feasibility of AmbulanCe-Based Telemedicine (FACT) Study : Safety, Feasibility and Reliability of Third Generation Ambulance Telemedicine

NARCIS (Netherlands)

Yperzeele, Laetitia; Van Hooff, Robbert-Jan; De Smedt, Ann; Espinoza, Alexis Valenzuela; Van Dyck, Rita; Van de Casseye, Rohny; Convents, Andre; Hubloue, Ives; Lauwaert, Door; De Keyser, Jacques; Brouns, Raf

2014-01-01

Background: Telemedicine is currently mainly applied as an in-hospital service, but this technology also holds potential to improve emergency care in the prehospital arena. We report on the safety, feasibility and reliability of in-ambulance teleconsultation using a telemedicine system of the third

Enhancing Experiment Central Service Reliability: from delivery to security and virtualization

CERN Document Server

Donno, Flavia; Buzykaev, Alexey; Saiz Santos, Maria Dolores

2011-01-01

The four LHC experiments rely on experiment specific services running on machines mainly located at CERN. Some of these services have been rated by the experiments as very critical: any loss or degradation of performance has a major impact on the experiment's production and analysis activities. It is therefore important to provide a reliable and robust operational environment. In this work we describe the strategy based on service deployment, security and virtualization adopted to enhance the reliability of ATLAS and CMS central services.

Safety parameter display system: an operator support system for enhancement of safety in Indian PHWRs

International Nuclear Information System (INIS)

Subramaniam, K.; Biswas, T.

1994-01-01

Ensuring operational safety in nuclear power plants is important as operator errors are observed to contribute significantly to the occurrence of accidents. Computerized operator support systems, which process and structure information, can help operators during both normal and transient conditions, and thereby enhance safety and aid effective response to emergency conditions. An important operator aid being developed and described in this paper, is the safety parameter display system (SPDS). The SPDS is an event-independent, symptom-based operator aid for safety monitoring. Knowledge-based systems can provide operators with an improved quality of information. An information processing model of a knowledge based operator support system (KBOSS) developed for emergency conditions using an expert system shell is also presented. The paper concludes with a discussion of the design issues involved in the use of a knowledge based systems for real time safety monitoring and fault diagnosis. (author). 8 refs., 4 figs., 1 tab

Reliability–based economic model predictive control for generalised flow–based networks including actuators’ health–aware capabilities

Directory of Open Access Journals (Sweden)

Grosso Juan M.

2016-09-01

Full Text Available This paper proposes a reliability-based economic model predictive control (MPC strategy for the management of generalised flow-based networks, integrating some ideas on network service reliability, dynamic safety stock planning, and degradation of equipment health. The proposed strategy is based on a single-layer economic optimisation problem with dynamic constraints, which includes two enhancements with respect to existing approaches. The first enhancement considers chance-constraint programming to compute an optimal inventory replenishment policy based on a desired risk acceptability level, leading to dynamical allocation of safety stocks in flow-based networks to satisfy non-stationary flow demands. The second enhancement computes a smart distribution of the control effort and maximises actuators’ availability by estimating their degradation and reliability. The proposed approach is illustrated with an application of water transport networks using the Barcelona network as the case study considered.

Dynamic reliability of digital-based transmitters

Energy Technology Data Exchange (ETDEWEB)

Brissaud, Florent, E-mail: florent.brissaud.2007@utt.f [Institut National de l' Environnement Industriel et des Risques (INERIS), Parc Technologique Alata, BP 2, 60550 Verneuil-en-Halatte (France) and Universite de Technologie de Troyes - UTT, Institut Charles Delaunay - ICD and UMR CNRS 6279 STMR, 12 rue Marie Curie, BP 2060, 10010 Troyes Cedex (France); Smidts, Carol [Ohio State University (OSU), Nuclear Engineering Program, Department of Mechanical Engineering, Scott Laboratory, 201 W 19th Ave, Columbus OH 43210 (United States); Barros, Anne; Berenguer, Christophe [Universite de Technologie de Troyes (UTT), Institut Charles Delaunay (ICD) and UMR CNRS 6279 STMR, 12 rue Marie Curie, BP 2060, 10010 Troyes Cedex (France)

2011-07-15

Dynamic reliability explicitly handles the interactions between the stochastic behaviour of system components and the deterministic behaviour of process variables. While dynamic reliability provides a more efficient and realistic way to perform probabilistic risk assessment than 'static' approaches, its industrial level applications are still limited. Factors contributing to this situation are the inherent complexity of the theory and the lack of a generic platform. More recently the increased use of digital-based systems has also introduced additional modelling challenges related to specific interactions between system components. Typical examples are the 'intelligent transmitters' which are able to exchange information, and to perform internal data processing and advanced functionalities. To make a contribution to solving these challenges, the mathematical framework of dynamic reliability is extended to handle the data and information which are processed and exchanged between systems components. Stochastic deviations that may affect system properties are also introduced to enhance the modelling of failures. A formalized Petri net approach is then presented to perform the corresponding reliability analyses using numerical methods. Following this formalism, a versatile model for the dynamic reliability modelling of digital-based transmitters is proposed. Finally the framework's flexibility and effectiveness is demonstrated on a substantial case study involving a simplified model of a nuclear fast reactor.

Design verification enhancement of field programmable gate array-based safety-critical I&C system of nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Ahmed, Ibrahim [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104 (Korea, Republic of); Jung, Jaecheon, E-mail: jcjung@kings.ac.kr [Department of Nuclear Power Plant Engineering, KEPCO International Nuclear Graduate School, 658-91 Haemaji-ro, Seosang-myeon, Ulju-gun, Ulsan 45014 (Korea, Republic of); Heo, Gyunyoung [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104 (Korea, Republic of)

2017-06-15

Highlights: • An enhanced, systematic and integrated design verification approach is proposed for V&V of FPGA-based I&C system of NPP. • RPS bistable fixed setpoint trip algorithm is designed, analyzed, verified and discussed using the proposed approaches. • The application of integrated verification approach simultaneously verified the entire design modules. • The applicability of the proposed V&V facilitated the design verification processes. - Abstract: Safety-critical instrumentation and control (I&C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. However, safety analysis for FPGA-based I&C systems, and verification and validation (V&V) assessments still remain important issues to be resolved, which are now become a global research point of interests. In this work, we proposed a systematic design and verification strategies from start to ready-to-use in form of model-based approaches for FPGA-based reactor protection system (RPS) that can lead to the enhancement of the design verification and validation processes. The proposed methodology stages are requirement analysis, enhanced functional flow block diagram (EFFBD) models, finite state machine with data path (FSMD) models, hardware description language (HDL) code development, and design verifications. The design verification stage includes unit test – Very high speed integrated circuit Hardware Description Language (VHDL) test and modified condition decision coverage (MC/DC) test, module test – MATLAB/Simulink Co-simulation test, and integration test – FPGA hardware test beds. To prove the adequacy of the proposed

Design verification enhancement of field programmable gate array-based safety-critical I&C system of nuclear power plant

International Nuclear Information System (INIS)

Ahmed, Ibrahim; Jung, Jaecheon; Heo, Gyunyoung

2017-01-01

Highlights: • An enhanced, systematic and integrated design verification approach is proposed for V&V of FPGA-based I&C system of NPP. • RPS bistable fixed setpoint trip algorithm is designed, analyzed, verified and discussed using the proposed approaches. • The application of integrated verification approach simultaneously verified the entire design modules. • The applicability of the proposed V&V facilitated the design verification processes. - Abstract: Safety-critical instrumentation and control (I&C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. However, safety analysis for FPGA-based I&C systems, and verification and validation (V&V) assessments still remain important issues to be resolved, which are now become a global research point of interests. In this work, we proposed a systematic design and verification strategies from start to ready-to-use in form of model-based approaches for FPGA-based reactor protection system (RPS) that can lead to the enhancement of the design verification and validation processes. The proposed methodology stages are requirement analysis, enhanced functional flow block diagram (EFFBD) models, finite state machine with data path (FSMD) models, hardware description language (HDL) code development, and design verifications. The design verification stage includes unit test – Very high speed integrated circuit Hardware Description Language (VHDL) test and modified condition decision coverage (MC/DC) test, module test – MATLAB/Simulink Co-simulation test, and integration test – FPGA hardware test beds. To prove the adequacy of the proposed

Enhancing product robustness in reliability-based design optimization

International Nuclear Information System (INIS)

Zhuang, Xiaotian; Pan, Rong; Du, Xiaoping

2015-01-01

Different types of uncertainties need to be addressed in a product design optimization process. In this paper, the uncertainties in both product design variables and environmental noise variables are considered. The reliability-based design optimization (RBDO) is integrated with robust product design (RPD) to concurrently reduce the production cost and the long-term operation cost, including quality loss, in the process of product design. This problem leads to a multi-objective optimization with probabilistic constraints. In addition, the model uncertainties associated with a surrogate model that is derived from numerical computation methods, such as finite element analysis, is addressed. A hierarchical experimental design approach, augmented by a sequential sampling strategy, is proposed to construct the response surface of product performance function for finding optimal design solutions. The proposed method is demonstrated through an engineering example. - Highlights: • A unifying framework for integrating RBDO and RPD is proposed. • Implicit product performance function is considered. • The design problem is solved by sequential optimization and reliability assessment. • A sequential sampling technique is developed for improving design optimization. • The comparison with traditional RBDO is provided

Patient safety in anesthesia: learning from the culture of high-reliability organizations.

Science.gov (United States)

Wright, Suzanne M

2015-03-01

There has been an increased awareness of and interest in patient safety and improved outcomes, as well as a growing body of evidence substantiating medical error as a leading cause of death and injury in the United States. According to The Joint Commission, US hospitals demonstrate improvements in health care quality and patient safety. Although this progress is encouraging, much room for improvement remains. High-reliability organizations, industries that deliver reliable performances in the face of complex working environments, can serve as models of safety for our health care system until plausible explanations for patient harm are better understood. Copyright © 2015 Elsevier Inc. All rights reserved.

Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

International Nuclear Information System (INIS)

Lowe, Andrew; Hayward, Brent

2006-08-01

This report documents a study with the objective of enhancing safety culture in the Swedish nuclear power industry. A primary objective of this study was to ensure that the latest thinking on human factors principles was being recognised and applied by nuclear power operators as a means of ensuring optimal safety performance. The initial phase of the project was conducted as a pilot study, involving the senior management group at one Swedish nuclear power-producing site. The pilot study enabled the project methodology to be validated after which it was repeated at other Swedish nuclear power industry sites, providing a broad-ranging analysis of opportunities across the industry to enhance safety culture. The introduction to this report contains an overview of safety culture, explains the background to the project and sets out the project rationale and objectives. The methodology used for understanding and analysing the important safety culture issues at each nuclear power site is then described. This section begins with a summary of the processes used in the information gathering and data analysis stage. The six components of the Management Workshops conducted at each site are then described. These workshops used a series of presentations, interactive events and group exercises to: (a) provide feedback to site managers on the safety culture and safety leadership issues identified at their site, and (b) stimulate further safety thinking and provide 'take-away' information and leadership strategies that could be applied to promote safety culture improvements. Section 3, project Findings, contains the main observations and output from the project. These include: - a brief overview of aspects of the local industry operating context that impinge on safety culture; - a summary of strengths or positive attributes observed within the safety culture of the Swedish nuclear industry; - a set of identified opportunities for further improvement; - the aggregated results of the

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

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

Optimal design of water supply networks for enhancing seismic reliability

International Nuclear Information System (INIS)

Yoo, Do Guen; Kang, Doosun; Kim, Joong Hoon

2016-01-01

The goal of the present study is to construct a reliability evaluation model of a water supply system taking seismic hazards and present techniques to enhance hydraulic reliability of the design into consideration. To maximize seismic reliability with limited budgets, an optimal design model is developed using an optimization technique called harmony search (HS). The model is applied to actual water supply systems to determine pipe diameters that can maximize seismic reliability. The reliabilities between the optimal design and existing designs were compared and analyzed. The optimal design would both enhance reliability by approximately 8.9% and have a construction cost of approximately 1.3% less than current pipe construction cost. In addition, the reinforcement of the durability of individual pipes without considering the system produced ineffective results in terms of both cost and reliability. Therefore, to increase the supply ability of the entire system, optimized pipe diameter combinations should be derived. Systems in which normal status hydraulic stability and abnormal status available demand could be maximally secured if configured through the optimal design. - Highlights: • We construct a seismic reliability evaluation model of water supply system. • We present technique to enhance hydraulic reliability in the aspect of design. • Harmony search algorithm is applied in optimal designs process. • The effects of the proposed optimal design are improved reliability about by 9%. • Optimized pipe diameter combinations should be derived indispensably.

Human reliability in probabilistic safety assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

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

Reliability Based Optimization of Fire Protection

DEFF Research Database (Denmark)

Thoft-Christensen, Palle

fire protection (PFP) of firewalls and structural members. The paper is partly based on research performed within the EU supported research project B/E-4359 "Optimized Fire Safety of Offshore Structures" and partly on research supported by the Danish Technical Research Council (see Thoft-Christensen [1......]). Special emphasis is put on the optimization software developed within the project.......It is well known that fire is one of the major risks of serious damage or total loss of several types of structures such as nuclear installations, buildings, offshore platforms/topsides etc. This paper presents a methodology and software for reliability based optimization of the layout of passive...

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Computer-based and web-based radiation safety training

Energy Technology Data Exchange (ETDEWEB)

Owen, C., LLNL

1998-03-01

The traditional approach to delivering radiation safety training has been to provide a stand-up lecture of the topic, with the possible aid of video, and to repeat the same material periodically. New approaches to meeting training requirements are needed to address the advent of flexible work hours and telecommuting, and to better accommodate individuals learning at their own pace. Computer- based and web-based radiation safety training can provide this alternative. Computer-based and web- based training is an interactive form of learning that the student controls, resulting in enhanced and focused learning at a time most often chosen by the student.

Reliability Based Calibration of Fatigue Design Guidelines for Ship Structures

DEFF Research Database (Denmark)

Folsø, Rasmus; Otto, S.; Parmentier, G.

2002-01-01

A simple reliability based framework is applied to calibrate a new set of fatigue design guidelines. This new guideline considers two different approaches for the assessment of both loads, stresses and local stress raising effects, and partial safety factors must be given for any combination...

Use of reliability analysis for the safety evaluation of technical facilities

International Nuclear Information System (INIS)

Balfanz, H.P.; Eggert, H.; Lindauer, E.

1975-01-01

Using examples from nuclear technology, the following is discussed: how efficient the present practical measures are for increasing reliability, which weak points can be recognized and what appears to be the most promising direction to take for improvements. The following are individually dealt with: 1) determination of the relevant parameters for the safety of a plant; 2) definition and fixing of reliability requirements; 3) process to prove the fulfilment of requirements; 4) measures to guarantee the reliability; 5) data feed-back to check and improve the reliability. (HP/LH) [de

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.)

Engineering Solutions to Enhance Traffic Safety Performance on Two-Lane Highways

Directory of Open Access Journals (Sweden)

Lina Wu

2015-01-01

Full Text Available Improving two-lane highway traffic safety conditions is of practical importance to the traffic system, which has attracted significant research attention within the last decade. Many cost-effective and proactive solutions such as low-cost treatments and roadway safety monitoring programs have been developed to enhance traffic safety performance under prevailing conditions. This study presents research perspectives achieved from the Highway Safety Enhancement Project (HSEP that assessed safety performance on two-lane highways in Beijing, China. Potential causal factors are identified based on proposed evaluation criteria, and primary countermeasures are developed against inferior driving conditions such as sharp curves, heavy gradients, continuous downgrades, poor sight distance, and poor clear zones. Six cost-effective engineering solutions were specifically implemented to improve two-lane highway safety conditions, including (1 traffic sign replacement, (2 repainting pavement markings, (3 roadside barrier installation, (4 intersection channelization, (5 drainage optimization, and (6 sight distance improvement. The effectiveness of these solutions was examined and evaluated based on Empirical Bayes (EB models. The results indicate that the proposed engineering solutions effectively improved traffic safety performance by significantly reducing crash occurrence risks and crash severities.

Safety management of pipelines based on structural reliability: historical perspective and progress

Energy Technology Data Exchange (ETDEWEB)

Beck, Andre Teofilo; Bazan, Felipe Alexander Vargas [Universidade de Sao Paulo (USP), SP (Brazil); Mendes, Renato [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil); Donato, Guilherme [Petroleo Brasileiro S.A. (PETROBRAS/CENPES), Rio de Janeiro, RJ (Brazil). Centro de Pesquisas e Desenvolvimento; Gardiner, Michael [GL Noble Denton, Loughborough (United Kingdom)

2012-07-01

Since the early beginnings in the sixties and seventies, Structural Reliability theory has reached a mature stage encompassing solid theoretical developments and increasing practical applications. Structural reliability methods have permeated the engineering profession, finding applications in code calibration, structural optimization, life extension of existing structures, life-cycle management of infrastructure risks and costs, and so on. This review paper shows that the ground work for Reliability Based Design and Assessment (RBDA) of onshore pipeline systems is already developed. Hence, this allows the economic management of the risks involved in operation of pipeline systems. It is shown that RBDA is a rational tool to safely manage the operational life of pipeline systems, optimizing initial design and the expenditures in inspection and maintenance operations. (author)

Reliability calculations

International Nuclear Information System (INIS)

Petersen, K.E.

1986-03-01

Risk and reliability analysis is increasingly being used in evaluations of plant safety and plant reliability. The analysis can be performed either during the design process or during the operation time, with the purpose to improve the safety or the reliability. Due to plant complexity and safety and availability requirements, sophisticated tools, which are flexible and efficient, are needed. Such tools have been developed in the last 20 years and they have to be continuously refined to meet the growing requirements. Two different areas of application were analysed. In structural reliability probabilistic approaches have been introduced in some cases for the calculation of the reliability of structures or components. A new computer program has been developed based upon numerical integration in several variables. In systems reliability Monte Carlo simulation programs are used especially in analysis of very complex systems. In order to increase the applicability of the programs variance reduction techniques can be applied to speed up the calculation process. Variance reduction techniques have been studied and procedures for implementation of importance sampling are suggested. (author)

Reliability analysis and computation of computer-based safety instrumentation and control used in German nuclear power plant. Final report; Zuverlaessigkeitsuntersuchung und -berechnung rechnerbasierter Sicherheitsleittechnik zum Einsatz in deutschen Kernkraftwerken. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Ding, Yongjian [Hochschule Magdeburg-Stendal, Magdeburg (Germany). Inst. fuer Elektrotechnik; Krause, Ulrich [Magdeburg Univ. (Germany). Inst. fuer Apparate- und Umwelttechnik; Gu, Chunlei

2014-08-21

The trend of technological advancement in the field of safety instrumentation and control (I and C) leads to increasingly frequent use of computer-based (digital) control systems which consisting of distributed, connected bus communications computers and their functionalities are freely programmable by qualified software. The advantages of the new I and C system over the old I and C system with hard-wired technology are e.g. in the higher flexibility, cost-effective procurement of spare parts, higher hardware reliability (through higher integration density, intelligent self-monitoring mechanisms, etc.). On the other hand, skeptics see the new technology with the computer-based I and C a higher potential by influences of common cause failures (CCF), and the easier manipulation by sabotage (IT Security). In this joint research project funded by the Federal Ministry for Economical Affaires and Energy (BMWi) (2011-2014, FJZ 1501405) the Otto-von-Guericke-University Magdeburg and Magdeburg-Stendal University of Applied Sciences are therefore trying to develop suitable methods for the demonstration of the reliability of the new instrumentation and control systems with the focus on the investigation of CCF. This expertise of both houses shall be extended to this area and a scientific contribution to the sound reliability judgments of the digital safety I and C in domestic and foreign nuclear power plants. First, the state of science and technology will be worked out through the study of national and international standards in the field of functional safety of electrical and I and C systems and accompanying literature. On the basis of the existing nuclear Standards the deterministic requirements on the structure of the new digital I and C system will be determined. The possible methods of reliability modeling will be analyzed and compared. A suitable method called multi class binomial failure rate (MCFBR) which was successfully used in safety valve applications will be

Reliability analysis of digital based I and C system

Energy Technology Data Exchange (ETDEWEB)

Kang, I. S.; Cho, B. S.; Choi, M. J. [KOPEC, Yongin (Korea, Republic of)

1999-10-01

Rapidly, digital technology is being widely applied in replacing analog component installed in existing plant and designing new nuclear power plant for control and monitoring system in Korea as well as in foreign countries. Even though many merits of digital technology, it is being faced with a new problem of reliability assurance. The studies for solving this problem are being performed vigorously in foreign countries. The reliability of KNGR Engineered Safety Features Component Control System (ESF-CCS), digital based I and C system, was analyzed to verify fulfillment of the ALWR EPRI-URD requirement for reliability analysis and eliminate hazards in design applied new technology. The qualitative analysis using FMEA and quantitative analysis using reliability block diagram were performed. The results of analyses are shown in this paper.

The ConCom Safety Management Scale: developing and testing a measurement instrument for control-based and commitment-based safety management approaches in hospitals.

Science.gov (United States)

Alingh, Carien W; Strating, Mathilde M H; van Wijngaarden, Jeroen D H; Paauwe, Jaap; Huijsman, Robbert

2018-03-06

Nursing management is considered important for patient safety. Prior research has predominantly focused on charismatic leadership styles, although it is questionable whether these best characterise the role of nurse managers. Managerial control is also relevant. Therefore, we aimed to develop and test a measurement instrument for control-based and commitment-based safety management of nurse managers in clinical hospital departments. A cross-sectional survey design was used to test the newly developed questionnaire in a sample of 2378 nurses working in clinical departments. The nurses were asked about their perceptions of the leadership behaviour and management practices of their direct supervisors. Psychometric properties were evaluated using confirmatory factor analysis and reliability estimates. The final 33-item questionnaire showed acceptable goodness-of-fit indices and internal consistency (Cronbach's α of the subscales range: 0.59-0.90). The factor structure revealed three subdimensions for control-based safety management: (1) stressing the importance of safety rules and regulations; (2) monitoring compliance; and (3) providing employees with feedback. Commitment-based management consisted of four subdimensions: (1) showing role modelling behaviour; (2) creating safety awareness; (3) showing safety commitment; and (4) encouraging participation. Construct validity of the scale was supported by high factor loadings and provided preliminary evidence that control-based and commitment-based safety management are two distinct yet related constructs. The findings were reconfirmed in a cross-validation procedure. The results provide initial support for the construct validity and reliability of our ConCom Safety Management Scale. Both management approaches were found to be relevant for managing patient safety in clinical hospital departments. The scale can be used to deepen our understanding of the influence of patient safety management on healthcare professionals

A simple reliability block diagram method for safety integrity verification

International Nuclear Information System (INIS)

Guo Haitao; Yang Xianhui

2007-01-01

IEC 61508 requires safety integrity verification for safety related systems to be a necessary procedure in safety life cycle. PFD avg must be calculated to verify the safety integrity level (SIL). Since IEC 61508-6 does not give detailed explanations of the definitions and PFD avg calculations for its examples, it is difficult for common reliability or safety engineers to understand when they use the standard as guidance in practice. A method using reliability block diagram is investigated in this study in order to provide a clear and feasible way of PFD avg calculation and help those who take IEC 61508-6 as their guidance. The method finds mean down times (MDTs) of both channel and voted group first and then PFD avg . The calculated results of various voted groups are compared with those in IEC61508 part 6 and Ref. [Zhang T, Long W, Sato Y. Availability of systems with self-diagnostic components-applying Markov model to IEC 61508-6. Reliab Eng System Saf 2003;80(2):133-41]. An interesting outcome can be realized from the comparison. Furthermore, although differences in MDT of voted groups exist between IEC 61508-6 and this paper, PFD avg of voted groups are comparatively close. With detailed description, the method of RBD presented can be applied to the quantitative SIL verification, showing a similarity of the method in IEC 61508-6

Human Reliability in Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

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

Enhancement of reliability of PLT-safety devices by utilization of process control components; Steigerung der Verfuegbarkeit von PLT-Schutzeinrichtungen durch Mitbenutzung von Komponenten des Prozessleitsystems

Energy Technology Data Exchange (ETDEWEB)

Gabriel, T.; Litz, L. [Technische Univ. Kaiserslautern (Germany); Schroers, B. [Material-Science AG, Leverkusen (Germany)

2008-01-15

According to the standard IEC 61511 each safety-related loop is assigned to one of the four Safety Integrity Levels (SILs). For every safety-related loop a SIL-specific Probability of Failure on Demand (PFD) must be proven. Usually, the PFD calculation is performed based upon the failure rates of each loop component aided by commercial software tools. However, this bottom-up approach suffers from many uncertainties. Especially, a lack of reliable failure rate data causes many problems. Reference data collected in different environments are available to solve this situation. However, this pragmatism leads to a PFD bandwidth, not to a single PFD value as desired. In order to make a decision for a numerical value appropriate for the chemical and pharmaceutical process industry a data ascertainment has been initiated by the European NAMUR. Its results display large deficiencies for the bottom-up approach. The error sources leading to this situation are located and analyzed. (GL)

Analysis Testing of Sociocultural Factors Influence on Human Reliability within Sociotechnical Systems: The Algerian Oil Companies

Directory of Open Access Journals (Sweden)

Abdelbaki Laidoune

2016-09-01

Conclusion: The explored sociocultural factors influence the human reliability both in qualitative and quantitative manners. The proposed model shows how reliability can be enhanced by some measures such as experience feedback based on, for example, safety improvements, training, and information. With that is added the continuous systems improvements to improve sociocultural reality and to reduce negative behaviors.

Reliability-based approaches for safety margin assessment in the French nuclear industry

International Nuclear Information System (INIS)

Ardillon, E.; Barthelet, B.; Meister, E.; Cambefort, P.; Hornet, P.; Le Delliou, P.

2003-01-01

The prevention of the fast fracture damage of the mechanical equipment important for the safety of nuclear islands of the French PWR relies on deterministic rules. These rules include flaw acceptance criteria involving safety factors applied to characteristic values (implicit margins) of the physical variables. The sets of safety factors that are currently under application in the industrial analyses with the agreement of the Safety Authority, are distributed across the two main physical parameters and have partly been based on a semi-probabilistic approach. After presenting the generic probabilistic pro-codification approach this paper shows its application to the evaluation of the performances of the existing regulatory flaw acceptance criteria. This application can be carried out in a realistic manner or in a more simplified one. These two approaches are applied to representative mechanical components. Their results are consistent. (author)

Connectivity-Based Reliable Multicast MAC Protocol for IEEE 802.11 Wireless LANs

Directory of Open Access Journals (Sweden)

Woo-Yong Choi

2009-01-01

Full Text Available We propose the efficient reliable multicast MAC protocol based on the connectivity information among the recipients. Enhancing the BMMM (Batch Mode Multicast MAC protocol, the reliable multicast MAC protocol significantly reduces the RAK (Request for ACK frame transmissions in a reasonable computational time and enhances the MAC performance. By the analytical performance analysis, the throughputs of the BMMM protocol and our proposed MAC protocol are derived. Numerical examples show that our proposed MAC protocol increases the reliable multicast MAC performance for IEEE 802.11 wireless LANs.

Development of a Nursing Handoff Tool: A Web-Based Application to Enhance Patient Safety

Science.gov (United States)

Goldsmith, Denise; Boomhower, Marc; Lancaster, Diane R.; Antonelli, Mary; Kenyon, Mary Anne Murphy; Benoit, Angela; Chang, Frank; Dykes, Patricia C.

2010-01-01

Dynamic and complex clinical environments present many challenges for effective communication among health care providers. The omission of accurate, timely, easily accessible vital information by health care providers significantly increases risk of patient harm and can have devastating consequences for patient care. An effective nursing handoff supports the standardized transfer of accurate, timely, critical patient information, as well as continuity of care and treatment, resulting in enhanced patient safety. The Brigham and Women’s/Faulkner Hospital Healthcare Information Technology Innovation Program (HIP) is supporting the development of a web based nursing handoff tool (NHT). The goal of this project is to develop a “proof of concept” handoff application to be evaluated by nurses on the inpatient intermediate care units. The handoff tool would enable nurses to use existing knowledge of evidence-based handoff methodology in their everyday practice to improve patient care and safety. In this paper, we discuss the results of nursing focus groups designed to identify the current state of handoff practice as well as the functional and data element requirements of a web based Nursing Handoff Tool (NHT). PMID:21346980

Analysis of the reliability of the active injection safety systems of Angra I

International Nuclear Information System (INIS)

Frutuoso e Melo, P.F.F.

1981-01-01

The reliability of the active emergency core cooling systems of Angra I nuclear power plant is evaluated. The fault tree analysis is employed. The unavailability of the above cited systems, is calculated. A parametric sensitivity analysis has been performed, due to the existing scattering in the failure and repair rate data of these system's components. The minimal cut sets were determined and, as a final step, a reliability importance analysis has been performed. This final step has required the development of a computer program. The methodology and data from the 'Reactor Safety Study' (Wash-1400) (in which the reliability of safety systems of a tipical PWR plant is calculated), is employed. The unavailability values for the safety systems analysed are too low, thus showing that in most cases the systems analysed are available to mitigate the effects of a loss-of-coolant accident. (Author) [pt

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)

Effects of image enhancement on reliability of landmark identification in digital cephalometry

Directory of Open Access Journals (Sweden)

M Oshagh

2013-01-01

Full Text Available Introduction: Although digital cephalometric radiography is gaining popularity in orthodontic practice, the most important source of error in its tracing is uncertainty in landmark identification. Therefore, efforts to improve accuracy in landmark identification were directed primarily toward the improvement in image quality. One of the more useful techniques of this process involves digital image enhancement which can increase overall visual quality of image, but this does not necessarily mean a better identification of landmarks. The purpose of this study was to evaluate the effectiveness of digital image enhancements on reliability of landmark identification. Materials and Methods: Fifteen common landmarks including 10 skeletal and 5 soft tissues were selected on the cephalograms of 20 randomly selected patients, prepared in Natural Head Position (NHP. Two observers (orthodontists identified landmarks on the 20 original photostimulable phosphor (PSP digital cephalogram images and 20 enhanced digital images twice with an intervening time interval of at least 4 weeks. The x and y coordinates were further analyzed to evaluate the pattern of recording differences in horizontal and vertical directions. Reliability of landmarks identification was analyzed by paired t test. Results: There was a significant difference between original and enhanced digital images in terms of reliability of points Ar and N in vertical and horizontal dimensions, and enhanced images were significantly more reliable than original images. Identification of A point, Pogonion and Pronasal points, in vertical dimension of enhanced images was significantly more reliable than original ones. Reliability of Menton point identification in horizontal dimension was significantly more in enhanced images than original ones. Conclusion: Direct digital image enhancement by altering brightness and contrast can increase reliability of some landmark identification and this may lead to more

Quantification of human reliability in probabilistic safety assessment

International Nuclear Information System (INIS)

Hirschberg, S.; Dankg, Vinh N.

1996-01-01

Human performance may substantially influence the reliability and safety of complex technical systems. For this reason, Human Reliability Analysis (HRA) constitutes an important part of Probabilistic Safety Assessment (PSAs) or Quantitative Risk Analyses (QRAs). The results of these studies as well as analyses of past accidents and incidents clearly demonstrate the importance of human interactions. The contribution of human errors to the core damage frequency (CDF), as estimated in the Swedish nuclear PSAs, are between 15 and 88%. A survey of the FRAs in the Swiss PSAs shows that also for the Swiss nuclear power plants the estimated HE contributions are substantial (49% of the CDF due to internal events in the case of Beznau and 70% in the case of Muehleberg; for the total CDF, including external events, 25% respectively 20%). Similar results can be extracted from the PSAs carried out for French, German, and US plants. In PSAs or QRAs, the adequate treatment of the human interactions with the system is a key to the understanding of accident sequences and their relative importance to overall risk. The main objectives of HRA are: first, to ensure that the key human interactions are systematically identified and incorporated into the safety analysis in a traceable manner, and second, to quantify the probabilities of their success and failure. Adopting a structured and systematic approach to the assessment of human performance makes it possible to provide greater confidence that the safety and availability of human-machine systems is not unduly jeopardized by human performance problems. Section 2 discusses the different types of human interactions analysed in PSAs. More generally, the section presents how HRA fits in the overall safety analysis, that is, how the human interactions to be quantified are identified. Section 3 addresses the methods for quantification. Section 4 concludes the paper by presenting some recommendations and pointing out the limitations of the

Risk-based rules for crane safety systems

Energy Technology Data Exchange (ETDEWEB)

Ruud, Stian [Section for Control Systems, DNV Maritime, 1322 Hovik (Norway)], E-mail: Stian.Ruud@dnv.com; Mikkelsen, Age [Section for Lifting Appliances, DNV Maritime, 1322 Hovik (Norway)], E-mail: Age.Mikkelsen@dnv.com

2008-09-15

The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented.

Risk-based rules for crane safety systems

International Nuclear Information System (INIS)

Ruud, Stian; Mikkelsen, Age

2008-01-01

The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented

Procedures for controlling the risks of reliability, safety, and availability of technical systems

International Nuclear Information System (INIS)

1987-01-01

The reference book covers four sections. Apart from the fundamental aspects of the reliability problem, of risk and safety and the relevant criteria with regard to reliability, the material presented explains reliability in terms of maintenance, logistics and availability, and presents procedures for reliability assessment and determination of factors influencing the reliability, together with suggestions for systems technical integration. The reliability assessment consists of diagnostic and prognostic analyses. The section on factors influencing reliability discusses aspects of organisational structures, programme planning and control, and critical activities. (DG) [de

Suitability review of FMEA and reliability analysis for digital plant protection system and digital engineered safety features actuation system

Energy Technology Data Exchange (ETDEWEB)

Kim, I. S.; Kim, T. K.; Kim, M. C.; Kim, B. S.; Hwang, S. W.; Ryu, K. C. [Hanyang Univ., Seoul (Korea, Republic of)

2000-11-15

Of the many items that should be checked out during a review stage of the licensing application for the I and C system of Ulchin 5 and 6 units, this report relates to a suitability review of the reliability analysis of Digital Plant Protection System (DPPS) and Digital Engineered Safety Features Actuation System (DESFAS). In the reliability analysis performed by the system designer, ABB-CE, fault tree analysis was used as the main methods along with Failure Modes and Effect Analysis (FMEA). However, the present regulatory technique dose not allow the system reliability analysis and its results to be appropriately evaluated. Hence, this study was carried out focusing on the following four items ; development of general review items by which to check the validity of a reliability analysis, and the subsequent review of suitability of the reliability analysis for Ulchin 5 and 6 DPPS and DESFAS L development of detailed review items by which to check the validity of an FMEA, and the subsequent review of suitability of the FMEA for Ulchin 5 and 6 DPPS and DESFAS ; development of detailed review items by which to check the validity of a fault tree analysis, and the subsequent review of suitability of the fault tree for Ulchin 5 and 6 DPPS and DESFAS ; an integrated review of the safety and reliability of the Ulchin 5 and 6 DPPS and DESFAS based on the results of the various reviews above and also of a reliability comparison between the digital systems and the comparable analog systems, i.e., and analog Plant Protection System (PPS) and and analog Engineered Safety Features Actuation System (ESFAS). According to the review mentioned above, the reliability analysis of Ulchin 5 and 6 DPPS and DESFAS generally satisfies the review requirements. However, some shortcomings of the analysis were identified in our review such that the assumed test periods for several equipment were not properly incorporated in the analysis, and failures of some equipment were not included in the

Nuclear electric propulsion operational reliability and crew safety study

International Nuclear Information System (INIS)

Karns, J.J.; Fragola, J.R.; Kahan, L.; Pelaccio, D.

1993-01-01

The central purpose of this analysis is to assess the ''achievability'' of a nuclear electric propulsion (NEP) system in a given mission. ''Achievability'' is a concept introduced to indicate the extent to which a system that meets or achieves its design goals might be implemented using the existing technology base. In the context of this analysis, the objective is to assess the achievability of an NEP system for a manned Mars mission as it pertains to operational reliability and crew safety goals. By varying design parameters, then examining the resulting system achievability, the design and mission risk drivers can be identified. Additionally, conceptual changes in design approach or mission strategy which are likely to improve overall achievability of the NEP system can be examined

Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

Energy Technology Data Exchange (ETDEWEB)

Kalchev, B [Energoproekt, Sofia (Bulgaria)

1996-12-31

The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E{sup -3}, 1/RY to 1.07E{sup -3}, 1/RY. 5 refs., 2 tabs., 5 figs.

Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

International Nuclear Information System (INIS)

Kalchev, B.

1995-01-01

The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E -3 , 1/RY to 1.07E -3 , 1/RY. 5 refs., 2 tabs., 5 figs

IAEA TC Project 'Strengthening safety and reliability of fuel and materials in nuclear power plants'

International Nuclear Information System (INIS)

Makihara, Y.

2008-01-01

The Regional TC Project in Europe RER9076 'Strengthening Safety and Reliability of Fuel and Materials in Nuclear Power Plants' was launched in 2003 as a four-year project and was subsequently extended in 2006 to run through 2008. The purpose of the Project is to support the Central and Eastern European countries with the necessary tools to fulfill their own fuel and material licensing needs. The main objective will be to provide quality data on fuel and materials irradiated in power reactors and in dedicated experiments carried out in material test reactors (MTRs). Within the framework of the Project, ten tasks were implemented. These included experiments performed at the test facilities in the region, training courses and workshops related to fuel safety. While several tasks are expected to be completed by the end of RER9076, some remain. It would be desirable to initiate a new RER Project from the next TC cycle (2009-2011) in order to take over RER9076 and to implement new tasks required for enhancing fuel safety in the region. (author)

A study on the quantitative evaluation of the reliability for safety critical software using Bayesian belief nets

International Nuclear Information System (INIS)

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

2003-01-01

Despite the efforts to avoid undesirable risks, or at least to bring them under control in the world, new risks that are highly difficult to manage continue to emerge from the use of new technologies, such as the use of digital instrumentation and control (I and C) components in nuclear power plant. Whenever new risk issues came out by now, we have endeavored to find the most effective ways to reduce risks, or to allocate limited resources to do this. One of the major challenges is the reliability analysis of safety-critical software associated with digital safety systems. Though many activities such as testing, verification and validation (V and V) techniques have been carried out in the design stage of software, however, the process of quantitatively evaluating the reliability of safety-critical software has not yet been developed because of the irrelevance of the conventional software reliability techniques to apply for the digital safety systems. This paper focuses on the applicability of Bayesian Belief Net (BBN) techniques to quantitatively estimate the reliability of safety-critical software adopted in digital safety system. In this paper, a typical BBN model was constructed using the dedication process of the Commercial-Off-The-Shelf (COTS) installed by KAERI. In conclusion, the adoption of BBN technique can facilitate the process of evaluating the safety-critical software reliability in nuclear power plant, as well as provide very useful information (e.g., 'what if' analysis) associated with software reliability in the viewpoint of practicality

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.

International cooperation - a way to improve reliability and safety

International Nuclear Information System (INIS)

John, A.

1998-01-01

The mission of the World Association of Nuclear Operators (WANO) is highlighted, and WANO's Peer Review programme is described. At the Dukovany nuclear power plant, a Peer Review was undertaken in December 1997. The results gave evidence of a good level of safety, reliability and culture of operation of the plant. (P.A.)

System principles, mathematical models and methods to ensure high reliability of safety systems

Science.gov (United States)

Zaslavskyi, V.

2017-04-01

Modern safety and security systems are composed of a large number of various components designed for detection, localization, tracking, collecting, and processing of information from the systems of monitoring, telemetry, control, etc. They are required to be highly reliable in a view to correctly perform data aggregation, processing and analysis for subsequent decision making support. On design and construction phases of the manufacturing of such systems a various types of components (elements, devices, and subsystems) are considered and used to ensure high reliability of signals detection, noise isolation, and erroneous commands reduction. When generating design solutions for highly reliable systems a number of restrictions and conditions such as types of components and various constrains on resources should be considered. Various types of components perform identical functions; however, they are implemented using diverse principles, approaches and have distinct technical and economic indicators such as cost or power consumption. The systematic use of different component types increases the probability of tasks performing and eliminates the common cause failure. We consider type-variety principle as an engineering principle of system analysis, mathematical models based on this principle, and algorithms for solving optimization problems of highly reliable safety and security systems design. Mathematical models are formalized in a class of two-level discrete optimization problems of large dimension. The proposed approach, mathematical models, algorithms can be used for problem solving of optimal redundancy on the basis of a variety of methods and control devices for fault and defects detection in technical systems, telecommunication networks, and energy systems.

Technical feasibility and reliability of passive safety systems of AC600

International Nuclear Information System (INIS)

Niu, W.; Zeng, X.

1996-01-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished by the Nuclear Power Institute of China. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also

Technical feasibility and reliability of passive safety systems of AC600

Energy Technology Data Exchange (ETDEWEB)

Niu, W; Zeng, X [Nuclear Power Inst. of China, Chendu (China)

1996-12-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also involved. (author). 3 figs, 1 tab.

Operational experience review and methods to enhance safety and reliability in the NPP-Leibstadt (KKL)

Energy Technology Data Exchange (ETDEWEB)

Haeusermann, R [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland)

1997-10-01

In the nuclear community it became clear that an integrated feedback system of operating experience must also include the unsuccessful results. The deviations, expected to achieved performance are analysed to the failure mode and its effect. KKL has lowered the number of safety significant events since commercial operation started. The thoroughness of the review/analysis of the events has increased with high priority set to human factor induced events in operation and maintenance. Since the participation of the author in the ASSET-Mission in Smolensk in 1993, KKL introduced the ASSET-Root-Cause method and has supplemented it by the HPES (Human Performance Enhancement System). 4 refs, 6 figs.

Operational experience review and methods to enhance safety and reliability in the NPP-Leibstadt (KKL)

International Nuclear Information System (INIS)

Haeusermann, R.

1997-01-01

In the nuclear community it became clear that an integrated feedback system of operating experience must also include the unsuccessful results. The deviations, expected to achieved performance are analysed to the failure mode and its effect. KKL has lowered the number of safety significant events since commercial operation started. The thoroughness of the review/analysis of the events has increased with high priority set to human factor induced events in operation and maintenance. Since the participation of the author in the ASSET-Mission in Smolensk in 1993, KKL introduced the ASSET-Root-Cause method and has supplemented it by the HPES (Human Performance Enhancement System). 4 refs, 6 figs

Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

International Nuclear Information System (INIS)

1996-01-01

An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs

Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs.

Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software.

Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

International Nuclear Information System (INIS)

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

2001-03-01

As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software

A Nuclear Safety System based on Industrial Computer

International Nuclear Information System (INIS)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack

2011-01-01

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

A Nuclear Safety System based on Industrial Computer

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack [Korea Electric Power Corporation Engineering and Construction, Daejeon (Korea, Republic of)

2011-05-15

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

CONSIDERING TRAVEL TIME RELIABILITY AND SAFETY FOR EVALUATION OF CONGESTION RELIEF SCHEMES ON EXPRESSWAY SEGMENTS

Directory of Open Access Journals (Sweden)

Babak MEHRAN

2009-01-01

Full Text Available Evaluation of the efficiency of congestion relief schemes on expressways has generally been based on average travel time analysis. However, road authorities are much more interested in knowing the possible impacts of improvement schemes on safety and travel time reliability prior to implementing them in real conditions. A methodology is presented to estimate travel time reliability based on modeling travel time variations as a function of demand, capacity and weather conditions. For a subject expressway segment, patterns of demand and capacity were generated for each 5-minute interval over a year by using the Monte-Carlo simulation technique, and accidents were generated randomly according to traffic conditions. A whole year analysis was performed by comparing demand and available capacity for each scenario and shockwave analysis was used to estimate the queue length at each time interval. Travel times were estimated from refined speed-flow relationships and buffer time index was estimated as a measure of travel time reliability. it was shown that the estimated reliability measures and predicted number of accidents are very close to observed values through empirical data. After validation, the methodology was applied to assess the impact of two alternative congestion relief schemes on a subject expressway segment. one alternative was to open the hard shoulder to traffic during the peak period, while the other was to reduce the peak period demand by 15%. The extent of improvements in travel conditions and safety, likewise the reduction in road users' costs after implementing each improvement scheme were estimated. it was shown that both strategies can result in up to 23% reduction in the number of occurred accidents and significant improvements in travel time reliability. Finally, the advantages and challenging issues of selecting each improvement scheme were discussed.

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

International Nuclear Information System (INIS)

Lee, Seung Woo; Lee, Hwa Ki

2008-01-01

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

The enhancement of Ignalina NPP in design and operational safety

International Nuclear Information System (INIS)

Negrivoda, G.

1999-01-01

Enhancement of Ignalina NPP design include: core design improvements; fuel channel integrity (multiple pressure tube rupture); improvements of shutdown systems; improvements of instrumentation and control devices; containment strength and tightness; design basis accident analysis; improvements of safety and support systems; seismic safety enhancement; Year 2000 project; cracks in pipes. Enhancement of operational safety includes: quality assurance; configuration management; safety management and safety culture; emergency operating procedures; training and full scope simulator; in-service inspection; fire protection and ageing monitoring and management

Reliability analysis of microcomputer boards and computer based systems important to safety of nuclear plants

International Nuclear Information System (INIS)

Shrikhande, S.V.; Patil, V.K.; Ganesh, G.; Biswas, B.; Patil, R.K.

2010-01-01

Computer Based Systems (CBS) are employed in Indian nuclear plants for protection, control and monitoring purpose. For forthcoming CBS, Reactor Control Division has designed and developed a new standardized family of microcomputer boards qualified to stringent requirements of nuclear industry. These boards form the basic building blocks of CBS. Reliability analysis of these boards is being carried out using analysis package based on MIL-STD-217Plus methodology. The estimated failure rate values of these standardized microcomputer boards will be useful for reliability assessment of these systems. The paper presents reliability analysis of microcomputer boards and case study of a CBS system built using these boards. (author)

Design measures to increase safety and reliability of power station control and protection systems

International Nuclear Information System (INIS)

Edelmann, J.; Spieth, W.

1977-06-01

The paper reviews a few criteria which exert a considerable influence on the safety and reliability of monitoring and control systems. When judging the safety and reliability of a system, it is of importance not only to look at the failures of just one part of a system but also to take into account the effect these failures have on the overall process. In this respect there is a marked difference between a centralized and a decentralized system. With the technical equipment nowadays at our disposal a high safety standard has been reached. Redundant and dynamic protection systems make the occurrence of a dangerous failure hypothetic. (Author)

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.

Design provisions for safety

International Nuclear Information System (INIS)

Birkhofer, A.

1983-01-01

Design provisions for safety of nuclear power plants are based on a well balanced concept: the public is protected against a release of radioactive material by multiple barriers. These barriers are protected according to a 'defence-in-depth' principle. The reactor safety concept is primarily aimed at the prevention of accidents, especially fuel damage. Additionally, measures for consequence limitation are provided in order to prevent a severe release of radioactivity to the environment. However, it is difficult to judge the overall effectiveness of such devices. In a comprehensive safety analysis it has to be shown that the protection systems and safeguards work with sufficient reliability in the event of an accident. For the reliability assessment deterministic criteria (single failure, redundancy, fail-safe, demand for diversity) play an important role. Increasing efforts have been made to assess reliability quantitatively by means of probabilistic methods. It is now usual to perform reliability analyses of essential systems of nuclear power plants in the course of licensing procedures. As an additional level of emergency measures for a further reduction of hazards a reasonable amount of accident information has to be transferred. Operational experience may be considered as an important feedback to the design of plant safety features. Operator training has to include, besides skill in performing of operating procedures, the training of a flexible response to different accident situations. Experience has shown that the design provisions for safety could prevent dangerous release of the radioactive material to the environment after an accident has occurred. For future developments of reactor safety, extensive analyses of operating experience are of great importance. The main goal should be to enhance the reliability of measures for accident prevention, which prevent the core from meltdown or other damages

Application of reliability analysis methods to the comparison of two safety circuits

International Nuclear Information System (INIS)

Signoret, J.-P.

1975-01-01

Two circuits of different design, intended for assuming the ''Low Pressure Safety Injection'' function in PWR reactors are analyzed using reliability methods. The reliability analysis of these circuits allows the failure trees to be established and the failure probability derived. The dependence of these results on test use and maintenance is emphasized as well as critical paths. The great number of results obtained may allow a well-informed choice taking account of the reliability wanted for the type of circuits [fr

A Hybrid Approach for Reliability Analysis Based on Analytic Hierarchy Process and Bayesian Network

International Nuclear Information System (INIS)

Zubair, Muhammad

2014-01-01

By using analytic hierarchy process (AHP) and Bayesian Network (BN) the present research signifies the technical and non-technical issues of nuclear accidents. The study exposed that the technical faults was one major reason of these accidents. Keep an eye on other point of view it becomes clearer that human behavior like dishonesty, insufficient training, and selfishness are also play a key role to cause these accidents. In this study, a hybrid approach for reliability analysis based on AHP and BN to increase nuclear power plant (NPP) safety has been developed. By using AHP, best alternative to improve safety, design, operation, and to allocate budget for all technical and non-technical factors related with nuclear safety has been investigated. We use a special structure of BN based on the method AHP. The graphs of the BN and the probabilities associated with nodes are designed to translate the knowledge of experts on the selection of best alternative. The results show that the improvement in regulatory authorities will decrease failure probabilities and increase safety and reliability in industrial area.

Development of advanced methods and related software for human reliability evaluation within probabilistic safety analyses

International Nuclear Information System (INIS)

Kosmowski, K.T.; Mertens, J.; Degen, G.; Reer, B.

1994-06-01

Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Analysis (PSA). The first part of this report consists of an overview of types of human behaviour and human error including the effect of significant performance shaping factors on human reliability. Particularly with regard to safety assessments for nuclear power plants a lot of HRA methods have been developed. The most important of these methods are presented and discussed in the report, together with techniques for incorporating HRA into PSA and with models of operator cognitive behaviour. Based on existing HRA methods the concept of a software system is described. For the development of this system the utilization of modern programming tools is proposed; the essential goal is the effective application of HRA methods. A possible integration of computeraided HRA within PSA is discussed. The features of Expert System Technology and examples of applications (PSA, HRA) are presented in four appendices. (orig.) [de

Implementation of an Enhanced Measurement Control Program for handling nuclear safety samples at WSRC

International Nuclear Information System (INIS)

Boler-Melton, C.; Holland, M.K.

1991-01-01

In the separation and purification of nuclear material, nuclear criticality safety (NCS) is of primary concern. The primary nuclear criticality safety controls utilized by the Savannah River Site (SRS) Separations Facilities involve administrative and process equipment controls. Additional assurance of NCS is obtained by identifying key process hold points where sampling is used to independently verify the effectiveness of production control. Nuclear safety measurements of samples from these key process locations provide a high degree of assurance that processing conditions are within administrative and procedural nuclear safety controls. An enhanced procedure management system aimed at making improvements in the quality, safety, and conduct of operation was implemented for Nuclear Safety Sample (NSS) receipt, analysis, and reporting. All procedures with nuclear safety implications were reviewed for accuracy and adequate detail to perform the analytical measurements safely, efficiently, and with the utmost quality. Laboratory personnel worked in a ''Deliberate Operating'' mode (a systematic process requiring continuous expert oversight during all phases of training, testing, and implementation) to initiate the upgrades. Thus, the effort to revise and review nuclear safety sample procedures involved a team comprised of a supervisor, chemist, and two technicians for each procedure. Each NSS procedure was upgraded to a ''Use Every Time'' (UET) procedure with sign-off steps to ensure compliance with each step for every nuclear safety sample analyzed. The upgrade program met and exceeded both the long and short term customer needs by improving measurement reliability, providing objective evidence of rigid adherence to program principles and requirements, and enhancing the system for independent verification of representative sampling from designated NCS points

Review of Reliability-Based Design Optimization Approach and Its Integration with Bayesian Method

Science.gov (United States)

Zhang, Xiangnan

2018-03-01

A lot of uncertain factors lie in practical engineering, such as external load environment, material property, geometrical shape, initial condition, boundary condition, etc. Reliability method measures the structural safety condition and determine the optimal design parameter combination based on the probabilistic theory. Reliability-based design optimization (RBDO) is the most commonly used approach to minimize the structural cost or other performance under uncertainty variables which combines the reliability theory and optimization. However, it cannot handle the various incomplete information. The Bayesian approach is utilized to incorporate this kind of incomplete information in its uncertainty quantification. In this paper, the RBDO approach and its integration with Bayesian method are introduced.

The advantages of reliability centered maintenance for standby safety systems

International Nuclear Information System (INIS)

Dam, R.F.; Ayazzudin, S.; Nickerson, J.H.; DeLong, A.I.

2002-01-01

Full text: On standby safety systems, nuclear plants have to balance the requirements of demonstrating the reliability of each system, while maintaining the system and plant availability. With the goal of demonstrating statistical reliability, these systems have extensive testing programs, which often makes the system unavailable and this can impact the plant capacity. The inputs to the process are often safety and regulatory related, resulting in programs that provide a high level of scrutiny on the systems being considered. In such cases, the value of the application of a maintenance optimization strategy, such as Reliability Centered Maintenance (RCM), is questioned. Part of the question stems from the use of the word 'Reliability' in RCM, which implies a level of redundancy when applied to a system maintenance program driven by reliability requirements. A deeper look at the RCM process, however, shows that RCM has the goal of ensuring that the system operates 'reliably' through the application of an integrated maintenance strategy. This is a subtle, but important distinction. Although the system reliability requirements are an important part of the strategy evaluation, RCM provides a broader context where testing is only one part of an overall strategy focused on ensuring that component function is maintained through a combination of monitoring technologies (including testing), predictive techniques, and intrusive maintenance strategies. Each strategy is targeted to identify known component degradation mechanisms. The conclusion is that a maintenance program driven by reliability requirements will tend to have testing defined at a frequency intended to support the needed statistics. The testing demonstrates that the desired function is available today. Maintenance driven by functional requirements and known failure causes, as developed through an RCM assessment, will have frequencies tied to industry experience with components and rely on a higher degree of

Safety improvement of Paks nuclear power plant

International Nuclear Information System (INIS)

Vamos, G.

1999-01-01

Safety upgrading completed in the early nineties at the Paks NPP include: replacement of steam generator safety valves and control valves; reliability improvement of the electrical supply system; modification of protection logic; enhancement of the fire protection; construction of full scope Training Simulator. Design safety upgrading measures achieved in recent years were concerned with: relocation of steam generator emergency feed-water supply; emergency gas removal from the primary coolant system; hydrogen management in the containment; protection against sumps; preventing of emergency core cooling system tanks from refilling. Increasing seismic resistance, containment assessment, refurbishment of reactor protection system, improving reliability of emergency electrical supply, analysis of internal hazards are now being implemented. Safety upgrading measures which are being prepared include: bleed and feed procedures; reactor over-pressurisation protection in cold state; treatment of steam generator primary to secondary leak accidents. Operational safety improvements are dealing with safety culture, training measures and facilities; symptom based emergency operating procedures; in-service inspection; fire protection. The significance of international cooperation is emphasised in view of achieving nuclear safety standards recognised in EU

Use of reliability engineering tools in safety and risk assessment of nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

Use of reliability engineering tools in safety and risk assessment of nuclear facilities

International Nuclear Information System (INIS)

Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias

2017-01-01

Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

Improving the Efficiency of Administrative Decision-Making when Monitoring Reliability and Safety of Oil and Gas Equipment

Directory of Open Access Journals (Sweden)

Zemenkova Maria

2016-01-01

Full Text Available Methodology of rapid assessment of reliability index was developed based on system analysis of technological parameters. Within functioning of on-line monitoring system of reliability index of industrial facility this method allows to increase efficiency of making managerial decisions on technical and preventive maintenance. The technique is based on the analysis of technological parameters of operational modes of pipeline transport facilities registered by dispatcher controls. The created technique can be used by the operating, research, design institutes and oil and gas transport enterprises when declaring industrial safety. The received mathematical models allow federal services of supervision, the independent expert organizations to predict the development of reliability in the registered block of dispatching data either in real time mode, or taking into account the dynamics of service conditions of the object.

The Use of Questionnaires in Safety Culture Studies in High Reliability Organizations. Literature Review and an Application in the Spanish Nuclear Sector

International Nuclear Information System (INIS)

German, S.; Navajas, J.; Silla, I.

2014-01-01

This report examines two aspects related to the use of questionnaires in safety culture research conducted in high reliability organizations. First, a literature review of recent studies that address safety culture through questionnaires is presented. Literature review showed that most studies used only questionnaires as a research technique, were cross-sectional, applied paper-based questionnaires, and were conducted in one type of high reliability organization. Second, a research project on safety culture that used electronic surveys in a sample of experts on safety culture is discussed. This project, developed by CISOT-CIEMAT research institute, was carry out in the Spanish nuclear sector and illustrates relevant aspects of the methodological design and administration processes that must be considered to encourage participation in the study.. (Author)

Reliability Analysis on NPP's Safety-Related Control Module with Field Data

International Nuclear Information System (INIS)

Lee, Sang Yong; Jung, Jae Hyun; Kim, Seong Hun

2006-01-01

The automatic control systems used in nuclear power plant (NPP) consists of numerous control modules that can be considered to be a network of components various complex ways. The control modules require relatively high reliability than industrial electronic products. Reliability prediction provides the rational basis of system designs and also provides the safety significance of system operations. The aim of this paper is to minimize the deficiencies of the traditional reliability prediction method calculation using the available field return data. This way is possible to do more realistic reliability assessment. SAMCHANG Enterprise Company (SEC) has established database containing high quality data at the module and component level from module maintenance in NPP. On the basis of these, this paper compares results that add failure record (field data) to Telcordia-SR-332 reliability prediction model with MIL-HDBK-217F prediction results

Methods and Effects of Safety Enhancement in Korean PSR

International Nuclear Information System (INIS)

Kim, Young Gab; Park, Jong Woon

2009-01-01

Periodic Safety Review (PSR) is a comprehensive study on a nuclear power plant safety, taking into account aspects such as operational history, ageing, safety analyses and advances in code and standards since the time of construction. In Korea, PSRs have been performed for 20 units and have been effectively used to obtain an overall view of actual plant safety to determine reasonable and practical modifications that should be made in order to obtain a higher level of safety approaching that of modern plants. Among many safety enhancements achieved from Korean PSRs, new safety analyses are the important methods to confirm plant safety by increasing safety margin for specific safety issues. Methods and effects of safety enhancements applied in Korean PSRs are reviewed in this paper in light of new safety analyses to obtain additional safety margins

Licensing process for safety-critical software-based systems

Energy Technology Data Exchange (ETDEWEB)

Haapanen, P. [VTT Automation, Espoo (Finland); Korhonen, J. [VTT Electronics, Espoo (Finland); Pulkkinen, U. [VTT Automation, Espoo (Finland)

2000-12-01

System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications

Licensing process for safety-critical software-based systems

International Nuclear Information System (INIS)

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

2000-12-01

System vendors nowadays propose software-based technology even for the most critical safety functions in nuclear power plants. Due to the nature of software faults and the way they cause system failures new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)', financed together by the Radiation and Nuclear Safety Authority (STUK), the Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland (VTT), various safety assessment methods and tools for software based systems are developed and evaluated. As a part of the OHA-work a reference model for the licensing process for software-based safety automation systems is defined. The licensing process is defined as the set of interrelated activities whose purpose is to produce and assess evidence concerning the safety and reliability of the system/application to be licensed and to make the decision about the granting the construction and operation permissions based on this evidence. The parties of the licensing process are the authority, the licensee (the utility company), system vendors and their subcontractors and possible external independent assessors. The responsibility about the production of the evidence in first place lies at the licensee who in most cases rests heavily on the vendor expertise. The evaluation and gauging of the evidence is carried out by the authority (possibly using external experts), who also can acquire additional evidence by using their own (independent) methods and tools. Central issue in the licensing process is to combine the quality evidence about the system development process with the information acquired through tests, analyses and operational experience. The purpose of the licensing process described in this report is to act as a reference model both for the authority and the licensee when planning the licensing of individual applications. Many of the

Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

International Nuclear Information System (INIS)

Lee, Sangseok; Sohn, Kwangyoung; Lee, Junku; Park, Geunok

2013-01-01

Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of controllers

Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangseok; Sohn, Kwangyoung [Korea Reliability Technology and System, Daejeon (Korea, Republic of); Lee, Junku; Park, Geunok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of

Current activities and future trends in reliability analysis and probabilistic safety assessment in Hungary

International Nuclear Information System (INIS)

Hollo, E.; Toth, J.

1986-01-01

In Hungary reliability analysis (RA) and probabilistic safety assessment (PSA) of nuclear power plants was initiated 3 years ago. First, computer codes for automatic fault tree analysis (CAT, PREP) and numerical evaluation (REMO, KITT1,2) were adapted. Two main case studies - detailed availability/reliability calculation of diesel sets and analysis of safety systems influencing event sequences induced by large LOCA - were performed. Input failure data were taken from publications, a need for failure and reliability data bank was revealed. Current and future activities involves: setup of national data bank for WWER-440 units; full-scope level-I PSA of PAKS NPP in Hungary; operational safety assessment of particular problems at PAKS NPP. In the present article the state of RA and PSA activities in Hungary, as well as the main objectives of ongoing work are described. A need for international cooperation (for unified data collection of WWER-440 units) and for IAEA support (within Interregional Program INT/9/063) is emphasized. (author)

Distributed fiber optic sensing enhances pipeline safety and security

Energy Technology Data Exchange (ETDEWEB)

Frings, Jochen; Walk, Tobias [ILF Consulting Engineers, Munich (Germany)

2011-09-15

Pipelines are efficient, highly reliable and safe means of transportation. However, despite intensive right of way surveillance by foot, car and out of the air, pipeline leaks and illegal tappings are a reality - sometimes with catastrophic results. These events show a gap in real-time monitoring caused by the highly distributed nature of pipelines. Parts of this gap now can be closed with distributed fiber optic sensing technology. Using various physical effects this technology is apt to detect temperature, strain, vibrations and sound with very good localization over spans up to 50 km with a single sensor cable. Various field tested applications like leakage detection, third party activity monitoring and intrusion detection or ground movement detection as well as integrity monitoring proof that distributed fiber optic sensing can enhance pipeline safety and security. (orig.)

Feasibility of AmbulanCe-Based Telemedicine (FACT study: safety, feasibility and reliability of third generation in-ambulance telemedicine.

Directory of Open Access Journals (Sweden)

Laetitia Yperzeele

Full Text Available Telemedicine is currently mainly applied as an in-hospital service, but this technology also holds potential to improve emergency care in the prehospital arena. We report on the safety, feasibility and reliability of in-ambulance teleconsultation using a telemedicine system of the third generation.A routine ambulance was equipped with a system for real-time bidirectional audio-video communication, automated transmission of vital parameters, glycemia and electronic patient identification. All patients ( ≥ 18 years transported during emergency missions by a Prehospital Intervention Team of the Universitair Ziekenhuis Brussel were eligible for inclusion. To guarantee mobility and to facilitate 24/7 availability, the teleconsultants used lightweight laptop computers to access a dedicated telemedicine platform, which also provided functionalities for neurological assessment, electronic reporting and prehospital notification of the in-hospital team. Key registrations included any safety issue, mobile connectivity, communication of patient information, audiovisual quality, user-friendliness and accuracy of the prehospital diagnosis.Prehospital teleconsultation was obtained in 41 out of 43 cases (95.3%. The success rates for communication of blood pressure, heart rate, blood oxygen saturation, glycemia, and electronic patient identification were 78.7%, 84.8%, 80.6%, 64.0%, and 84.2%. A preliminary prehospital diagnosis was formulated in 90.2%, with satisfactory agreement with final in-hospital diagnoses. Communication of a prehospital report to the in-hospital team was successful in 94.7% and prenotification of the in-hospital team via SMS in 90.2%. Failures resulted mainly from limited mobile connectivity and to a lesser extent from software, hardware or human error. The user acceptance was high.Ambulance-based telemedicine of the third generation is safe, feasible and reliable but further research and development, especially with regard to high

A methodology for a quantitative assessment of safety culture in NPPs based on Bayesian networks

International Nuclear Information System (INIS)

Kim, Young Gab; Lee, Seung Min; Seong, Poong Hyun

2017-01-01

Highlights: • A safety culture framework and a quantitative methodology to assess safety culture were proposed. • The relation among Norm system, Safety Management System and worker's awareness was established. • Safety culture probability at NPPs was updated by collecting actual organizational data. • Vulnerable areas and the relationship between safety culture and human error were confirmed. - Abstract: For a long time, safety has been recognized as a top priority in high-reliability industries such as aviation and nuclear power plants (NPPs). Establishing a safety culture requires a number of actions to enhance safety, one of which is changing the safety culture awareness of workers. The concept of safety culture in the nuclear power domain was established in the International Atomic Energy Agency (IAEA) safety series, wherein the importance of employee attitudes for maintaining organizational safety was emphasized. Safety culture assessment is a critical step in the process of enhancing safety culture. In this respect, assessment is focused on measuring the level of safety culture in an organization, and improving any weakness in the organization. However, many continue to think that the concept of safety culture is abstract and unclear. In addition, the results of safety culture assessments are mostly subjective and qualitative. Given the current situation, this paper suggests a quantitative methodology for safety culture assessments based on a Bayesian network. A proposed safety culture framework for NPPs would include the following: (1) a norm system, (2) a safety management system, (3) safety culture awareness of worker, and (4) Worker behavior. The level of safety culture awareness of workers at NPPs was reasoned through the proposed methodology. Then, areas of the organization that were vulnerable in terms of safety culture were derived by analyzing observational evidence. We also confirmed that the frequency of events involving human error

Research for enhancing reactor safety

International Nuclear Information System (INIS)

1989-05-01

Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

Use of a web site to enhance criticality safety training

International Nuclear Information System (INIS)

Huang, Song T.; Morman, James A.

2003-01-01

Establishment of the NCSP (Nuclear Criticality Safety Program) website represents one attempt by the NCS (Nuclear Criticality Safety) community to meet the need to enhance communication and disseminate NCS information to a wider audience. With the aging work force in this important technical field, there is a common recognition of the need to capture the corporate knowledge of these people and provide an easily accessible, web-based training opportunity to those people just entering the field of criticality safety. A multimedia-based site can provide a wide range of possibilities for criticality safety training. Training modules could range from simple text-based material, similar to the NCSET (Nuclear Criticality Safety Engineer Training) modules, to interactive web-based training classes, to video lecture series. For example, the Los Alamos National Laboratory video series of interviews with pioneers of criticality safety could easily be incorporated into training modules. Obviously, the development of such a program depends largely upon the need and participation of experts who share the same vision and enthusiasm of training the next generation of criticality safety engineers. The NCSP website is just one example of the potential benefits that web-based training can offer. You are encouraged to browse the NCSP website at http://ncsp.llnl.gov. We solicit your ideas in the training of future NCS engineers and welcome your participation with us in developing future multimedia training modules. (author)

Defining safety culture and the nexus between safety goals and safety culture. 4. Enhancing Safety Culture Through the Establishment of Safety Goals

International Nuclear Information System (INIS)

Tateiwa, Kenji; Miyata, Koichi; Yahagi, Kimitoshi

2001-01-01

efficient management. To seek compatibility between safety culture and efficient management, and to build an agreeable common perception among the utility, regulatory body, and the public on rationalizing the safety level to the extent acceptable, two issues must be considered: (a) establishing safety goals and (b) quantifying the safety culture. As for the first issue, currently no generic safety goal for the nuclear industry is available in Japan. This causes difficulty in deciding whether or not a specific action that takes place in a plant leading to a certain amount of risk increment is acceptable. Therefore, it is important for us to have a safety goal established. By establishing the safety goal, we could utilize it for the following usage: 1. to enable prompt response in case a safety level has entered an unacceptable level; 2. to sustain and enhance the safety culture centered by risk information (sustaining safety culture and achieving rational management simultaneously); 3. to use public relations for plant activities, where condemnation for overlooking safety culture might arrive for a minor risk increasing activity. As for the second issue, we must devise a method to quantify the level of safety culture. This could be done based on the safety culture indicators listed in the appendix of IAEA INSAG-4 (Ref. 1). By quantifying the level of safety culture, a comprehensive safety level of a plant can be evaluated by integrating with the safety level based on PSA. For example, when considering on-line maintenance for specific equipment, it is important not only to assess the safety level in terms of PSA but also to assess the level of safety culture in order to have a comprehensive view of the safety level and to compare it with the safety goal. By both establishing safety goals and quantifying the level of safety culture in a nuclear power plant, the ability to assess the comprehensive safety level of a plant and acknowledge the quantitative margin from the safety

Design of Korean nuclear reliability data-base network using a two-stage Bayesian concept

International Nuclear Information System (INIS)

Kim, T.W.; Jeong, K.S.; Chae, S.K.

1987-01-01

In an analysis of probabilistic risk, safety, and reliability of a nuclear power plant, the reliability data base (DB) must be established first. As the importance of the reliability data base increases, event reporting systems such as the US Nuclear Regulatory Commission's Licensee Event Report and the International Atomic Energy Agency's Incident Reporting System have been developed. In Korea, however, the systematic reliability data base is not yet available. Therefore, foreign data bases have been directly quoted in reliability analyses of Korean plants. In order to develop a reliability data base for Korean plants, the problem is which methodology is to be used, and the application limits of the selected method must be solved and clarified. After starting the commercial operation of Korea Nuclear Unit-1 (KNU-1) in 1978, six nuclear power plants have begun operation. Of these, only KNU-3 is a Canada Deuterium Uranium pressurized heavy-water reactor, and the others are all pressurized water reactors. This paper describes the proposed reliability data-base network (KNRDS) for Korean nuclear power plants in the context of two-stage Bayesian (TSB) procedure of Kaplan. It describes the concept of TSB to obtain the Korean-specific plant reliability data base, which is updated with the incorporation of both the reported generic reliability data and the operation experiences of similar plants

Reliability and safety program plan outline for the operational phase of a waste isolation facility

International Nuclear Information System (INIS)

Ammer, H.G.; Wood, D.E.

1977-01-01

A Reliability and Safety Program plan outline has been prepared for the operational phase of a Waste Isolation Facility. The program includes major functions of risk assessment, technical support activities, quality assurance, operational safety, configuration monitoring, reliability analysis and support and coordination meetings. Detailed activity or task descriptions are included for each function. Activities are time-phased and presented in the PERT format for scheduling and interactions. Task descriptions include manloading, travel, and computer time estimates to provide data for future costing. The program outlined here will be used to provide guidance from a reliability and safety standpoint to design, procurement, construction, and operation of repositories for nuclear waste. These repositories are to be constructed under the National Waste Terminal Storage program under the direction of the Office of Waste Isolation, Union Carbide Corp. Nuclear Division

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

Assessment of reliability of a safety culture questionnaire in the cleanser and washer industries

Directory of Open Access Journals (Sweden)

2012-09-01

Full Text Available Introduction: Occupational injuries and accidents as one of the problems have always been considered important in occupational environments. Domino model that Heinrich was formed to pursue the idea of the cause of the accident is the man. Thus one of the effective way to reduce accidents will be control by the unsafe behaviors among workers by promoting safety culture. .Material and Method: In this descriptive - analytical study, the reliability and exploratory factor analysis was used to evaluate the reliability of the questionnaire. In total 303 questionnaires were analyzed using SPSS 17 software. . Result: The alpha crumbed, coefficient was 0/86. Structural factor of the questionnaire was evaluated using factor analysis. KMO and Bartlett’s sphericity test coefficient were 0/909 and 9785/057, respectively. The varimax rotation showed that all test questions are based on factors. .Conclusion: The results indicated favorable validity of this questionnaire for use in detergents and cleaners industries within the country. Considering the load factor safety culture in detergents and cleaners industries, contained 5 factors including “management commitment”, “education and information exchange,” “supportive environment”, “barriers” and “priority to safety”. The obtained the correlations, the highest positive correlation was belong to the “management commitment” (r=0/952, as the strongest correlation with the safety culture.

Software Reliability Issues Concerning Large and Safety Critical Software Systems

Science.gov (United States)

Kamel, Khaled; Brown, Barbara

1996-01-01

This research was undertaken to provide NASA with a survey of state-of-the-art techniques using in industrial and academia to provide safe, reliable, and maintainable software to drive large systems. Such systems must match the complexity and strict safety requirements of NASA's shuttle system. In particular, the Launch Processing System (LPS) is being considered for replacement. The LPS is responsible for monitoring and commanding the shuttle during test, repair, and launch phases. NASA built this system in the 1970's using mostly hardware techniques to provide for increased reliability, but it did so often using custom-built equipment, which has not been able to keep up with current technologies. This report surveys the major techniques used in industry and academia to ensure reliability in large and critical computer systems.

DOPO-Modified Two-Dimensional Co-Based Metal-Organic Framework: Preparation and Application for Enhancing Fire Safety of Poly(lactic acid).

Science.gov (United States)

Hou, Yanbei; Liu, Longxiang; Qiu, Shuilai; Zhou, Xia; Gui, Zhou; Hu, Yuan

2018-03-07

Co-based metal-organic framework (Co-MOF) nanosheets were successfully synthesized by the organic ligands with Schiff base structure. The laminated structure gives Co-MOF nanosheets a great advantage in the application in the flame retardant field. Meanwhile, -C═N- from Schiff base potentially provides active sites for further modification. In this work, 9,10-dihydro-9-oxa-10-phosphaphenanthrene-10-oxide (DOPO) was used to modify Co-MOF (DOPO@Co-MOF) to further enhance its flame retardant efficiency. It is attractive that DOPO has a synergistic effect with Co-MOF on improving fire safety of poly(lactic acid) (PLA). The obvious decrease in the values of peak heat release (27%), peak smoke production (56%), and total CO yield (20%) confirmed the enhanced fire safety of PLA composites. The possible flame retardant mechanism was proposed based on characterization results. Moreover, the addition of DOPO@Co-MOF had a positive influence on the mechanical performance, including tensile properties and impact resistance. This work designed and synthesized two-dimensional MOFs with active groups. As-prepared Co-MOF with expected structure shows a novel direction of preparing MOFs for flame retardant application.

Kullback-Leibler distance-based enhanced detection of incipient anomalies

KAUST Repository

Harrou, Fouzi; Sun, Ying; Madakyaru, Muddu

2016-01-01

Accurate and effective anomaly detection and diagnosis of modern engineering systems by monitoring processes ensure reliability and safety of a product while maintaining desired quality. In this paper, an innovative method based on Kullback

Proceedings of the Digital Systems Reliability and Nuclear Safety Workshop

Energy Technology Data Exchange (ETDEWEB)

Wallace, D. R.; Cuthill, B. B.; Ippolito, L. M. [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Beltracchi, L. [Nuclear Regulatory Commission, Washington, DC (United States) ed.

1994-03-01

The United States Nuclear Regulatory Commission (NRC), in cooperation with the National Institute of Standards and Technology conducted the.Digital Systems Reliability and Nuclear Safety Workshop on September 13--14, 1993, in Rockville, Maryland. The workshop provided a forum for the exchange of information among experts within the nuclear industry, experts from other industries, regulators and academia. The information presented at this workshop provided in-depth exposure of the NRC staff and the nuclear industry to digital systems design safety issues and also provided feedback to the NRC from outside experts regarding identified safety issues, proposed regulatory positions, and intended research associated with the use of digital systems in nuclear power plants. Technical presentations provided insights on areas where current software engineering practices may be inadequate for safety-critical systems, on potential solutions for development issues, and on methods for reducing risk in safety-critical systems. This report contains an analysis of results of the workshop, the papers presented panel presentations, and summaries of, discussions at this workshop. The individual papers have been cataloged separately.

Reliability Assessment and Reliability-Based Inspection and Maintenance of Offshore Wind Turbines

DEFF Research Database (Denmark)

Ramirez, José Rangel

actions are the most relevant and effective means of control of deterioration. The risk-based inspection planning methodology, based on Bayesian decision theory, represents an important tool to identify the suitable strategy to inspect and control the deterioration in structures such as offshore wind...... performance during the life cycle. The deterioration processes, such as fatigue and corrosion, are typically affecting offshore structural systems. This damage decreases the system performance and increases the risk of failure, thus not fulfilling the established safety criteria. Inspection and maintenance...... to their offshore location, no pollution risks and low human risks since they are unmanned. This allows the allocation of lower reliability level compared to e.g. oil & gas installations. With the incursion to water depths between 20 and 50 meters, the use of jacket and tripod structures represents a feasible...

Safety instrumented systems in the oil and gas industry : Concepts and methods for safety and reliability assessments in design and operation

Energy Technology Data Exchange (ETDEWEB)

Lundteigen, Mary Ann

2009-07-01

This thesis proposes new methods and gives new insight to safety and reliability assessments of safety instrumented systems (SISs). These systems play an important role in many industry sectors and are used to detect the onset of hazardous events and mitigate their consequences to humans, the environment, and material assets. The thesis focuses on SIS applications in the oil and gas industry. Here, the SIS must respond to hazardous events such as gas leakages, fires, and over pressurization. Because there are personnel onboard the oil and gas installations, the operations take place in a vulnerable marine environment, and substantial values are associated with the offshore facilities, the reliability of SIS is of great concern to the public, the authorities, and the plant owners. The objective of this project has been to identify some of the key factors that influence the SIS reliability, clarify their effects on reliability, and suggest means to improve the treatment of these factors in safety and reliability assessments in design and operation. The project builds on concepts, methods, and definitions in two key standards for SIS design, construction, and operation: IEC 61508 and IEC 61511. The main contributions from this project are: A product development model that integrates reliability, availability, maintainability, and safety (RAMS) requirements with product development. The contributions have been presented in ten articles, five published in international journals, two submitted for publication, and three presented at conferences and in conference proceedings. The contributions are also directed to the industry and the actors that are involved in SIS design, construction, and operation. Even if the oil and gas industry is the main focus area, the results may be relevant for other industry sectors as well. SIS manufacturers and SIS designers face a large number of requirements from authorities, oil companies, international standards, and so on. At the same

Near-misses are an opportunity to improve patient safety: adapting strategies of high reliability organizations to healthcare.

Science.gov (United States)

Van Spall, Harriette; Kassam, Alisha; Tollefson, Travis T

2015-08-01

Near-miss investigations in high reliability organizations (HROs) aim to mitigate risk and improve system safety. Healthcare settings have a higher rate of near-misses and subsequent adverse events than most high-risk industries, but near-misses are not systematically reported or analyzed. In this review, we will describe the strategies for near-miss analysis that have facilitated a culture of safety and continuous quality improvement in HROs. Near-miss analysis is routine and systematic in HROs such as aviation. Strategies implemented in aviation include the Commercial Aviation Safety Team, which undertakes systematic analyses of near-misses, so that findings can be incorporated into Standard Operating Procedures (SOPs). Other strategies resulting from incident analyses include Crew Resource Management (CRM) for enhanced communication, situational awareness training, adoption of checklists during operations, and built-in redundancy within systems. Health care organizations should consider near-misses as opportunities for quality improvement. The systematic reporting and analysis of near-misses, commonplace in HROs, can be adapted to health care settings to prevent adverse events and improve clinical outcomes.

KHNP Safety Culture Framework based on Global Standard, and Lessons learned from Safety Culture Evaluation

International Nuclear Information System (INIS)

Kim, Younggab; Hur, Nam Young; Jeong, Hyeon Jong

2015-01-01

In order to eliminate the vague fears of the people about the nuclear power and operate continuously NPPs, a strong safety culture of NPPs should be demonstrated. Strong safety culture awareness of workers can overcome social distrust about NPPs. KHNP has been a variety efforts to improve and establish safety culture of NPPs. Safety culture framework applying global standards was set up and safety culture assessment has been carried out periodically to enhance safety culture of workers. In addition, KHNP developed various safety culture contents and they are being used in NPPs by workers. As a result of these efforts, safety culture awareness of workers is changed positively and the safety environment of NPPs is expected to be improved. KHNP makes an effort to solve areas for improvement derived from safety culture assessment. However, there are some areas to take a long time in completing the work. Therefore, these actions are necessary to be carried out consistently and continuously. KHNP also developed recently safety culture enhancement system based on web. All information related to safety culture in KHNP will be shared through this web system and this system will be used to safety culture assessment. In addition to, KHNP plans to develop safety culture indicators for monitoring the symptoms of safety culture weakening

KHNP Safety Culture Framework based on Global Standard, and Lessons learned from Safety Culture Evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Younggab; Hur, Nam Young; Jeong, Hyeon Jong [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-05-15

In order to eliminate the vague fears of the people about the nuclear power and operate continuously NPPs, a strong safety culture of NPPs should be demonstrated. Strong safety culture awareness of workers can overcome social distrust about NPPs. KHNP has been a variety efforts to improve and establish safety culture of NPPs. Safety culture framework applying global standards was set up and safety culture assessment has been carried out periodically to enhance safety culture of workers. In addition, KHNP developed various safety culture contents and they are being used in NPPs by workers. As a result of these efforts, safety culture awareness of workers is changed positively and the safety environment of NPPs is expected to be improved. KHNP makes an effort to solve areas for improvement derived from safety culture assessment. However, there are some areas to take a long time in completing the work. Therefore, these actions are necessary to be carried out consistently and continuously. KHNP also developed recently safety culture enhancement system based on web. All information related to safety culture in KHNP will be shared through this web system and this system will be used to safety culture assessment. In addition to, KHNP plans to develop safety culture indicators for monitoring the symptoms of safety culture weakening.

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

Information about robustness, reliability and safety in early design phases

DEFF Research Database (Denmark)

Marini, Vinicius Kaster

methods, and an industrial case to assess how the use of information about robustness, reliability and safety as practised by current methods influences concept development. Current methods cannot be used in early design phases due to their dependence on detailed design information for the identification...... alternatives. This prompts designers to reuse working principles that are inherently flawed, as they are liable to disturbances, failures and hazards. To address this issue, an approach based upon individual records of early design issues consists of comparing failures and benefits from prior working...... principles, before making a decision, and improving the more suitable alternatives through this feedback. Workshops were conducted with design practitioners to evaluate the potential of the approach and to simulate decision-making and gain feedback on a proof-of-concept basis. The evaluation has demonstrated...

Improved reliability, maintainability and safety through elastomer upgrading

International Nuclear Information System (INIS)

Wensel, R.; Wittich, K.C.

1995-01-01

Equipment in nuclear plants has historically contained whatever elastomer each component supplier traditionally used for corresponding non-nuclear service. The resulting proliferation of elastomer compounds, many of which are far from optimal for the service conditions (e.g., pressure, temperature, radiation, etc.), has multiplied the costs to provide station reliability, maintainability and safety. Cost-effective improvements are being achieved in CANDU plants by upgrading and standardizing on a handful of high performing elastomer compounds. These upgraded materials offer significant gains in service life over the materials they replace (often by factors of 2 or more). This rationalization of elastomer compounds also facilitates the EQ process for safety-related equipment. Detailed test data on aging is currently being generated for these specific elastomers, encompassing the conditions and media (air, water, oil) common in CANDU service. Two key elements characterize this testing. First, each result is specific to the compound used in the test, and second, it is specific to the tested failure mode (e.g., compression set, extrusion, fracture, etc.). Having fewer, but more thoroughly tested compounds, avoids the penalty (associated with poorly characterized materials) of having to replace parts prematurely because of conservatism, while maintaining safe, reliable service. This paper provides an overview of this approach covering: the benefits of compound rationalization; and the how and why of establishing relevant failure criteria; appropriate quality assurance to maintain EQ; procurement, storage and handling guidelines; and monitoring and predicting in-service degradation. (author)

Risk-based reconfiguration of safety monitoring system using dynamic Bayesian network

International Nuclear Information System (INIS)

Kohda, Takehisa; Cui Weimin

2007-01-01

To prevent an abnormal event from leading to an accident, the role of its safety monitoring system is very important. The safety monitoring system detects symptoms of an abnormal event to mitigate its effect at its early stage. As the operation time passes by, the sensor reliability decreases, which implies that the decision criteria of the safety monitoring system should be modified depending on the sensor reliability as well as the system reliability. This paper presents a framework for the decision criteria (or diagnosis logic) of the safety monitoring system. The logic can be dynamically modified based on sensor output data monitored at regular intervals to minimize the expected loss caused by two types of safety monitoring system failure events: failed-dangerous (FD) and failed-safe (FS). The former corresponds to no response under an abnormal system condition, while the latter implies a spurious activation under a normal system condition. Dynamic Bayesian network theory can be applied to modeling the entire system behavior composed of the system and its safety monitoring system. Using the estimated state probabilities, the optimal decision criterion is given to obtain the optimal diagnosis logic. An illustrative example of a three-sensor system shows the merits and characteristics of the proposed method, where the reasonable interpretation of sensor data can be obtained

Contribution to a quantitative assessment model for reliability-based metrics of electronic and programmable safety-related functions; Contribution a un modele d'evaluation quantitative des performances fiabilistes de fonctions electroniques et programmables dediees a la securite

Energy Technology Data Exchange (ETDEWEB)

Hamidi, K

2005-10-15

The use of fault-tolerant EP architectures has induced growing constraints, whose influence on reliability-based performance metrics is no more negligible. To face up the growing influence of simultaneous failure, this thesis proposes, for safety-related functions, a new-trend assessment method of reliability, based on a better taking into account of time-aspect. This report introduces the concept of information and uses it to interpret the failure modes of safety-related function as the direct result of the initiation and propagation of erroneous information until the actuator-level. The main idea is to distinguish the apparition and disappearance of erroneous states, which could be defined as intrinsically dependent of HW-characteristic and maintenance policies, and their possible activation, constrained through architectural choices, leading to the failure of safety-related function. This approach is based on a low level on deterministic SED models of the architecture and use non homogeneous Markov chains to depict the time-evolution of probabilities of errors. (author)

[Does simulator-based team training improve patient safety?].

Science.gov (United States)

Trentzsch, H; Urban, B; Sandmeyer, B; Hammer, T; Strohm, P C; Lazarovici, M

2013-10-01

Patient safety became paramount in medicine as well as in emergency medicine after it was recognized that preventable, adverse events significantly contributed to morbidity and mortality during hospital stay. The underlying errors cannot usually be explained by medical technical inadequacies only but are more due to difficulties in the transition of theoretical knowledge into tasks under the conditions of clinical reality. Crew Resource Management and Human Factors which determine safety and efficiency of humans in complex situations are suitable to control such sources of error. Simulation significantly improved safety in high reliability organizations, such as the aerospace industry.Thus, simulator-based team training has also been proposed for medical areas. As such training is consuming in cost, time and human resources, the question of the cost-benefit ratio obviously arises. This review outlines the effects of simulator-based team training on patient safety. Such course formats are not only capable of creating awareness and improvements in safety culture but also improve technical team performance and emphasize team performance as a clinical competence. A few studies even indicated improvement of patient-centered outcome, such as a reduced rate of adverse events but further studies are required in this respect. In summary, simulator-based team training should be accepted as a suitable strategy to improve patient safety.

Proceedings of the international symposium on safety and reliability systems of PWRs and BWRs

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-05-01

Out of 33 contributions presented at the conference, 30 were submitted to INIS. The conference programme was divided into three sections: (i) Diagnostics and in-service inspection; (ii) Safety and reliability of NPP operation; (iii) Experience of NPP operation and new approaches to nuclear safety. (J.B.).

Proceedings of the international symposium on safety and reliability systems of PWRs and BWRs

International Nuclear Information System (INIS)

1996-02-01

Out of 33 contributions presented at the conference, 30 were submitted to INIS. The conference programme was divided into three sections: (i) Diagnostics and in-service inspection; (ii) Safety and reliability of NPP operation; (iii) Experience of NPP operation and new approaches to nuclear safety. (J.B.)

A probabilistic bridge safety evaluation against floods.

Science.gov (United States)

Liao, Kuo-Wei; Muto, Yasunori; Chen, Wei-Lun; Wu, Bang-Ho

2016-01-01

To further capture the influences of uncertain factors on river bridge safety evaluation, a probabilistic approach is adopted. Because this is a systematic and nonlinear problem, MPP-based reliability analyses are not suitable. A sampling approach such as a Monte Carlo simulation (MCS) or importance sampling is often adopted. To enhance the efficiency of the sampling approach, this study utilizes Bayesian least squares support vector machines to construct a response surface followed by an MCS, providing a more precise safety index. Although there are several factors impacting the flood-resistant reliability of a bridge, previous experiences and studies show that the reliability of the bridge itself plays a key role. Thus, the goal of this study is to analyze the system reliability of a selected bridge that includes five limit states. The random variables considered here include the water surface elevation, water velocity, local scour depth, soil property and wind load. Because the first three variables are deeply affected by river hydraulics, a probabilistic HEC-RAS-based simulation is performed to capture the uncertainties in those random variables. The accuracy and variation of our solutions are confirmed by a direct MCS to ensure the applicability of the proposed approach. The results of a numerical example indicate that the proposed approach can efficiently provide an accurate bridge safety evaluation and maintain satisfactory variation.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

Program integration of predictive maintenance with reliability centered maintenance

International Nuclear Information System (INIS)

Strong, D.K. Jr; Wray, D.M.

1990-01-01

This paper addresses improving the safety and reliability of power plants in a cost-effective manner by integrating the recently developed reliability centered maintenance techniques with the traditional predictive maintenance techniques of nuclear power plants. The topics of the paper include a description of reliability centered maintenance (RCM), enhancing RCM with predictive maintenance, predictive maintenance programs, condition monitoring techniques, performance test techniques, the mid-Atlantic Reliability Centered Maintenance Users Group, test guides and the benefits of shared guide development

Reliability and validity enhancement: a treatment package for increasing fidelity of self-report.

Science.gov (United States)

Bornstein, P H; Hamilton, S B; Miller, R K; Quevillon, R P; Spitzform, M

1977-07-01

This study investigated the effects of reliability and validity "enhancers" on fidelity of self-report data in an analogue therapy situation. Under the guise of a Concentration Skills Training Program, 57 Ss were assigned randomly to one of the following conditions: (a) Reliability Enhancement; (b) Truth Talk; (c) No Comment Control. Results indicated significant differences among groups (p less than .05). In addition, tests of multiple comparisons revealed that Reliability Enhancement was significantly different from Truth Talk in occurrences of unreliability (p less than .05). These findings are discussed in light of the increased reliance on self-report data in behavioral intervention, and recommendations are made for future research.

Application of Cold Chain Logistics Safety Reliability in Fresh Food Distribution Optimization

OpenAIRE

Zou Yifeng; Xie Ruhe

2013-01-01

In view of the nature of fresh food’s continuous decrease of safety during distribution process, this study applied safety reliability of food cold chain logistics to establish fresh food distribution routing optimization model with time windows, and solved the model using MAX-MIN Ant System (MMAS) with case analysis. Studies have shown that the mentioned model and algorithm can better solve the problem of fresh food distribution routing optimization with time windows.

Naive Bayes Bearing Fault Diagnosis Based on Enhanced Independence of Data.

Science.gov (United States)

Zhang, Nannan; Wu, Lifeng; Yang, Jing; Guan, Yong

2018-02-05

The bearing is the key component of rotating machinery, and its performance directly determines the reliability and safety of the system. Data-based bearing fault diagnosis has become a research hotspot. Naive Bayes (NB), which is based on independent presumption, is widely used in fault diagnosis. However, the bearing data are not completely independent, which reduces the performance of NB algorithms. In order to solve this problem, we propose a NB bearing fault diagnosis method based on enhanced independence of data. The method deals with data vector from two aspects: the attribute feature and the sample dimension. After processing, the classification limitation of NB is reduced by the independence hypothesis. First, we extract the statistical characteristics of the original signal of the bearings effectively. Then, the Decision Tree algorithm is used to select the important features of the time domain signal, and the low correlation features is selected. Next, the Selective Support Vector Machine (SSVM) is used to prune the dimension data and remove redundant vectors. Finally, we use NB to diagnose the fault with the low correlation data. The experimental results show that the independent enhancement of data is effective for bearing fault diagnosis.

Naive Bayes Bearing Fault Diagnosis Based on Enhanced Independence of Data

Science.gov (United States)

Zhang, Nannan; Wu, Lifeng; Yang, Jing; Guan, Yong

2018-01-01

The bearing is the key component of rotating machinery, and its performance directly determines the reliability and safety of the system. Data-based bearing fault diagnosis has become a research hotspot. Naive Bayes (NB), which is based on independent presumption, is widely used in fault diagnosis. However, the bearing data are not completely independent, which reduces the performance of NB algorithms. In order to solve this problem, we propose a NB bearing fault diagnosis method based on enhanced independence of data. The method deals with data vector from two aspects: the attribute feature and the sample dimension. After processing, the classification limitation of NB is reduced by the independence hypothesis. First, we extract the statistical characteristics of the original signal of the bearings effectively. Then, the Decision Tree algorithm is used to select the important features of the time domain signal, and the low correlation features is selected. Next, the Selective Support Vector Machine (SSVM) is used to prune the dimension data and remove redundant vectors. Finally, we use NB to diagnose the fault with the low correlation data. The experimental results show that the independent enhancement of data is effective for bearing fault diagnosis. PMID:29401730

Design, construction, qualification and reliability of main components, from the safety aspect

International Nuclear Information System (INIS)

Crette, J.P.

1982-01-01

In FRANCE, the design and construction of reliable components, which condition the safe operation and availability of breeder plants, is based on the experience acquired during the operation of RAPSODIE, PHENIX and the various test facilities. The technical progress achieved on all main components is illustrated by examples taken from the CREYS-MALVILLE plant. In parallel with the development of these components, an extensive program covering research, development and the definition of design, construction and inspection rules, together with scheduling and quality assurance methods, prepares the industrialization of this reactor system, in compliance with the rules and recommendations issued by the pertinent safety authorities

Perspective channel-type reactor with enhanced safety

International Nuclear Information System (INIS)

Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

1994-01-01

Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

Reliability analysis of safety systems of nuclear power plant and utility experience with reliability safeguarding of systems during specified normal operation

International Nuclear Information System (INIS)

Balfanz, H.P.

1989-01-01

The paper gives an outline of the methods applied for reliability analysis of safety systems in nuclear power plant. The main tasks are to check the system design for detection of weak points, and to find possibilities of optimizing the strategies for inspection, inspection intervals, maintenance periods. Reliability safeguarding measures include the determination and verification of the broundary conditions of the analysis with regard to the reliability parameters and maintenance parameters used in the analysis, and the analysis of data feedback reflecting the plant response during operation. (orig.) [de

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

International Nuclear Information System (INIS)

Boring, Ronald; Mandelli, Diego; Rasmussen, Martin; Ulrich, Thomas; Groth, Katrina; Smith, Curtis

2016-01-01

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rasmussen, Martin [Norwegian Univ. of Science and Technology, Trondheim (Norway). Social Research; Herberger, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ulrich, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

On Improving Reliability of SRAM-Based Physically Unclonable Functions

Directory of Open Access Journals (Sweden)

Arunkumar Vijayakumar

2017-01-01

Full Text Available Physically unclonable functions (PUFs have been touted for their inherent resistance to invasive attacks and low cost in providing a hardware root of trust for various security applications. SRAM PUFs in particular are popular in industry for key/ID generation. Due to intrinsic process variations, SRAM cells, ideally, tend to have the same start-up behavior. SRAM PUFs exploit this start-up behavior. Unfortunately, not all SRAM cells exhibit reliable start-up behavior due to noise susceptibility. Hence, design enhancements are needed for improving reliability. Some of the proposed enhancements in literature include fuzzy extraction, error-correcting codes and voting mechanisms. All enhancements involve a trade-off between area/power/performance overhead and PUF reliability. This paper presents a design enhancement technique for reliability that improves upon previous solutions. We present simulation results to quantify improvement in SRAM PUF reliability and efficiency. The proposed technique is shown to generate a 128-bit key in ≤0.2 μ s at an area estimate of 4538 μ m 2 with error rate as low as 10 − 6 for intrinsic error probability of 15%.

Operational safety performance indicator system - a management tool for the self assessment of safety and reliability of nuclear power plants

International Nuclear Information System (INIS)

Anil Kumar; Mandowara, S.L.; Mittal, S.

2006-01-01

Operational Safety Performance Indicator system is one of the self assessment tools for station management to monitor safety and reliability of nuclear power plants. It provides information to station management about the performance of various areas of the plants by means of different colours of relevant performance indicators. Such systems have been implemented at many nuclear power plants in the world and have been considered as strength during WANO Peer Review. IAEA had a Coordinated Research Programme (CRP) on this with several countries participating including India. In NPCIL this system has been implemented in KAPS about a year back and found very useful in identifying areas which needs to be given more attention. Based on the KAPS feedback Implementation of this system has been taken up in RAPS-3 and 4 and KGS-l and 2. (author)

Achieving High Reliability with People, Processes, and Technology.

Science.gov (United States)

Saunders, Candice L; Brennan, John A

2017-01-01

High reliability as a corporate value in healthcare can be achieved by meeting the "Quadruple Aim" of improving population health, reducing per capita costs, enhancing the patient experience, and improving provider wellness. This drive starts with the board of trustees, CEO, and other senior leaders who ingrain high reliability throughout the organization. At WellStar Health System, the board developed an ambitious goal to become a top-decile health system in safety and quality metrics. To achieve this goal, WellStar has embarked on a journey toward high reliability and has committed to Lean management practices consistent with the Institute for Healthcare Improvement's definition of a high-reliability organization (HRO): one that is committed to the prevention of failure, early identification and mitigation of failure, and redesign of processes based on identifiable failures. In the end, a successful HRO can provide safe, effective, patient- and family-centered, timely, efficient, and equitable care through a convergence of people, processes, and technology.

Reliability-Based Optimization in Structural Engineering

DEFF Research Database (Denmark)

Enevoldsen, I.; Sørensen, John Dalsgaard

1994-01-01

In this paper reliability-based optimization problems in structural engineering are formulated on the basis of the classical decision theory. Several formulations are presented: Reliability-based optimal design of structural systems with component or systems reliability constraints, reliability...

Physician attitudes towards pharmacological cognitive enhancement: safety concerns are paramount.

Directory of Open Access Journals (Sweden)

Opeyemi C Banjo

2010-12-01

Full Text Available The ethical dimensions of pharmacological cognitive enhancement have been widely discussed in academic circles and the popular media, but missing from the conversation have been the perspectives of physicians - key decision makers in the adoption of new technologies into medical practice. We queried primary care physicians in major urban centers in Canada and the United States with the aim of understanding their attitudes towards cognitive enhancement. Our primary hypothesis was that physicians would be more comfortable prescribing cognitive enhancers to older patients than to young adults. Physicians were presented with a hypothetical pharmaceutical cognitive enhancer that had been approved by the regulatory authorities for use in healthy adults, and was characterized as being safe, effective, and without significant adverse side effects. Respondents overwhelmingly reported increasing comfort with prescribing cognitive enhancers as the patient age increased from 25 to 65. When asked about their comfort with prescribing extant drugs that might be considered enhancements (sildenafil, modafinil, and methylphenidate or our hypothetical cognitive enhancer to a normal, healthy 40 year old, physicians were more comfortable prescribing sildenafil than any of the other three agents. When queried as to the reasons they answered as they did, the most prominent concerns physicians expressed were issues of safety that were not offset by the benefit afforded the individual, even in the face of explicit safety claims. Moreover, many physicians indicated that they viewed safety claims with considerable skepticism. It has become routine for safety to be raised and summarily dismissed as an issue in the debate over pharmacological cognitive enhancement; the observation that physicians were so skeptical in the face of explicit safety claims suggests that such a conclusion may be premature. Thus, physician attitudes suggest that greater weight be placed upon the

Enhancing the effectiveness of IST through risk-based techniques

Energy Technology Data Exchange (ETDEWEB)

Floyd, S.D.

1996-12-01

Current IST requirements were developed mainly through deterministic-based methods. While this approach has resulted in an adequate level of safety and reliability for pumps and valves, insights from probabilistic safety assessments suggest a better safety focus can be achieved at lower costs. That is, some high safety impact pumps and valves are currently not tested under the IST program and should be added, while low safety impact valves could be tested at significantly greater intervals than allowed by the current IST program. The nuclear utility industry, through the Nuclear Energy Institute (NEI), has developed a draft guideline for applying risk-based techniques to focus testing on those pumps and valves with a high safety impact while reducing test frequencies on low safety impact pumps and valves. The guideline is being validated through an industry pilot application program that is being reviewed by the U.S. Nuclear Regulatory Commission. NEI and the ASME maintain a dialogue on the two groups` activities related to risk-based IST. The presenter will provide an overview of the NEI guideline, discuss the methodological approach for applying risk-based technology to IST and provide the status of the industry pilot plant effort.

Enhancing reliability of ultrasonic testing of welds of nuclear power plant equipment

International Nuclear Information System (INIS)

Shcherbinskij, V.G.

1981-01-01

Results of investigation of factors influencing the reliability of manual ultrasonic testing of welded joints and weld deposited metal power-generating equipment are presented. Recommendations on the enhancing of reliability are given [ru

Contribution to a quantitative assessment model for reliability-based metrics of electronic and programmable safety-related functions; Contribution a un modele d'evaluation quantitative des performances fiabilistes de fonctions electroniques et programmables dediees a la securite

Energy Technology Data Exchange (ETDEWEB)

Hamidi, K

2005-10-15

The use of fault-tolerant EP architectures has induced growing constraints, whose influence on reliability-based performance metrics is no more negligible. To face up the growing influence of simultaneous failure, this thesis proposes, for safety-related functions, a new-trend assessment method of reliability, based on a better taking into account of time-aspect. This report introduces the concept of information and uses it to interpret the failure modes of safety-related function as the direct result of the initiation and propagation of erroneous information until the actuator-level. The main idea is to distinguish the apparition and disappearance of erroneous states, which could be defined as intrinsically dependent of HW-characteristic and maintenance policies, and their possible activation, constrained through architectural choices, leading to the failure of safety-related function. This approach is based on a low level on deterministic SED models of the architecture and use non homogeneous Markov chains to depict the time-evolution of probabilities of errors. (author)

An Approach to Enhancement of the Safety Culture of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The culture of an organization is very complex to study and evaluate, but it is possible to examine the specific norms that figure out the culture for the safety of a system. This paper describes an initiation plan to study the safety culture issue in Korean NPPs. Recently there happened successively events that turned out to be socially prominent in Korea. Many issues on the safety culture aspect of NPPs have been raised including the types of errors such as violations, an intended concealment of safety-related information, counterfeit items, forgery process in procurement, and so on. Those were investigated in detail for the root causes of these issues as human and organizational errors and for the countermeasures to prevent those events. They are integrated into a correspondent long-term plan including the establishment of a fundamental infrastructure of safety culture management for operating NPPs in Korea. A monitoring system with analysis functions utilizing system dynamics simulation and data mining is proposed to be incorporated into a safety culture management system. Additionally, a set of training and support programs are to be developed for the enhancement of some selected competence of the operating personnel in Korean NPPs. The safe operation of NPPs requires the typical safety culture characteristics of the high reliability organization (HRO). The culture of an organization is very complex to study and evaluate, but it is possible to examine the specific norms that figure out the culture for the safety of a system. This paper describes an integrated systems approach as an initiating plan to study the safety culture issue in Korean NPPs.

An Approach to Enhancement of the Safety Culture of Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Yong Hee

2014-01-01

The culture of an organization is very complex to study and evaluate, but it is possible to examine the specific norms that figure out the culture for the safety of a system. This paper describes an initiation plan to study the safety culture issue in Korean NPPs. Recently there happened successively events that turned out to be socially prominent in Korea. Many issues on the safety culture aspect of NPPs have been raised including the types of errors such as violations, an intended concealment of safety-related information, counterfeit items, forgery process in procurement, and so on. Those were investigated in detail for the root causes of these issues as human and organizational errors and for the countermeasures to prevent those events. They are integrated into a correspondent long-term plan including the establishment of a fundamental infrastructure of safety culture management for operating NPPs in Korea. A monitoring system with analysis functions utilizing system dynamics simulation and data mining is proposed to be incorporated into a safety culture management system. Additionally, a set of training and support programs are to be developed for the enhancement of some selected competence of the operating personnel in Korean NPPs. The safe operation of NPPs requires the typical safety culture characteristics of the high reliability organization (HRO). The culture of an organization is very complex to study and evaluate, but it is possible to examine the specific norms that figure out the culture for the safety of a system. This paper describes an integrated systems approach as an initiating plan to study the safety culture issue in Korean NPPs

John F. Kennedy Space Center, Safety, Reliability, Maintainability and Quality Assurance, Survey and Audit Program

Science.gov (United States)

1994-01-01

This document is the product of the KSC Survey and Audit Working Group composed of civil service and contractor Safety, Reliability, and Quality Assurance (SR&QA) personnel. The program described herein provides standardized terminology, uniformity of survey and audit operations, and emphasizes process assessments rather than a program based solely on compliance. The program establishes minimum training requirements, adopts an auditor certification methodology, and includes survey and audit metrics for the audited organizations as well as the auditing organization.

Study of evaluation techniques of software safety and reliability in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Youn, Cheong; Baek, Y. W.; Kim, H. C.; Park, N. J.; Shin, C. Y. [Chungnam National Univ., Taejon (Korea, Republic of)

1999-04-15

Software system development process and software quality assurance activities are examined in this study. Especially software safety and reliability requirements in nuclear power plant are investigated. For this purpose methodologies and tools which can be applied to software analysis, design, implementation, testing, maintenance step are evaluated. Necessary tasks for each step are investigated. Duty, input, and detailed activity for each task are defined to establish development process of high quality software system. This means applying basic concepts of software engineering and principles of system development. This study establish a guideline that can assure software safety and reliability requirements in digitalized nuclear plant systems and can be used as a guidebook of software development process to assure software quality many software development organization.

Cognitive human reliability analysis for an assessment of the safety significance of complex transients

International Nuclear Information System (INIS)

Amico, P.J.; Hsu, C.J.; Youngblood, R.W.; Fitzpatrick, R.G.

1989-01-01

This paper reports that as part of a probabilistic assessment of the safety significance of complex transients at certain PWR power plants, it was necessary to perform a cognitive human reliability analysis. To increase the confidence in the results, it was desirable to make use of actual observations of operator response which were available for the assessment. An approach was developed which incorporated these observations into the human cognitive reliability (HCR) modeling approach. The results obtained provided additional insights over what would have been found using other approaches. These insights were supported by the observations, and it is suggested that this approach be considered for use in future probabilistic safety assessments

Reliability Analysis of Safety Grade PLC(POSAFE-Q) for Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, J. Y.; Lyou, J.; Lee, D. Y.; Choi, J. G.; Park, W. M.

2006-01-01

The Part Count Method of the military standard MILHDK- 217F has been used for the reliability prediction of the nuclear field. This handbook determines the Programmable Logic Controller (PLC) failure rate by summing the failure rates of the individual component included in the PLC. Normally it is easily predictable that the components added for the fault detection improve the reliability of the PLC. But the application of this handbook is estimated with poor reliability because of the increased component number for the fault detection. To compensate this discrepancy, the quantitative reliability analysis method is suggested using the functional separation model in this paper. And it is applied to the Reactor Protection System (RPS) being developed in Korea to identify any design weak points from a safety point of view

Reliability analysis of idealized tunnel support system using probability-based methods with case studies

Science.gov (United States)

Gharouni-Nik, Morteza; Naeimi, Meysam; Ahadi, Sodayf; Alimoradi, Zahra

2014-06-01

In order to determine the overall safety of a tunnel support lining, a reliability-based approach is presented in this paper. Support elements in jointed rock tunnels are provided to control the ground movement caused by stress redistribution during the tunnel drive. Main support elements contribute to stability of the tunnel structure are recognized owing to identify various aspects of reliability and sustainability in the system. The selection of efficient support methods for rock tunneling is a key factor in order to reduce the number of problems during construction and maintain the project cost and time within the limited budget and planned schedule. This paper introduces a smart approach by which decision-makers will be able to find the overall reliability of tunnel support system before selecting the final scheme of the lining system. Due to this research focus, engineering reliability which is a branch of statistics and probability is being appropriately applied to the field and much effort has been made to use it in tunneling while investigating the reliability of the lining support system for the tunnel structure. Therefore, reliability analysis for evaluating the tunnel support performance is the main idea used in this research. Decomposition approaches are used for producing system block diagram and determining the failure probability of the whole system. Effectiveness of the proposed reliability model of tunnel lining together with the recommended approaches is examined using several case studies and the final value of reliability obtained for different designing scenarios. Considering the idea of linear correlation between safety factors and reliability parameters, the values of isolated reliabilities determined for different structural components of tunnel support system. In order to determine individual safety factors, finite element modeling is employed for different structural subsystems and the results of numerical analyses are obtained in

Reliability data banks

International Nuclear Information System (INIS)

Cannon, A.G.; Bendell, A.

1991-01-01

Following an introductory chapter on Reliability, what is it, why it is needed, how it is achieved and measured, the principles of reliability data bases and analysis methodologies are the subject of the next two chapters. Achievements due to the development of data banks are mentioned for different industries in the next chapter, FACTS, a comprehensive information system for industrial safety and reliability data collection in process plants are covered next. CREDO, the Central Reliability Data Organization is described in the next chapter and is indexed separately, as is the chapter on DANTE, the fabrication reliability Data analysis system. Reliability data banks at Electricite de France and IAEA's experience in compiling a generic component reliability data base are also separately indexed. The European reliability data system, ERDS, and the development of a large data bank come next. The last three chapters look at 'Reliability data banks, - friend foe or a waste of time'? and future developments. (UK)

Mobile phone radiation health risk controversy: the reliability and sufficiency of science behind the safety standards.

Science.gov (United States)

Leszczynski, Dariusz; Xu, Zhengping

2010-01-27

There is ongoing discussion whether the mobile phone radiation causes any health effects. The International Commission on Non-Ionizing Radiation Protection, the International Committee on Electromagnetic Safety and the World Health Organization are assuring that there is no proven health risk and that the present safety limits protect all mobile phone users. However, based on the available scientific evidence, the situation is not as clear. The majority of the evidence comes from in vitro laboratory studies and is of very limited use for determining health risk. Animal toxicology studies are inadequate because it is not possible to "overdose" microwave radiation, as it is done with chemical agents, due to simultaneous induction of heating side-effects. There is a lack of human volunteer studies that would, in unbiased way, demonstrate whether human body responds at all to mobile phone radiation. Finally, the epidemiological evidence is insufficient due to, among others, selection and misclassification bias and the low sensitivity of this approach in detection of health risk within the population. This indicates that the presently available scientific evidence is insufficient to prove reliability of the current safety standards. Therefore, we recommend to use precaution when dealing with mobile phones and, whenever possible and feasible, to limit body exposure to this radiation. Continuation of the research on mobile phone radiation effects is needed in order to improve the basis and the reliability of the safety standards.

Mobile phone radiation health risk controversy: the reliability and sufficiency of science behind the safety standards

Directory of Open Access Journals (Sweden)

Leszczynski Dariusz

2010-01-01

Full Text Available Abstract There is ongoing discussion whether the mobile phone radiation causes any health effects. The International Commission on Non-Ionizing Radiation Protection, the International Committee on Electromagnetic Safety and the World Health Organization are assuring that there is no proven health risk and that the present safety limits protect all mobile phone users. However, based on the available scientific evidence, the situation is not as clear. The majority of the evidence comes from in vitro laboratory studies and is of very limited use for determining health risk. Animal toxicology studies are inadequate because it is not possible to "overdose" microwave radiation, as it is done with chemical agents, due to simultaneous induction of heating side-effects. There is a lack of human volunteer studies that would, in unbiased way, demonstrate whether human body responds at all to mobile phone radiation. Finally, the epidemiological evidence is insufficient due to, among others, selection and misclassification bias and the low sensitivity of this approach in detection of health risk within the population. This indicates that the presently available scientific evidence is insufficient to prove reliability of the current safety standards. Therefore, we recommend to use precaution when dealing with mobile phones and, whenever possible and feasible, to limit body exposure to this radiation. Continuation of the research on mobile phone radiation effects is needed in order to improve the basis and the reliability of the safety standards.

Approach to design of future FBRs with enhanced safety and economy

International Nuclear Information System (INIS)

Raghupathy, S.; Balasubramaniyan, V.; Puthiyavinayagam, P.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C.; Raj, Baldev

2009-01-01

After the techno economic demonstration of FBR technology through PFBR, it is essential to achieve high economic competitiveness with enhanced safety on par with other power generation options. Towards this, design studies were made with the objectives of identifying means to achieve enhanced safety, design features for improving economy, conceptualise the plant layout, ways to reduce construction time and to achieve higher capacity factor. This paper describes the outcome of the above studies and the approach to the design of future FBRs with enhanced safety and economy. High economic competitiveness and public acceptance are essential for commercial deployment of fast reactors. India has a well-laid out three phase nuclear power programme with Pressurised Heavy Water Reactors (PHWRs) utilizing natural uranium as fuel in the first phase and fast breeder reactors (FBRs) using plutonium and thorium in the second and third phases. Deploying FBRs on a commercial scale is vital for India in order to utilize the vast thorium reserves and to meet the long term energy needs of the country. As a first step, a 40 MWt, 13.5 MWe Fast Breeder Test Reactor (FBTR) was constructed and has been in operation since 1985. Operation of FBTR has given valuable operating experience with critical systems of FBRs including sodium systems and has given confidence to undertake the design of 500 MWe Prototype Fast Breeder Reactor (PFBR). Construction of PFBR is currently under progress and is expected to be completed by 2010. It is planned to construct four more oxide fuelled FBRs of 500 MWe capacity by 2020 and metal fuelled FBRs of 1000 MWe thereafter. Enhanced safety is proposed to be achieved through improvements in shut down systems, adopting in-vessel primary sodium purification concept, improving the reliability of decay heat removal system, shifting component decontamination activities to a separate building located outside the reactor containment building, providing additional

Architecture for interlock systems: reliability analysis with regard to safety and availability

International Nuclear Information System (INIS)

Wagner, S.; Apollonio, A.; Schmidt, R.; Zerlauth, M.; Vergara-Fernandez, A.

2012-01-01

For particle accelerators like LHC and other large experimental physics facilities like ITER, the machine protection relies on complex interlock systems. In the design of interlock loops for the signal exchange in machine protection systems, the choice of the hardware architecture impacts on machine safety and availability. The reliable performance of a machine stop (leaving the machine in a safe state) in case of an emergency, is an inherent requirement. The constraints in terms of machine availability on the other hand may differ from one facility to another. Spurious machine stops, lowering machine availability, may to a certain extent be tolerated in facilities where they do not cause undue equipment wear-out. In order to compare various interlock loop architectures in terms of safety and availability, the occurrence frequencies of related scenarios have been calculated in a reliability analysis, using a generic analytical model. This paper presents the results and illustrates the potential of the analysis method for supporting the choice of interlock system architectures. The results show the advantages of a 2003 (3 redundant lines with 2-out-of-3 voting) over the 6 architectures under consideration for systems with high requirements in both safety and availability

Reliability assurance for regulation of advanced reactors

International Nuclear Information System (INIS)

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

1992-01-01

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

Reliability assurance for regulation of advanced reactors

International Nuclear Information System (INIS)

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

1991-01-01

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

A holistic framework of degradation modeling for reliability analysis and maintenance optimization of nuclear safety systems

International Nuclear Information System (INIS)

Lin, Yanhui

2016-01-01

Components of nuclear safety systems are in general highly reliable, which leads to a difficulty in modeling their degradation and failure behaviors due to the limited amount of data available. Besides, the complexity of such modeling task is increased by the fact that these systems are often subject to multiple competing degradation processes and that these can be dependent under certain circumstances, and influenced by a number of external factors (e.g. temperature, stress, mechanical shocks, etc.). In this complicated problem setting, this PhD work aims to develop a holistic framework of models and computational methods for the reliability-based analysis and maintenance optimization of nuclear safety systems taking into account the available knowledge on the systems, degradation and failure behaviors, their dependencies, the external influencing factors and the associated uncertainties.The original scientific contributions of the work are: (1) For single components, we integrate random shocks into multi-state physics models for component reliability analysis, considering general dependencies between the degradation and two types of random shocks. (2) For multi-component systems (with a limited number of components):(a) a piecewise-deterministic Markov process modeling framework is developed to treat degradation dependency in a system whose degradation processes are modeled by physics-based models and multi-state models; (b) epistemic uncertainty due to incomplete or imprecise knowledge is considered and a finite-volume scheme is extended to assess the (fuzzy) system reliability; (c) the mean absolute deviation importance measures are extended for components with multiple dependent competing degradation processes and subject to maintenance; (d) the optimal maintenance policy considering epistemic uncertainty and degradation dependency is derived by combining finite-volume scheme, differential evolution and non-dominated sorting differential evolution; (e) the

V and V based Fault Estimation Method for Safety-Critical Software using BNs

International Nuclear Information System (INIS)

Eom, Heung Seop; Park, Gee Yong; Jang, Seung Cheol; Kang, Hyun Gook

2011-01-01

Quantitative software reliability measurement approaches have severe limitations in demonstrating the proper level of reliability for safety-critical software. These limitations can be overcome by using some other means of assessment. One of the promising candidates is based on the quality of the software development. Particularly in the nuclear industry, regulatory bodies in most countries do not accept the concept of quantitative goals as a sole means of meeting their regulations for the reliability of digital computers in NPPs, and use deterministic criteria for both hardware and software. The point of deterministic criteria is to assess the whole development process and its related activities during the software development life cycle for the acceptance of safety-critical software, and software V and V plays an important role in this process. In this light, we studied a V and V based fault estimation method using Bayesian Nets (BNs) to assess the reliability of safety-critical software, especially reactor protection system software in a NPP. The BNs in the study were made for an estimation of software faults and were based on the V and V frame, which governs the development of safety-critical software in the nuclear field. A case study was carried out for a reactor protection system that was developed as a part of the Korea Nuclear Instrumentation and Control System. The insight from the case study is that some important factors affecting the fault number of the target software include the residual faults in the system specification, maximum number of faults introduced in the development phase, ratio between process/function characteristic, uncertainty sizing, and fault elimination rate by inspection activities

Inverse Reliability Task: Artificial Neural Networks and Reliability-Based Optimization Approaches

OpenAIRE

Lehký , David; Slowik , Ondřej; Novák , Drahomír

2014-01-01

Part 7: Genetic Algorithms; International audience; The paper presents two alternative approaches to solve inverse reliability task – to determine the design parameters to achieve desired target reliabilities. The first approach is based on utilization of artificial neural networks and small-sample simulation Latin hypercube sampling. The second approach considers inverse reliability task as reliability-based optimization task using double-loop method and also small-sample simulation. Efficie...

Practice of value-based distribution reliability assessment (VBDRA) at Scarborough Public Utilities

International Nuclear Information System (INIS)

Chen, R-L.

1995-01-01

The development of value-based distribution reliability assessment (VBDRA) at Scarborough Public Utilities was described. Load point reliability indices, customer interruption costs (CIC), continuity and service reliability, accuracy of CIC, and the aspects of application of VBDRA were addressed. The application of VBDRA to a long-term rebuild plan for 4.16 kV distribution system was described. The importance of a cost-benefit analysis for implementation of VBDRA was emphasized. In the case of the Scarborough Public Utilities Commission the enhanced feeder reliability assessment was found to influence the allocation of funding to where it provided the most value to customers. 14 refs., 3 tabs., 3 figs

Use of a Web Site to Enhance Criticality Safety Training

International Nuclear Information System (INIS)

Huang, S T; Morman, J

2003-01-01

Currently, a website dedicated to enhancing communication and dissemination of criticality safety information is sponsored by the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This website was developed as part of the DOE response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. The website is the focal point for DOE nuclear criticality safety (NCS) activities, resources and references, including hyperlinks to other sites actively involved in the collection and dissemination of criticality safety information. The website is maintained by the Lawrence Livermore National Laboratory (LLNL) under auspices of the NCSP management. One area of the website contains a series of Nuclear Criticality Safety Engineer Training (NCSET) modules. During the past few years, many users worldwide have accessed the NCSET section of the NCSP website and have downloaded the training modules as an aid for their training programs. This trend was remarkable in that it points out a continuing need of the criticality safety community across the globe. It has long been recognized that training of criticality safety professionals is a continuing process involving both knowledge-based training and experience-based operations floor training. As more of the experienced criticality safety professionals reach retirement age, the opportunities for mentoring programs are reduced. It is essential that some method be provided to assist the training of young criticality safety professionals to replenish this limited human expert resource to support on-going and future nuclear operations. The main objective of this paper is to present the features of the NCSP website, including its mission, contents, and most importantly its use for the dissemination of training modules to the criticality safety community. We will discuss lessons learned and several ideas

Nuclear Safety R and D Programs and trend in the U. S. Utility Industry

International Nuclear Information System (INIS)

Kim, Jong Hyun

1992-01-01

First of all, the deterministic approach to safety analysis, which had dominated safety research in the earlier years, has given much ground to probabilistic approach. Secondly, human factors analysis has become an important part of safety research. Third, safety research relevant to reliability, or safety combined with reliability, are gradually taking place of purely safety-oriented or stand-alone safety research. More and more nuclear utilities in the U. S. are integrating safety with reliability. This evolution is in part due to the successful completion of major safety testing and analyses of deterministic nature, and partially due to the utility industry's desire to harvest synergistic nature, and partially due to the utility industry's desire to harvest synergistic results by combining safety with reliability, as the utility industry is more and more concerned about reducing operation and maintenance costs by enhancing reliability while maintaining plant safety. Nuclear safety is a complex and comprehensive concept, defying a simple categorization or interpretation. Thus, research and development in nuclear safety is necessarily diverse, and the program areas and trend presented in this paper are not meant to be all inclusive. For instance, there are some other active areas that were not mentioned, such as seismic risk assessment program and others. Nuclear safety research and development activities have undergone a perceptible shift of emphasis in recent years. They have become more focused and product-oriented. Also, except for the severe accident analysis, the emphasis on prevention and mitigation of accident, rather than analyzing the consequences of accident, is very much in evidence; that is, reliability-based technologies using PIRA methodology, and upgrading of instrumentation and control technologies are in the main stream of activities

Demonstrating the Safety and Reliability of a New System or Spacecraft: Incorporating Analyses and Reviews of the Design and Processing in Determining the Number of Tests to be Conducted

Science.gov (United States)

Vesely, William E.; Colon, Alfredo E.

2010-01-01

Design Safety/Reliability is associated with the probability of no failure-causing faults existing in a design. Confidence in the non-existence of failure-causing faults is increased by performing tests with no failure. Reliability-Growth testing requirements are based on initial assurance and fault detection probability. Using binomial tables generally gives too many required tests compared to reliability-growth requirements. Reliability-Growth testing requirements are based on reliability principles and factors and should be used.

Food safety security: a new concept for enhancing food safety measures.

Science.gov (United States)

Iyengar, Venkatesh; Elmadfa, Ibrahim

2012-06-01

The food safety security (FSS) concept is perceived as an early warning system for minimizing food safety (FS) breaches, and it functions in conjunction with existing FS measures. Essentially, the function of FS and FSS measures can be visualized in two parts: (i) the FS preventive measures as actions taken at the stem level, and (ii) the FSS interventions as actions taken at the root level, to enhance the impact of the implemented safety steps. In practice, along with FS, FSS also draws its support from (i) legislative directives and regulatory measures for enforcing verifiable, timely, and effective compliance; (ii) measurement systems in place for sustained quality assurance; and (iii) shared responsibility to ensure cohesion among all the stakeholders namely, policy makers, regulators, food producers, processors and distributors, and consumers. However, the functional framework of FSS differs from that of FS by way of: (i) retooling the vulnerable segments of the preventive features of existing FS measures; (ii) fine-tuning response systems to efficiently preempt the FS breaches; (iii) building a long-term nutrient and toxicant surveillance network based on validated measurement systems functioning in real time; (iv) focusing on crisp, clear, and correct communication that resonates among all the stakeholders; and (v) developing inter-disciplinary human resources to meet ever-increasing FS challenges. Important determinants of FSS include: (i) strengthening international dialogue for refining regulatory reforms and addressing emerging risks; (ii) developing innovative and strategic action points for intervention {in addition to Hazard Analysis and Critical Control Points (HACCP) procedures]; and (iii) introducing additional science-based tools such as metrology-based measurement systems.

Enhancements in SCALE 6.1

International Nuclear Information System (INIS)

Rearden, Bradley T.

2010-01-01

The Standardized Computer Analysis for Licensing Evaluation (SCALE) code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides a 'plug-and-play' framework with nearly 80 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE's graphical user interfaces assist with accurate system modeling and convenient access to desired results. SCALE 6.1, scheduled for release in the fall of 2010, provides improved reliability and introduces a number of enhanced features, some of which are briefly described here. SCALE 6.1 provides state-of-the-art capabilities for criticality safety, reactor physics, and radiation shielding in a robust yet user-friendly package. The new features and improved reliability of this latest release of SCALE are intended to improve safety and efficiency throughout the nuclear community.

Aptamer-Based Analysis: A Promising Alternative for Food Safety Control

Directory of Open Access Journals (Sweden)

Sonia Amaya-González

2013-11-01

Full Text Available Ensuring food safety is nowadays a top priority of authorities and professional players in the food supply chain. One of the key challenges to determine the safety of food and guarantee a high level of consumer protection is the availability of fast, sensitive and reliable analytical methods to identify specific hazards associated to food before they become a health problem. The limitations of existing methods have encouraged the development of new technologies, among them biosensors. Success in biosensor design depends largely on the development of novel receptors with enhanced affinity to the target, while being stable and economical. Aptamers fulfill these characteristics, and thus have surfaced as promising alternatives to natural receptors. This Review describes analytical strategies developed so far using aptamers for the control of pathogens, allergens, adulterants, toxins and other forbidden contaminants to ensure food safety. The main progresses to date are presented, highlighting potential prospects for the future.

Nuclear safety policy statement in korea

International Nuclear Information System (INIS)

Kim, W.S.; Kim, H.J.; Choi, K.S.; Choi, Y.S.; Park, D.K.

2006-01-01

Full text: Wide varieties of programs to enhance nuclear safety have been established and implemented by the Korean government in accordance with the Nuclear Safety Policy Statement announced in September 1994. The policy statement was intended to set the long-term policy goals for maintaining and achieving high-level of nuclear safety and also help the public understand the national policy and a strong will of the government toward nuclear safety. It has been recognized as very effective in developing safety culture in nuclear-related organizations and also enhancing nuclear safety in Korea. However, ageing of operating nuclear power plants and increasing of new nuclear facilities have demanded a new comprehensive national safety policy to cover the coming decade, taking the implementation results of the policy statement of 1994 and the changing environment of nuclear industries into consideration. Therefore, the results of safety policy implementation have been reviewed and, considering changing environment and future prospects, a new nuclear safety policy statement as a highest level national policy has been developed. The implementation results of 11 regulatory policy directions such as the use of Probabilistic Safety Assessment, introduction of Periodic Safety Review, strengthening of safety research, introduction of Risk Based Regulation stipulated in the safety policy statement of 1994 were reviewed and measures taken after various symposia on nuclear safety held in Nuclear Safety Days since 1995 were evaluated. The changing international and domestic environment of nuclear industry were analysed and future prospects were explored. Based on the analysis and review results, a draft of new nuclear safety policy statement was developed. The draft was finalized after the review of many prominent experts in Korea. Considering changing environment and future prospects, new policy statement that will show government's persistent will for nuclear safety has been

Human reliability

International Nuclear Information System (INIS)

Embrey, D.E.

1987-01-01

Concepts and techniques of human reliability have been developed and are used mostly in probabilistic risk assessment. For this, the major application of human reliability assessment has been to identify the human errors which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. Some of the major issues within human reliability studies are reviewed and it is shown how these are applied to the assessment of human failures in systems. This is done under the following headings; models of human performance used in human reliability assessment, the nature of human error, classification of errors in man-machine systems, practical aspects, human reliability modelling in complex situations, quantification and examination of human reliability, judgement based approaches, holistic techniques and decision analytic approaches. (UK)

Standardization of domestic human reliability analysis and experience of human reliability analysis in probabilistic safety assessment for NPPs under design

International Nuclear Information System (INIS)

Kang, D. I.; Jung, W. D.

2002-01-01

This paper introduces the background and development activities of domestic standardization of procedure and method for Human Reliability Analysis (HRA) to avoid the intervention of subjectivity by HRA analyst in Probabilistic Safety Assessment (PSA) as possible, and the review of the HRA results for domestic nuclear power plants under design studied by Korea Atomic Energy Research Institute. We identify the HRA methods used for PSA for domestic NPPs and discuss the subjectivity of HRA analyst shown in performing a HRA. Also, we introduce the PSA guidelines published in USA and review the HRA results based on them. We propose the system of a standard procedure and method for HRA to be developed

Identification of Black Spots Based on Reliability Approach

Directory of Open Access Journals (Sweden)

Ahmadreza Ghaffari

2013-12-01

Full Text Available Identifying crash “black-spots”, “hot-spots” or “high-risk” locations is one of the most important and prevalent concerns in traffic safety and various methods have been devised and presented for solving this issue until now. In this paper, a new method based on the reliability analysis is presented to identify black-spots. Reliability analysis has an ordered framework to consider the probabilistic nature of engineering problems, so crashes with their probabilistic na -ture can be applied. In this study, the application of this new method was compared with the commonly implemented Frequency and Empirical Bayesian methods using simulated data. The results indicated that the traditional methods can lead to an inconsistent prediction due to their inconsider -ation of the variance of the number of crashes in each site and their dependence on the mean of the data.

Analyzing Reliability and Performance Trade-Offs of HLS-Based Designs in SRAM-Based FPGAs Under Soft Errors

Science.gov (United States)

Tambara, Lucas Antunes; Tonfat, Jorge; Santos, André; Kastensmidt, Fernanda Lima; Medina, Nilberto H.; Added, Nemitala; Aguiar, Vitor A. P.; Aguirre, Fernando; Silveira, Marcilei A. G.

2017-02-01

The increasing system complexity of FPGA-based hardware designs and shortening of time-to-market have motivated the adoption of new designing methodologies focused on addressing the current need for high-performance circuits. High-Level Synthesis (HLS) tools can generate Register Transfer Level (RTL) designs from high-level software programming languages. These tools have evolved significantly in recent years, providing optimized RTL designs, which can serve the needs of safety-critical applications that require both high performance and high reliability levels. However, a reliability evaluation of HLS-based designs under soft errors has not yet been presented. In this work, the trade-offs of different HLS-based designs in terms of reliability, resource utilization, and performance are investigated by analyzing their behavior under soft errors and comparing them to a standard processor-based implementation in an SRAM-based FPGA. Results obtained from fault injection campaigns and radiation experiments show that it is possible to increase the performance of a processor-based system up to 5,000 times by changing its architecture with a small impact in the cross section (increasing up to 8 times), and still increasing the Mean Workload Between Failures (MWBF) of the system.

Reliability analysis of idealized tunnel support system using probability-based methods with case studies

NARCIS (Netherlands)

Gharouni-Nik, M.; Naeimi, M.; Ahadi, S.; Alimoradi, Z.

2014-01-01

In order to determine the overall safety of a tunnel support lining, a reliability-based approach is presented in this paper. Support elements in jointed rock tunnels are provided to control the ground movement caused by stress redistribution during the tunnel drive. Main support elements contribute

Study on reliability analysis based on multilevel flow models and fault tree method

International Nuclear Information System (INIS)

Chen Qiang; Yang Ming

2014-01-01

Multilevel flow models (MFM) and fault tree method describe the system knowledge in different forms, so the two methods express an equivalent logic of the system reliability under the same boundary conditions and assumptions. Based on this and combined with the characteristics of MFM, a method mapping MFM to fault tree was put forward, thus providing a way to establish fault tree rapidly and realizing qualitative reliability analysis based on MFM. Taking the safety injection system of pressurized water reactor nuclear power plant as an example, its MFM was established and its reliability was analyzed qualitatively. The analysis result shows that the logic of mapping MFM to fault tree is correct. The MFM is easily understood, created and modified. Compared with the traditional fault tree analysis, the workload is greatly reduced and the modeling time is saved. (authors)

Experience gained in enhancing operational safety at ComEd's nuclear power plants

International Nuclear Information System (INIS)

Elias, D.

1997-01-01

The following aspects of experience gained in enhancing operational safety at Comed's nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight

Use of reliability in the LMFBR industry

International Nuclear Information System (INIS)

Penland, J.R.; Smith, A.M.; Goeser, D.K.

1977-01-01

This mission of a Reliability Program for an LMFBR should be to enhance the design and operational characteristics relative to safety and to plant availability. Successful accomplishment of this mission requires proper integration of several reliability engineering tasks--analysis, testing, parts controls and program controls. Such integration requires, in turn, that the program be structured, planned and managed. This paper describes the technical integration necessary and the management activities required to achieve mission success for LMFBR's

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

Accuracy and Metrological Reliability Enhancing of Thermoelectric Transducers

Directory of Open Access Journals (Sweden)

Bogdan Stadnyk

2010-12-01

Full Text Available This article is devoted to development and use of thermoelectric thermotransducers with an enhanced accuracy and metrological reliability. The actuality of a problem is stipulated. Investigating changes at typical external environments, the mechanisms of transformation function instability are considered; possibilities of thermodynamic presentation use are analyzed concerning a thermometric substance. The algorithm of thermotransducer instrumental errors’ minimization is developed.

A Step Toward High Reliability: Implementation of a Daily Safety Brief in a Children's Hospital.

Science.gov (United States)

Saysana, Michele; McCaskey, Marjorie; Cox, Elaine; Thompson, Rachel; Tuttle, Lora K; Haut, Paul R

2017-09-01

Health care is a high-risk industry. To improve communication about daily events and begin the journey toward a high reliability organization, the Riley Hospital for Children at Indiana University Health implemented a daily safety brief. Various departments in our children's hospital were asked to participate in a daily safety brief, reporting daily events and unexpected outcomes within their scope of responsibility. Participants were surveyed before and after implementation of the safety brief about communication and awareness of events in the hospital. The length of the brief and percentage of departments reporting unexpected outcomes were measured. The analysis of the presurvey and the postsurvey showed a statistically significant improvement in the questions related to the awareness of daily events as well as communication and relationships between departments. The monthly mean length of time for the brief was 15 minutes or less. Unexpected outcomes were reported by 50% of the departments for 8 months. A daily safety brief can be successfully implemented in a children's hospital. Communication between departments and awareness of daily events were improved. Implementation of a daily safety brief is a step toward becoming a high reliability organization.

Enhanced Reliability of Photovoltaic Systems with Energy Storage and Controls

Energy Technology Data Exchange (ETDEWEB)

Manz, D.; Schelenz, O.; Chandra, R.; Bose, S.; de Rooij, M.; Bebic, J.

2008-02-01

This report summarizes efforts to reconfigure loads during outages to allow individual customers the opportunity to enhance the reliability of their electric service through the management of their loads, photovoltaics, and energy storage devices.

Safety culture in nuclear installations. Guidance for the use in enhancement of safety culture

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-12-01

This guidance has been developed for use in the IAEA Safety Culture Services, which provides support to Member States in their efforts to develop a sound safety culture of their organizations. It will be of particular use in seminars and training workshops that are part of these services. Much of the information in this publication reflects the approach the IAEA has adopted to assist nuclear organizations in Member States in improving their safety culture. This guidance covers topics such as: what is culture, and in particular what is safety culture; what are the stages of development of safety culture, and how you can assess its development using employee surveys; what practices can be used to develop safety culture, and what indicators will help monitor progress. The symptoms of a weakening safety culture are described, as well as the lessons learned from organizations who have experienced safety culture problems. This guide also contains information on how to undertake the process of transforming the existing safety culture, and develop a learning culture in an organization that is based on continuous improvement. The relationship between quality and safety is discussed. The safety culture services offered by the IAEA are also described. The IAEA perspective of safety culture has expanded with time as its understanding of the complexities of the concept developed. The concept of safety culture was first introduced by the International Nuclear Safety Advisory Group formed by the IAEA. In their report (INSAG-4, 1991) they maintained that the establishment of a safety culture within an organization is one of the fundamental management principles necessary for the safe operation of a nuclear facility. The definition recognized that safety culture is both structural and attitudinal in nature and relates to the organization and its style, as well as to attitudes, approaches and the commitment of individuals at all levels in the organization. In the framework of the

Safety culture in nuclear installations. Guidance for the use in enhancement of safety culture

International Nuclear Information System (INIS)

2002-12-01

This guidance has been developed for use in the IAEA Safety Culture Services, which provides support to Member States in their efforts to develop a sound safety culture of their organizations. It will be of particular use in seminars and training workshops that are part of these services. Much of the information in this publication reflects the approach the IAEA has adopted to assist nuclear organizations in Member States in improving their safety culture. This guidance covers topics such as: what is culture, and in particular what is safety culture; what are the stages of development of safety culture, and how you can assess its development using employee surveys; what practices can be used to develop safety culture, and what indicators will help monitor progress. The symptoms of a weakening safety culture are described, as well as the lessons learned from organizations who have experienced safety culture problems. This guide also contains information on how to undertake the process of transforming the existing safety culture, and develop a learning culture in an organization that is based on continuous improvement. The relationship between quality and safety is discussed. The safety culture services offered by the IAEA are also described. The IAEA perspective of safety culture has expanded with time as its understanding of the complexities of the concept developed. The concept of safety culture was first introduced by the International Nuclear Safety Advisory Group formed by the IAEA. In their report (INSAG-4, 1991) they maintained that the establishment of a safety culture within an organization is one of the fundamental management principles necessary for the safe operation of a nuclear facility. The definition recognized that safety culture is both structural and attitudinal in nature and relates to the organization and its style, as well as to attitudes, approaches and the commitment of individuals at all levels in the organization. In the framework of the

A novel mechanical design of broken rope protection device for enhancing the safety performances of overhead manned equipment in coal mine

Directory of Open Access Journals (Sweden)

Xiaoguang Zhang

2015-08-01

Full Text Available A novel mechanical design of the broken rope protection device is proposed to enhance the safety performances of the overhead manned equipment. According to the operating characteristics and functional requirements of the overhead manned equipment, a three-dimensional mechanical model of the broken rope protection device was redesigned. Based on the known parameters of the mechanical model, the stress and strength of the main components are readjusted using the statics characteristics of finite element analysis. To ensure the reliability of the control system of the broken rope protection device, the process of people’s falling, the response performance of the tension sensor, and the signal extraction of the broken rope are analyzed under different loading and unloading speeds. The working principle of the broken rope protection device is expounded in detail. The experimental results showed that better effect is obtained by the new broken rope protection device, which is characterized by good durability, low investment, and high reliability.

Reliability data book

International Nuclear Information System (INIS)

Bento, J.P.; Boerje, S.; Ericsson, G.; Hasler, A.; Lyden, C.O.; Wallin, L.; Poern, K.; Aakerlund, O.

1985-01-01

The main objective for the report is to improve failure data for reliability calculations as parts of safety analyses for Swedish nuclear power plants. The work is based primarily on evaluations of failure reports as well as information provided by the operation and maintenance staff of each plant. In the report are presented charts of reliability data for: pumps, valves, control rods/rod drives, electrical components, and instruments. (L.E.)

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)

Knowledge management and networking for enhancing nuclear safety

International Nuclear Information System (INIS)

Taniguchi, T.; Lederman, L.

2004-01-01

Striving for innovative solutions to enhance efficiency of programme delivery and a wider outreach of its nuclear safety activities, the International Atomic Energy Agency (IAEA) has developed an Integrated Safety Approach as a platform for linking its safety related statutory functions and its many associated activities. The approach recognizes the vital importance of effective management of the knowledge base and builds on the integration between the IAEA's safety standards and all aspects of the provision for their application, including peer reviews and technical meetings to share lessons learned. The IAEA is using knowledge management techniques to develop process flows, map safety knowledge and to promote knowledge sharing. The first practical application was the establishment of a knowledge base related to safety aspects of ageing and long-term operation of nuclear power plants. The IAEA is also promoting and facilitating the establishment of regional nuclear and radiation safety networks to preserve existing knowledge and expertise as well as to strengthen sharing and creation of new knowledge in these fields. Prominent examples are the Asian Nuclear Safety Network established in the frame of the IAEA's Programme on the Safety of Nuclear Installations in South East Asia, Pacific and Far East Countries, and the Ibero-American Radiation Safety Network in the frame of the Ibero-American Forum of Nuclear Regulators. Results to date are most encouraging and suggest that this pioneer work should be extended to other regions and eventually to a global nuclear safety network. Responsive to the need of Member States, the IAEA Secretariat has prepared and made available a large number of up-to-date training packages in nuclear, radiation, transport and waste safety, using IAEA safety standards as a basis. It is also providing instruction to trainers in Member States on the use of these modules. This ensures that the material is properly used and that the IAEA

Reliability and safety of a new upper cervical spine injury treatment algorithm

Directory of Open Access Journals (Sweden)

Andrei Fernandes Joaquim

Full Text Available ABSTRACT In the present study, we evaluated the reliability and safety of a new upper cervical spine injury treatment algorithm to help in the selection of the best treatment modality for these injuries. Methods Thirty cases, previously treated according to the new algorithm, were presented to four spine surgeons who were questioned about their personal suggestion for treatment, and the treatment suggested according to the application of the algorithm. After four weeks, the same questions were asked again to evaluate reliability (intra- and inter-observer using the Kappa index. Results The reliability of the treatment suggested by applying the algorithm was superior to the reliability of the surgeons’ personal suggestion for treatment. When applying the upper cervical spine injury treatment algorithm, an agreement with the treatment actually performed was obtained in more than 89% of the cases. Conclusion The system is safe and reliable for treating traumatic upper cervical spine injuries. The algorithm can be used to help surgeons in the decision between conservative versus surgical treatment of these injuries.

Probabilistic optimization of safety coefficients

International Nuclear Information System (INIS)

Marques, M.; Devictor, N.; Magistris, F. de

1999-01-01

This article describes a reliability-based method for the optimization of safety coefficients defined and used in design codes. The purpose of the optimization is to determine the partial safety coefficients which minimize an objective function for sets of components and loading situations covered by a design rule. This objective function is a sum of distances between the reliability of the components designed using the safety coefficients and a target reliability. The advantage of this method is shown on the examples of the reactor vessel, a vapour pipe and the safety injection circuit. (authors)

First evidence on the validity and reliability of the Safety Organizing Scale-Nursing Home version (SOS-NH).

Science.gov (United States)

Ausserhofer, Dietmar; Anderson, Ruth A; Colón-Emeric, Cathleen; Schwendimann, René

2013-08-01

The Safety Organizing Scale is a valid and reliable measure on safety behaviors and practices in hospitals. This study aimed to explore the psychometric properties of the Safety Organizing Scale-Nursing Home version (SOS-NH). In a cross-sectional analysis of staff survey data, we examined validity and reliability of the 9-item Safety SOS-NH using American Educational Research Association guidelines. This substudy of a larger trial used baseline survey data collected from staff members (n = 627) in a variety of work roles in 13 nursing homes (NHs) in North Carolina and Virginia. Psychometric evaluation of the SOS-NH revealed good response patterns with low average of missing values across all items (3.05%). Analyses of the SOS-NH's internal structure (eg, comparative fit indices = 0.929, standardized root mean square error of approximation = 0.045) and consistency (composite reliability = 0.94) suggested its 1-dimensionality. Significant between-facility variability, intraclass correlations, within-group agreement, and design effect confirmed appropriateness of the SOS-NH for measurement at the NH level, justifying data aggregation. The SOS-NH showed discriminate validity from one related concept: communication openness. Initial evidence regarding validity and reliability of the SOS-NH supports its utility in measuring safety behaviors and practices among a wide range of NH staff members, including those with low literacy. Further psychometric evaluation should focus on testing concurrent and criterion validity, using resident outcome measures (eg, patient fall rates). Copyright © 2013 American Medical Directors Association, Inc. All rights reserved.

RSAM: An enhanced architecture for achieving web services reliability in mobile cloud computing

Directory of Open Access Journals (Sweden)

Amr S. Abdelfattah

2018-04-01

Full Text Available The evolution of the mobile landscape is coupled with the ubiquitous nature of the internet with its intermittent wireless connectivity and the web services. Achieving the web service reliability results in low communication overhead and retrieving the appropriate response. The middleware approach (MA is highly tended to achieve the web service reliability. This paper proposes a Reliable Service Architecture using Middleware (RSAM that achieves the reliable web services consumption. The enhanced architecture focuses on ensuring and tracking the request execution under the communication limitations and service temporal unavailability. It considers the most measurement factors including: request size, response size, and consuming time. We conducted experiments to compare the enhanced architecture with the traditional one. In these experiments, we covered several cases to prove the achievement of reliability. Results also show that the request size was found to be constant, the response size is identical to the traditional architecture, and the increase in the consuming time was less than 5% of the transaction time with the different response sizes. Keywords: Reliable web service, Middleware architecture, Mobile cloud computing

The contribution of quality assurance to safety and reliability in nuclear power plants

International Nuclear Information System (INIS)

Raisic, N.

1978-01-01

The potential contribution of quality assurance to nuclear power plant safety and reliability is analysed. An attempt is made to establish a relationship between quality and reliability. The reliability may be expressed in quantitative terms as ''the probability that an item will perform a required function for a stated period of time''. Quality, however, cannot be expressed in simple quantitative terms but only as a set of required properties which an item should have for a specific application. The achievement of quality and additional reliability objectives is a task of project activities such as design, construction, installation, operation, etc. The elements of a quality assurance system and its functions in nuclear power projects are presented in some detail. Confidence in plant quality, which should be a basis for the regulatory body issuing the construction permit or operation licence, should be based on the capability of quality assurance activities to prevent errors and correct deficiencies in nuclear power plants. An analysis is made of those errors in plant design, manufacture, construction and operation which contribute most frequently to plant outages. It is concluded that these errors can be avoided or corrected by strict adherence to quality assurance principles and by the efficient functioning of quality assurance systems. In fact, quality assurance may be considered an effective defence against common cause failures originating in errors in the design, manufacture, installation or operation of a nuclear power plant

An Enhanced Preventive Maintenance Optimization Model Based on a Three-Stage Failure Process

Directory of Open Access Journals (Sweden)

Ruifeng Yang

2015-01-01

Full Text Available Nuclear power plants are highly complex systems and the issues related to their safety are of primary importance. Probabilistic safety assessment is regarded as the most widespread methodology for studying the safety of nuclear power plants. As maintenance is one of the most important factors for affecting the reliability and safety, an enhanced preventive maintenance optimization model based on a three-stage failure process is proposed. Preventive maintenance is still a dominant maintenance policy due to its easy implementation. In order to correspond to the three-color scheme commonly used in practice, the lifetime of system before failure is divided into three stages, namely, normal, minor defective, and severe defective stages. When the minor defective stage is identified, two measures are considered for comparison: one is that halving the inspection interval only when the minor defective stage is identified at the first time; the other one is that if only identifying the minor defective stage, the subsequent inspection interval is halved. Maintenance is implemented immediately once the severe defective stage is identified. Minimizing the expected cost per unit time is our objective function to optimize the inspection interval. Finally, a numerical example is presented to illustrate the effectiveness of the proposed models.

A Method of Nuclear Software Reliability Estimation

International Nuclear Information System (INIS)

Park, Gee Yong; Eom, Heung Seop; Cheon, Se Woo; Jang, Seung Cheol

2011-01-01

A method on estimating software reliability for nuclear safety software is proposed. This method is based on the software reliability growth model (SRGM) where the behavior of software failure is assumed to follow the non-homogeneous Poisson process. Several modeling schemes are presented in order to estimate and predict more precisely the number of software defects based on a few of software failure data. The Bayesian statistical inference is employed to estimate the model parameters by incorporating the software test cases into the model. It is identified that this method is capable of accurately estimating the remaining number of software defects which are on-demand type directly affecting safety trip functions. The software reliability can be estimated from a model equation and one method of obtaining the software reliability is proposed

PWR core safety analysis with 3-dimensional methods

International Nuclear Information System (INIS)

Gensler, A.; Kühnel, K.; Kuch, S.

2015-01-01

Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

Reliability Analysis of Sealing Structure of Electromechanical System Based on Kriging Model

Science.gov (United States)

Zhang, F.; Wang, Y. M.; Chen, R. W.; Deng, W. W.; Gao, Y.

2018-05-01

The sealing performance of aircraft electromechanical system has a great influence on flight safety, and the reliability of its typical seal structure is analyzed by researcher. In this paper, we regard reciprocating seal structure as a research object to study structural reliability. Having been based on the finite element numerical simulation method, the contact stress between the rubber sealing ring and the cylinder wall is calculated, and the relationship between the contact stress and the pressure of the hydraulic medium is built, and the friction force on different working conditions are compared. Through the co-simulation, the adaptive Kriging model obtained by EFF learning mechanism is used to describe the failure probability of the seal ring, so as to evaluate the reliability of the sealing structure. This article proposes a new idea of numerical evaluation for the reliability analysis of sealing structure, and also provides a theoretical basis for the optimal design of sealing structure.

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

EC6 safety design improvements

Energy Technology Data Exchange (ETDEWEB)

Yu, S.; Lee, A.G.; Soulard, M. [Candu Energy Inc., Mississauga, ON (Canada)

2014-07-01

The Enhanced CANDU 6 (EC6) builds on the proven high performance design such as the Qinshan CANDU 6 reactor, and has made improvements to safety, operational performance, and has incorporated extensive operational feedback. Completion of all three phases of the pre-licensing design review by the Canadian Regulator - the Canadian Nuclear Safety Commission has provided a higher level of assurance that the EC6 reference design has taken modern regulatory requirements and expectations into account and further confirmed that there are no fundamental barriers to licensing the EC6 design in Canada. The EC6 design is based on the defence-in-depth principles in INSAG-10 and provides further safety features that address the lessons learned from Fukushima. With these safety features, the EC6 design has strengthened accident prevention as the first priority in the defence-in-depth strategy, as outlined in INSAG-10. As well, the EC6 design has incorporated further mitigation measures to provide additional protection of the public and the environment if the preventive measures fail. The EC6 design has an appropriate combination of inherent, passive safety characteristics, engineered features and administrative safety measures to effectively prevent and mitigate severe accident progressions. A strong contributor to the robustness and redundancy of CANDU design is the two-group separation philosophy. This ensures a high degree of independence between safety systems as well as physical separation and functional independence in how fundamental safety functions are provided. This paper will describe the following safety features based on the application of defence-in-depth and design approach to prevent beyond design basis events progressing to severe accidents and to mitigate the consequences if it occurs: Improved steam generator heat sink via a more reliable emergency heat removal system; Increased time before manual field actions are required via enhanced capacity of

Reliability-based optimization of engineering structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard

2008-01-01

The theoretical basis for reliability-based structural optimization within the framework of Bayesian statistical decision theory is briefly described. Reliability-based cost benefit problems are formulated and exemplitied with structural optimization. The basic reliability-based optimization...... problems are generalized to the following extensions: interactive optimization, inspection and repair costs, systematic reconstruction, re-assessment of existing structures. Illustrative examples are presented including a simple introductory example, a decision problem related to bridge re...

[Examination of safety improvement by failure record analysis that uses reliability engineering].

Science.gov (United States)

Kato, Kyoichi; Sato, Hisaya; Abe, Yoshihisa; Ishimori, Yoshiyuki; Hirano, Hiroshi; Higashimura, Kyoji; Amauchi, Hiroshi; Yanakita, Takashi; Kikuchi, Kei; Nakazawa, Yasuo

2010-08-20

How the maintenance checks of the medical treatment system, including start of work check and the ending check, was effective for preventive maintenance and the safety improvement was verified. In this research, date on the failure of devices in multiple facilities was collected, and the data of the trouble repair record was analyzed by the technique of reliability engineering. An analysis of data on the system (8 general systems, 6 Angio systems, 11 CT systems, 8 MRI systems, 8 RI systems, and the radiation therapy system 9) used in eight hospitals was performed. The data collection period assumed nine months from April to December 2008. Seven items were analyzed. (1) Mean time between failures (MTBF) (2) Mean time to repair (MTTR) (3) Mean down time (MDT) (4) Number found by check in morning (5) Failure generation time according to modality. The classification of the breakdowns per device, the incidence, and the tendency could be understood by introducing reliability engineering. Analysis, evaluation, and feedback on the failure generation history are useful to keep downtime to a minimum and to ensure safety.

Safety Criteria and Standards for Bearing Capacity of Foundation

Directory of Open Access Journals (Sweden)

Yanlong Li

2017-01-01

Full Text Available This paper focuses on the evaluation standards of factor of safety for foundation stability analysis. The problem of foundation stability is analyzed via the methods of risk analysis of engineering structures and reliability-based design, and the factor of safety for foundation stability is determined by using bearing capacity safety-factor method (BSFM and strength safety-factor method (SSFM. Based on a typical example, the admissible factors of safety were calibrated with a target reliability index specified in relevant standards. Two safety criteria and their standards of bearing capacity of foundation for these two methods (BSFM and SSFM were established. The universality of the safety criteria and their standards for foundation reliability was verified based on the concept of the ratio of safety margin (RSM.

Experience gained in enhancing operational safety at ComEd`s nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Elias, D [Commonwealth Edison Co. (United States)

1997-09-01

The following aspects of experience gained in enhancing operational safety at Comed`s nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight.

Nordic perspectives on safety management in high reliability organizations: Theory and applications

International Nuclear Information System (INIS)

Svenson, Ola; Salo, I.; Sjerve, A.B.; Reiman, T.; Oedewald, P.

2006-04-01

The chapters in this volume are written on a stand-alone basis meaning that the chapters can be read in any order. The first 4 chapters focus on theory and method in general with some applied examples illustrating the methods and theories. Chapters 5 and 6 are about safety management in the aviation industry with some additional information about incident reporting in the aviation industry and the health care sector. Chapters 7 through 9 cover safety management with applied examples from the nuclear power industry and with considerable validity for safety management in any industry. Chapters 10 through 12 cover generic safety issues with examples from the oil industry and chapter 13 presents issues related to organizations with different internal organizational structures. Although the many of the chapters use a specific industry to illustrate safety management, the messages in all the chapters are of importance for safety management in any high reliability industry or risky activity. The interested reader is also referred to, e.g., a document by an international NEA group (SEGHOF), who is about to publish a state of the art report on Systematic Approaches to Safety Management (cf., CSNI/NEA/SEGHOF, home page: www.nea.fr). (au)

Nordic perspectives on safety management in high reliability organizations: Theory and applications

Energy Technology Data Exchange (ETDEWEB)

Svenson, Ola; Salo, I; Sjerve, A B; Reiman, T; Oedewald, P [Stockholm Univ. (Sweden)

2006-04-15

The chapters in this volume are written on a stand-alone basis meaning that the chapters can be read in any order. The first 4 chapters focus on theory and method in general with some applied examples illustrating the methods and theories. Chapters 5 and 6 are about safety management in the aviation industry with some additional information about incident reporting in the aviation industry and the health care sector. Chapters 7 through 9 cover safety management with applied examples from the nuclear power industry and with considerable validity for safety management in any industry. Chapters 10 through 12 cover generic safety issues with examples from the oil industry and chapter 13 presents issues related to organizations with different internal organizational structures. Although the many of the chapters use a specific industry to illustrate safety management, the messages in all the chapters are of importance for safety management in any high reliability industry or risky activity. The interested reader is also referred to, e.g., a document by an international NEA group (SEGHOF), who is about to publish a state of the art report on Systematic Approaches to Safety Management (cf., CSNI/NEA/SEGHOF, home page: www.nea.fr). (au)

A formal safety analysis for PLC software-based safety critical system using Z

International Nuclear Information System (INIS)

Koh, Jung Soo

1997-02-01

This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

Reliability analysis on passive residual heat removal of AP1000 based on Grey model

Energy Technology Data Exchange (ETDEWEB)

Qi, Shi; Zhou, Tao; Shahzad, Muhammad Ali; Li, Yu [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Jiang, Guangming [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Laboratory

2017-06-15

It is common to base the design of passive systems on the natural laws of physics, such as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can affect the function of the system. The reliability of the passive residual heat removal is analysed.

Adaptation of the ToxRTool to Assess the Reliability of Toxicology Studies Conducted with Genetically Modified Crops and Implications for Future Safety Testing.

Science.gov (United States)

Koch, Michael S; DeSesso, John M; Williams, Amy Lavin; Michalek, Suzanne; Hammond, Bruce

2016-01-01

To determine the reliability of food safety studies carried out in rodents with genetically modified (GM) crops, a Food Safety Study Reliability Tool (FSSRTool) was adapted from the European Centre for the Validation of Alternative Methods' (ECVAM) ToxRTool. Reliability was defined as the inherent quality of the study with regard to use of standardized testing methodology, full documentation of experimental procedures and results, and the plausibility of the findings. Codex guidelines for GM crop safety evaluations indicate toxicology studies are not needed when comparability of the GM crop to its conventional counterpart has been demonstrated. This guidance notwithstanding, animal feeding studies have routinely been conducted with GM crops, but their conclusions on safety are not always consistent. To accurately evaluate potential risks from GM crops, risk assessors need clearly interpretable results from reliable studies. The development of the FSSRTool, which provides the user with a means of assessing the reliability of a toxicology study to inform risk assessment, is discussed. Its application to the body of literature on GM crop food safety studies demonstrates that reliable studies report no toxicologically relevant differences between rodents fed GM crops or their non-GM comparators.

Good performance in Japan is proof of continuing safety and reliability improvement practice

International Nuclear Information System (INIS)

Sumi, Y.

1987-01-01

Nuclear power is a vital energy supply source for both security and economy for such countries as Japan whose sources of energy are dependent on imported materials. This is the very reason why Japan gives her national priority to the improvement of nuclear power safety and reliability. As of the end of 1986, total nuclear power capacity owned and operated by private utility companies in Japan amounted to 24521 MW with 32 units sharing -- 19% of the total generating capacity. Moreover, during 1986 these units scored a remarkably high capacity factor of 76.2% and shared almost 28% of the nationwide electric power production, thereby contributing to a considerable saving of imported sources of energy. This outstanding record has been achieved by the parties concerned who dedicated themselves to furthering nuclear plant safety and reliability improvement. In this connection, this paper summarizes those key factors contributing to the good nuclear power plant performance of the Kansai Electric Power Company

Reliability and maintainability

International Nuclear Information System (INIS)

1994-01-01

Several communications in this conference are concerned with nuclear plant reliability and maintainability; their titles are: maintenance optimization of stand-by Diesels of 900 MW nuclear power plants; CLAIRE: an event-based simulation tool for software testing; reliability as one important issue within the periodic safety review of nuclear power plants; design of nuclear building ventilation by the means of functional analysis; operation characteristic analysis for a power industry plant park, as a function of influence parameters

The engineering project and reliability research of the safety interlock slow control system in BESIII

International Nuclear Information System (INIS)

Zhang Yinhong; Zhao Jingwei; Li Xiaonan; Xie Xiaoxi; Gao Cuishan; Bai Jingzhi; Chen Xihui; Min Jian; Nie Zhendong

2008-01-01

The new safety interlock slow control system of BESIII is designed to ensure that the BESIII interior equipments and the accelerator control center to work in coordination, and to guarantee the safety of the operating staff and all the important equipments at the same time. This paper introduces the hardware and software design of safety interlock system from the engineering requirements angle, including a detailed research on the software implementation technique of the state machine on PLC and the reliability of the system. (authors)