WorldWideScience

Sample records for safety systems reliability

  1. The reliability of nuclear power plant safety systems

    International Nuclear Information System (INIS)

    Susnik, J.

    1978-01-01

    A criterion was established concerning the protection that nuclear power plant (NPP) safety systems should afford. An estimate of the necessary or adequate reliability of the total complex of safety systems was derived. The acceptable unreliability of auxiliary safety systems is given, provided the reliability built into the specific NPP safety systems (ECCS, Containment) is to be fully utilized. A criterion for the acceptable unreliability of safety (sub)systems which occur in minimum cut sets having three or more components of the analysed fault tree was proposed. A set of input MTBF or MTTF values which fulfil all the set criteria and attain the appropriate overall reliability was derived. The sensitivity of results to input reliability data values was estimated. Numerical reliability evaluations were evaluated by the programs POTI, KOMBI and particularly URSULA, the last being based on Vesely's kinetic fault tree theory. (author)

  2. 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)

  3. 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

  4. 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)

  5. 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.

  6. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

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

  7. Software reliability and safety in nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1993-11-01

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

  8. Reliability and safety engineering

    CERN Document Server

    Verma, Ajit Kumar; Karanki, Durga Rao

    2016-01-01

    Reliability and safety are core issues that must be addressed throughout the life cycle of engineering systems. Reliability and Safety Engineering presents an overview of the basic concepts, together with simple and practical illustrations. The authors present reliability terminology in various engineering fields, viz.,electronics engineering, software engineering, mechanical engineering, structural engineering and power systems engineering. The book describes the latest applications in the area of probabilistic safety assessment, such as technical specification optimization, risk monitoring and risk informed in-service inspection. Reliability and safety studies must, inevitably, deal with uncertainty, so the book includes uncertainty propagation methods: Monte Carlo simulation, fuzzy arithmetic, Dempster-Shafer theory and probability bounds. Reliability and Safety Engineering also highlights advances in system reliability and safety assessment including dynamic system modeling and uncertainty management. Cas...

  9. 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.

  10. 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...

  11. 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

  12. 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

  13. 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

  14. Reliability Improved Design for a Safety System Channel

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Eung Se; Kim, Yun Goo [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced.

  15. Reliability Improved Design for a Safety System Channel

    International Nuclear Information System (INIS)

    Oh, Eung Se; Kim, Yun Goo

    2016-01-01

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced

  16. 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

  17. Reliability estimation of safety-critical software-based systems using Bayesian networks

    International Nuclear Information System (INIS)

    Helminen, A.

    2001-06-01

    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 software-based safety-critical automation systems in nuclear power plants. In the research project 'Programmable automation system safety integrity assessment (PASSI)', belonging to the Finnish Nuclear Safety Research Programme (FINNUS, 1999-2002), various safety assessment methods and tools for software based systems are developed and evaluated. The project is 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). In this report the applicability of Bayesian networks to the reliability estimation of software-based systems is studied. The applicability is evaluated by building Bayesian network models for the systems of interest and performing simulations for these models. In the simulations hypothetical evidence is used for defining the parameter relations and for determining the ability to compensate disparate evidence in the models. Based on the experiences from modelling and simulations we are able to conclude that Bayesian networks provide a good method for the reliability estimation of software-based systems. (orig.)

  18. 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

  19. 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.

  20. 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.

  1. 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

  2. Safety and reliability criteria

    International Nuclear Information System (INIS)

    O'Neil, R.

    1978-01-01

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

  3. 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)

  4. 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)

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. 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)

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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.)

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. Reliability analysis of software based safety functions

    International Nuclear Information System (INIS)

    Pulkkinen, U.

    1993-05-01

    The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

  9. 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

  10. 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

  11. An approach for assessing ALWR passive safety system reliability

    International Nuclear Information System (INIS)

    Hake, T.M.

    1991-01-01

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

  12. 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)

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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

  19. Reliability analysis of repairable safety systems of a reprocessing plant allowing for tolerable system downtimes

    International Nuclear Information System (INIS)

    Schaefer, H.

    1987-01-01

    GRS has been engaged in safety analysises of the German Reprocessing Plant for several years. The development and verification of appropriate reliability analysis methods, the generation of data as well as the search for an adequate structural presentation of the results to form a basis of recommendations for technical or administrative measures or contributions to risk oriented evaluations have been or are in the process of being established. In contrast to NPP-studies, the reliability assessment of safety systems of a reprocessing plant is applied to repairable and often relatively small systems allowing for tolerable system downtimes. A sketch of the diverse cooling systems of a vessel containing a selfheating solution is given. The interruption of the cooling function for about one day might be tolerable before boiling will be reached. This interval is suitable for transfer of the solution to a spare vessel or for repairing the failed components, thus restoring the cooling function

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. Reliability Approach of a Compressor System using Reliability Block ...

    African Journals Online (AJOL)

    pc

    2018-03-05

    Mar 5, 2018 ... This paper presents a reliability analysis of such a system using reliability ... Keywords-compressor system, reliability, reliability block diagram, RBD .... the same structure has been kept with the three subsystems: air flow, oil flow and .... and Safety in Engineering Design", Springer, 2009. [3] P. O'Connor ...

  5. 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)

  6. 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)

  7. 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)

  8. A probabilistic approach to safety/reliability of space nuclear power systems

    International Nuclear Information System (INIS)

    Medford, G.; Williams, K.; Kolaczkowski, A.

    1989-01-01

    An ongoing effort is investigating the feasibility of using probabilistic risk assessment (PRA) modeling techniques to construct a living model of a space nuclear power system. This is being done in conjunction with a traditional reliability and survivability analysis of the SP-100 space nuclear power system. The initial phase of the project consists of three major parts with the overall goal of developing a top-level system model and defining initiating events of interest for the SP-100 system. The three major tasks were performing a traditional survivability analysis, performing a simple system reliability analysis, and constructing a top-level system fault-tree model. Each of these tasks and their interim results are discussed in this paper. Initial results from the study support the conclusion that PRA modeling techniques can provide a valuable design and decision-making tool for space reactors. The ability of the model to rank and calculate relative contributions from various failure modes allows design optimization for maximum safety and reliability. Future efforts in the SP-100 program will see data development and quantification of the model to allow parametric evaluations of the SP-100 system. Current efforts have shown the need for formal data development and test programs within such a modeling framework

  9. 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.

  10. The REPAS approach to the evaluation of passive safety systems reliability

    International Nuclear Information System (INIS)

    Bianchi, F.; Burgazzi, L.; D'Auria, F.; Ricotti, M.E.

    2002-01-01

    Scope of this research, carried out by ENEA in collaboration with University of Pisa and Polytechnic of Milano since 1999, is the identification of a methodology allowing the evaluation of the reliability of passive systems as a whole, in a more physical and phenomenal way. The paper describe the study, named REPAS (Reliability Evaluation of Passive Safety systems), carried out by the partners and finalised to the development and validation of such a procedure. The strategy of engagement moves from the consideration that a passive system should be theoretically more reliable than an active one. In fact it does not need any external input or energy to operate and it relies only upon natural physical laws (e.g. gravity, natural circulation, internally stored energy, etc.) and/or 'intelligent' use of the energy inherently available in the system (e.g. chemical reaction, decay heat, etc.). Nevertheless the passive system may fail its mission not only as a consequence of classical mechanical failure of components, but also for deviation from the expected behaviour, due to physical phenomena mainly related to thermal-hydraulics or due to different boundary and initial conditions. The main sources of physical failure are identified and a probability of occurrence is assigned. The reliability analysis is performed on a passive system which operates in two-phase, natural circulation. The selected system is a loop including a heat source and a heat sink where the condensation occurs. The system behaviour under different configurations has been simulated via best-estimate code (Relap5 mod3.2). The results are shown and can be treated in such a way to give qualitative and quantitative information on the system reliability. Main routes of development of the methodology are also depicted. The analysis of the results shows that the procedure is suitable to evaluate the performance of a passive system on a probabilistic / deterministic basis. Important information can also be

  11. 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

  12. 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

  13. Reliability of large and complex systems

    CERN Document Server

    Kolowrocki, Krzysztof

    2014-01-01

    Reliability of Large and Complex Systems, previously titled Reliability of Large Systems, is an innovative guide to the current state and reliability of large and complex systems. In addition to revised and updated content on the complexity and safety of large and complex mechanisms, this new edition looks at the reliability of nanosystems, a key research topic in nanotechnology science. The author discusses the importance of safety investigation of critical infrastructures that have aged or have been exposed to varying operational conditions. This reference provides an asympt

  14. Fundamentals and applications of systems reliability analysis

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  15. 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

  16. Kilowatt isotope power system. Phase II plan. Volume V. Safety, quality assurance and reliability

    International Nuclear Information System (INIS)

    1978-01-01

    The development of a Kilowatt Isotope Power System (KIPS) was begun in 1975 for the purpose of satisfying the power requirements of satellites in the 1980's. The KIPS is a 238 PuO 2 -fueled organic Rankine cycle turbine power system to provide a design output of 500 to 2000 W. Included in this volume are: launch and flight safety considerations; quality assurance techniques and procedures to be followed through system fabrication, assembly and inspection; and the reliability program made up of reliability prediction analysis, failure mode analysis and criticality analysis

  17. 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

  18. 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

  19. Proceedings of the SRESA national conference on reliability and safety engineering

    International Nuclear Information System (INIS)

    Varde, P.V.; Vaishnavi, P.; Sujatha, S.; Valarmathi, A.

    2014-01-01

    The objective of this conference was to provide a forum for technical discussions on recent developments in the area of risk based approach and Prognostic Health Management of critical systems in decision making. The reliability and safety engineering methods are concerned with the way which the product fails, and the effects of failure is to understand how a product works and assures acceptable levels of safety. The reliability engineering addresses all the anticipated and possibly unanticipated causes of failure to ensure the occurrence of failure is prevented or minimized. The topics discussed in the conference were: Reliability in Engineering Design, Safety Assessment and Management, Reliability analysis and Assessment , Stochastic Petri nets for reliability Modeling, Dynamic Reliability, Reliability Prediction, Hardware Reliability, Software Reliability in Safety Critical Issues, Probabilistic Safety Assessment, Risk Informed Approach, Dynamic Models for Reliability Analysis, Reliability based Design and Analysis, Prognostics and Health Management, Remaining Useful Life (RUL), Human Reliability Modeling, Risk Based Applications, Hazard and Operability Study (HAZOP), Reliability in Network Security and Quality Assurance and Management etc. The papers relevant to INIS are indexed separately

  20. Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

    Directory of Open Access Journals (Sweden)

    Franco Pierro

    2009-01-01

    Full Text Available The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way 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. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks' formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

  1. 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)

  2. 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)

  3. 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)

  4. 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

  5. Addressing Unison and Uniqueness of Reliability and Safety for Better Integration

    Science.gov (United States)

    Huang, Zhaofeng; Safie, Fayssal

    2015-01-01

    For a long time, both in theory and in practice, safety and reliability have not been clearly differentiated, which leads to confusion, inefficiency, and sometime counter-productive practices in executing each of these two disciplines. It is imperative to address the uniqueness and the unison of these two disciplines to help both disciplines become more effective and to promote a better integration of the two for enhancing safety and reliability in our products as an overall objective. There are two purposes of this paper. First, it will investigate the uniqueness and unison of each discipline and discuss the interrelationship between the two for awareness and clarification. Second, after clearly understanding the unique roles and interrelationship between the two in a product design and development life cycle, we offer suggestions to enhance the disciplines with distinguished and focused roles, to better integrate the two, and to improve unique sets of skills and tools of reliability and safety processes. From the uniqueness aspect, the paper identifies and discusses the respective uniqueness of reliability and safety from their roles, accountability, nature of requirements, technical scopes, detailed technical approaches, and analysis boundaries. It is misleading to equate unreliable to unsafe, since a safety hazard may or may not be related to the component, sub-system, or system functions, which are primarily what reliability addresses. Similarly, failing-to-function may or may not lead to hazard events. Examples will be given in the paper from aerospace, defense, and consumer products to illustrate the uniqueness and differences between reliability and safety. From the unison aspect, the paper discusses what the commonalities between reliability and safety are, and how these two disciplines are linked, integrated, and supplemented with each other to accomplish the customer requirements and product goals. In addition to understanding the uniqueness in

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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.

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-26

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

  13. Reliability study: digital engineered safety feature actuation system of Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

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

    2003-04-01

    The usage of digital Instrumentation and Control (I and C) in a nuclear power plant becomes more extensive, including safety related systems. The PSA application of these new designs are very important in order to evaluate their reliability. In particular, Korean Standard Nuclear Power Plants (KSNPPs), typically Ulchin 5 and 6 (UCN 5 and 6) reactor units, adopted the digital safety-critical systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS). In this research, we developed fault tree models for assessing the unavailability of the DESFAS functions. We also performed an analysis of the quantification results. The unavailability results of different DESFAS functions showed that their values are comprised from 5.461E-5 to 3.14E-4. The system unavailability of DESFAS AFAS-1 is estimated as 5.461E-5, which is about 27% less than that of analog system if we consider the difference of human failure probability estimation between both analyses. The results of this study could be utilized in risk-effect analysis of KSNPP. We expect that the safety analysis result will contribute to design feedback

  14. 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)

  15. 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.)

  16. 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

  17. X-real-time executive (X-RTE) an ultra-high reliable real-time executive for safety critical systems

    International Nuclear Information System (INIS)

    Suresh Babu, R.M.

    1995-01-01

    With growing number of application of computers in safety critical systems of nuclear plants there has been a need to assure high quality and reliability of the software used in these systems. One way to assure software quality is to use qualified software components. Since the safety systems and control systems are real-time systems there is a need for a real-time supervisory software to guarantee temporal response of the system. This report describes one such software package, called X-Real-Time Executive (or X-RTE), which was developed in Reactor Control Division, BARC. The report describes all the capabilities and unique features of X-RTE and compares it with a commercially available operating system. The features of X-RTE include pre-emptive scheduling, process synchronization, inter-process communication, multi-processor support, temporal support, debug facility, high portability, high reliability, high quality, and extensive documentation. Examples have been used very liberally to illustrate the underlying concepts. Besides, the report provides a brief description about the methods used, during the software development, to assure high quality and reliability of X-RTE. (author). refs., 11 figs., tabs

  18. 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

  19. Reliability analysis of reactor protection systems

    International Nuclear Information System (INIS)

    Alsan, S.

    1976-07-01

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

  20. Reliability prediction for the vehicles equipped with advanced driver assistance systems (ADAS and passive safety systems (PSS

    Directory of Open Access Journals (Sweden)

    Balbir S. Dhillon

    2012-10-01

    Full Text Available The human error has been reported as a major root cause in road accidents in today’s world. The human as a driver in road vehicles composed of human, mechanical and electrical components is constantly exposed to changing surroundings (e.g., road conditions, environmentwhich deteriorate the driver’s capacities leading to a potential accident. The auto industries and transportation authorities have realized that similar to other complex and safety sensitive transportation systems, the road vehicles need to rely on both advanced technologies (i.e., Advanced Driver Assistance Systems (ADAS and Passive Safety Systems (PSS (e.g.,, seatbelts, airbags in order to mitigate the risk of accidents and casualties. In this study, the advantages and disadvantages of ADAS as active safety systems as well as passive safety systems in road vehicles have been discussed. Also, this study proposes models that analyze the interactions between human as a driver and ADAS Warning and Crash Avoidance Systems and PSS in the design of vehicles. Thereafter, the mathematical models have been developed to make reliability prediction at any given time on the road transportation for vehicles equipped with ADAS and PSS. Finally, the implications of this study in the improvement of vehicle designs and prevention of casualties are discussed.

  1. New design of engineered safety features-component control system to improve performance and reliability

    International Nuclear Information System (INIS)

    Kim, S.T.; Jung, H.W.; Lee, S.J.; Cho, C.H.; Kim, D.H.; Kim, H.

    2006-01-01

    Full text: Full text: The Engineered Safety Features-Component Control System (ESF-CCS) controls the engineered safety features of a Nuclear Power Plant such as Solenoid Operated Valves (SOV), Motor Operated Valves (MOV), pumps, dampers, etc. to mitigate the effects of a Design Basis Accident (DBA) or an abnormal operation. ESF-CCS serves as an interface system between the Plant Protection System (PPS) and remote actuation devices. ESF-CCS is composed of fault tolerant Group Controllers GC, Loop Controllers (LC), ESF-CCS Test and Interface Processor (ETIP) and Cabinet Operator Module (COM) and Control Channel Gateway (CCG) etc. GCs in each division are designed to be fully independent triple configuration, which perform system level NSSS and BOP ESFAS logic (2-out-of-4 logic and l-out-of-2 logic, respectively) making it possible to test each GC individually during normal operation. In the existing configuration, the safety-related plant component control is part of the Plant Control System (PCS) non-safety system. For increased safety and reliability, this design change incorporates this part into the LCs, and is therefore designed according to the safety-critical system procedures. The test and diagnosis capabilities of ETIP and COM are reinforced. By means of an automatic periodic test for all main functions of the system, it is possible to quickly determine an abnormal status of the system, and to decrease the elapsed time for tests, thus effectively increasing availability. ESF-CCS consists of four independent divisions (A, B, C, and D) in the Advanced Power Reactor 1400 (APR1400). One prototype division is being manufactured and will be tested

  2. Advances in reliability and system engineering

    CERN Document Server

    Davim, J

    2017-01-01

    This book presents original studies describing the latest research and developments in the area of reliability and systems engineering. It helps the reader identifying gaps in the current knowledge and presents fruitful areas for further research in the field. Among others, this book covers reliability measures, reliability assessment of multi-state systems, optimization of multi-state systems, continuous multi-state systems, new computational techniques applied to multi-state systems and probabilistic and non-probabilistic safety assessment.

  3. 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)

  4. 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

  5. Safety and reliability in Europe

    International Nuclear Information System (INIS)

    Colombo, A.G.

    1985-01-01

    This volume contains the papers presented at the ESRA Pre-Launching Meeting. The meeting was attended by about eighty European reliability and safety experts from industry, research organizations and universities. This meeting was dealing with the following subjects: the historical perspective of safety and reliability in Europe and to the aims of ESRA. Status and Trends in Research and Development; Codes, Standards and Regulations; Academic and Technical Training. National and international Organizations. Twenty six papers have been analyzed and abstracted for inclusion in the data base

  6. 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

  7. 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...

  8. 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)

  9. 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)

  10. A Review: Passive System Reliability Analysis – Accomplishments and Unresolved Issues

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, Arun Kumar, E-mail: arunths@barc.gov.in [Reactor Engineering Division, Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai (India); Chandrakar, Amit [Homi Bhabha National Institute, Mumbai (India); Vinod, Gopika [Reactor Safety Division, Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai (India)

    2014-10-10

    Reliability assessment of passive safety systems is one of the important issues, since safety of advanced nuclear reactors rely on several passive features. In this context, a few methodologies such as reliability evaluation of passive safety system (REPAS), reliability methods for passive safety functions (RMPS), and analysis of passive systems reliability (APSRA) have been developed in the past. These methodologies have been used to assess reliability of various passive safety systems. While these methodologies have certain features in common, but they differ in considering certain issues; for example, treatment of model uncertainties, deviation of geometric, and process parameters from their nominal values. This paper presents the state of the art on passive system reliability assessment methodologies, the accomplishments, and remaining issues. In this review, three critical issues pertaining to passive systems performance and reliability have been identified. The first issue is applicability of best estimate codes and model uncertainty. The best estimate codes based phenomenological simulations of natural convection passive systems could have significant amount of uncertainties, these uncertainties must be incorporated in appropriate manner in the performance and reliability analysis of such systems. The second issue is the treatment of dynamic failure characteristics of components of passive systems. REPAS, RMPS, and APSRA methodologies do not consider dynamic failures of components or process, which may have strong influence on the failure of passive systems. The influence of dynamic failure characteristics of components on system failure probability is presented with the help of a dynamic reliability methodology based on Monte Carlo simulation. The analysis of a benchmark problem of Hold-up tank shows the error in failure probability estimation by not considering the dynamism of components. It is thus suggested that dynamic reliability methodologies must be

  11. 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

  12. Safety systems I/C reliability analysis of the Kozloduy NPP units 5 and 6; Analiz nadezhnosti KIP i sistem bezopasnosti pyatogo i shestogo bloka AEhS `Kozloduy`

    Energy Technology Data Exchange (ETDEWEB)

    Marinova, B [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    The purpose of the analysis is to assess the safety systems I/C equipment reliability of the Kozloduy-5 and the Kozloduy-6 reactors. The assessment of quantitative and qualitative effect of control systems unavailability on the safety systems unavailability is performed. The analysis is limited to the following systems: sprinkler management, low pressure emergency spray, emergency injection of boric acid, hydro accumulators, pressure compensator and compressed air. The code for probabilistic safety assessment PSAPACK has been used in analysis. Fault trees for all analysed safety systems have been constructed. Results indicates a high reliability of the safety systems management.

  13. RELOSS, Reliability of Safety System by Fault Tree Analysis

    International Nuclear Information System (INIS)

    Allan, R.N.; Rondiris, I.L.; Adraktas, A.

    1981-01-01

    1 - Description of problem or function: Program RELOSS is used in the reliability/safety assessment of any complex system with predetermined operational logic in qualitative and (if required) quantitative terms. The program calculates the possible system outcomes following an abnormal operating condition and the probability of occurrence, if required. Furthermore, the program deduces the minimal cut or tie sets of the system outcomes and identifies the potential common mode failures. 4. Method of solution: The reliability analysis performed by the program is based on the event tree methodology. Using this methodology, the program develops the event tree of a system or a module of that system and relates each path of this tree to its qualitative and/or quantitative impact on specified system or module outcomes. If the system being analysed is subdivided into modules the program assesses each module in turn as described previously and then combines the module information to obtain results for the overall system. Having developed the event tree of a module or a system, the program identifies which paths lead or do not lead to various outcomes depending on whether the cut or the tie sets of the outcomes are required and deduces the corresponding sets. Furthermore the program identifies for a specific system outcome, the potential common mode failures and the cut or tie sets containing potential dependent failures of some components. 5. Restrictions on the complexity of the problem: The present dimensions of the program are as follows. They can however be easily modified: Maximum number of modules (equivalent components): 25; Maximum number of components in a module: 15; Maximum number of levels of parentheses in a logical statement: 10 Maximum number of system outcomes: 3; Maximum number of module outcomes: 2; Maximum number of points in time for which quantitative analysis is required: 5; Maximum order of any cut or tie set: 10; Maximum order of a cut or tie of any

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

    International Nuclear Information System (INIS)

    1979-01-01

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

  15. Development of reliability-based safety enhancement technology

    International Nuclear Information System (INIS)

    Kim, Kil Yoo; Han, Sang Hoon; Jang, Seung Cherl

    2002-04-01

    This project aims to develop critical technologies and the necessary reliability DB for maximizing the economics in the NPP operation with keeping the safety using the information of the risk (or reliability). For the research goal, firstly the four critical technologies(Risk Informed Tech. Spec. Optimization, Risk Informed Inservice Testing, On-line Maintenance, Maintenance Rule) for RIR and A have been developed. Secondly, KIND (Korea Information System for Nuclear Reliability Data) has been developed. Using KIND, YGN 3,4 and UCN 3,4 component reliability DB have been established. A reactor trip history DB for all NPP in Korea also has been developed and analyzed. Finally, a detailed reliability analysis of RPS/ESFAS for KNSP has been performed. With the result of the analysis, the sensitivity analysis also has been performed to optimize the AOT/STI of tech. spec. A statistical analysis procedure and computer code have been developed for the set point drift analysis

  16. Increasing nuclear safety and operational reliability by upgrading the charging pump mechanical sealing system

    International Nuclear Information System (INIS)

    Loenhout, Gerard van; Nilsson, Peter; Jehander, Magnus

    2016-01-01

    For the Ringhals-2 nuclear power plant, three installed centrifugal pumps were designated to have a combined High Head Safety Injection function, as well as a Chemical Volume Control System function. The pumps were originally installed with rubber bellow type mechanical seals, which over time had demonstrated an unreliable sealing performance by displaying high leakages. In 2002, the Ringhals Maintenance engineers initiated to identify a more reliable and robust shaft sealing solution. In 2007, the project was launched and the installation of the first, new mechanical sealing solution took place in the autumn of 2011. In October 2014, these mechanical seals were dismantled and inspected. The inspection confirmed the expected reliability of the new solution.

  17. Increasing nuclear safety and operational reliability by upgrading the charging pump mechanical sealing system

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve Corporation, Etten-Leur (Netherlands); Nilsson, Peter [Flowsys Technologies AB, Moelndal (Sweden); Jehander, Magnus [Ringhals AB, Vaeroebacka (Sweden)

    2016-07-01

    For the Ringhals-2 nuclear power plant, three installed centrifugal pumps were designated to have a combined High Head Safety Injection function, as well as a Chemical Volume Control System function. The pumps were originally installed with rubber bellow type mechanical seals, which over time had demonstrated an unreliable sealing performance by displaying high leakages. In 2002, the Ringhals Maintenance engineers initiated to identify a more reliable and robust shaft sealing solution. In 2007, the project was launched and the installation of the first, new mechanical sealing solution took place in the autumn of 2011. In October 2014, these mechanical seals were dismantled and inspected. The inspection confirmed the expected reliability of the new solution.

  18. Increasing nuclear safety and operational reliability by upgrading the charging pump mechanical sealing system

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve Corporation, Etten-Leur (Netherlands); Nilsson, Peter [Flowsys Technologies AB, Moelndal (Sweden); Jehander, Magnus [Ringhals AB, Vaeroebacka (Sweden)

    2016-03-15

    For the Ringhals-2 nuclear power plant, three installed centrifugal pumps were designated to have a combined High Head Safety Injection function, as well as a Chemical Volume Control System function. The pumps were originally installed with rubber bellow type mechanical seals, which over time had demonstrated an unreliable sealing performance by displaying high leakages. In 2002, the Ringhals Maintenance engineers initiated to identify a more reliable and robust shaft sealing solution. In 2007, the project was launched and the installation of the first, new mechanical sealing solution took place in the autumn of 2011. In October 2014, these mechanical seals were dismantled and inspected. The inspection confirmed the expected reliability of the new solution.

  19. Reliability analysis of the recirculation phase of the safety injection system of Angra-1

    International Nuclear Information System (INIS)

    Rivera, R.R.J.M.

    1981-09-01

    The calculation of several reliability parameters-failure probability, unavailability and unreliability - of the recirculation phase of the safety injection system of Angra-1, was done. This system has two distinct modes of operation (short term and long term) which were fault tree analysed both separately and as a whole. To obtain quantitative results the computer codes SAMPLE and PRET-KITT were utilized. The former was used to consider the uncertainties in the failure data (drawn integrally from WASH-1400) and the latter to obtain time dependent unreliability values. Hardware failures and common-mode failures were considered. Altough the analysis methods employed here differ somewhat from those used in WASH-1400, the results which could be compared were found to have the order of magnitude. A viability study of some suggestions of system's modifications was performed, and it has shown that some significant reliability improvements can be achieved with reasonably simple changes. (Author) [pt

  20. IEEE standard requirements for reliability analysis in the design and operation of safety systems for nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The purpose of this standard is to provide uniform, minimum acceptable requirements for the performance of reliability analyses for safety-related systems found in nuclear-power generating stations, but not to define the need for an analysis. The need for reliability analysis has been identified in other standards which expand the requirements of regulations (e.g., IEEE Std 379-1972 (ANSI N41.2-1972), ''Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection System,'' which describes the application of the single-failure criterion). IEEE Std 352-1975, ''Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systems,'' provides guidance in the application and use of reliability techniques referred to in this standard

  1. 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.

  2. Photovoltaic power system reliability considerations

    Science.gov (United States)

    Lalli, V. R.

    1980-01-01

    This paper describes an example of how modern engineering and safety techniques can be used to assure the reliable and safe operation of photovoltaic power systems. This particular application was for a solar cell power system demonstration project in Tangaye, Upper Volta, Africa. The techniques involve a definition of the power system natural and operating environment, use of design criteria and analysis techniques, an awareness of potential problems via the inherent reliability and FMEA methods, and use of a fail-safe and planned spare parts engineering philosophy.

  3. SAFETY CRITERION IN ASSESSING THE IMPORTANCE OF AN ELEMENT IN THE COMPLEX TECHNOLOGICAL SYSTEM RELIABILITY STRUCTURE

    Directory of Open Access Journals (Sweden)

    Leszek CHYBOWSKI

    2012-01-01

    Full Text Available The paper presents the need to develop a description of the importance of the technological systems reliability structure elements in terms of security of the system. Basic issues related to the exploration of weak links and important elements in the system as well as a proposal to develop the current approach to assessing the importance of the system components have been presented. Moreover, the differences between the unreliability of suitability and unreliability of safety have been pointed out.

  4. 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)

  5. Reactor system safety assurance

    International Nuclear Information System (INIS)

    Mattson, R.J.

    1984-01-01

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  6. 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)

  7. 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)

  8. 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

  9. 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

  10. Power system reliability analysis using fault trees

    International Nuclear Information System (INIS)

    Volkanovski, A.; Cepin, M.; Mavko, B.

    2006-01-01

    The power system reliability analysis method is developed from the aspect of reliable delivery of electrical energy to customers. The method is developed based on the fault tree analysis, which is widely applied in the Probabilistic Safety Assessment (PSA). The method is adapted for the power system reliability analysis. The method is developed in a way that only the basic reliability parameters of the analysed power system are necessary as an input for the calculation of reliability indices of the system. The modeling and analysis was performed on an example power system consisting of eight substations. The results include the level of reliability of current power system configuration, the combinations of component failures resulting in a failed power delivery to loads, and the importance factors for components and subsystems. (author)

  11. 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

  12. 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.

  13. System reliability developments in structural engineering

    International Nuclear Information System (INIS)

    Moses, F.

    1982-01-01

    Two major limitations occur in present structural design code developments utilizing reliability theory. The notional system reliabilities may differ significantly from calibrated component reliabilities. Secondly, actual failures are often due to gross errors not reflected in most present code formats. A review is presented of system reliability methods and further new concepts are developed. The incremental load approach for identifying and expressing collapse modes is expanded by employing a strategy to identify and enumerate the significant structural collapse modes. It further isolates the importance of critical components in the system performance. Ductile and brittle component behavior and strength correlation is reflected in the system model and illustrated in several examples. Modal combinations for the system reliability are also reviewed. From these developments a system factor can be addended to component safety checking equations. Values may be derived from system behavior by substituting in a damage model which accounts for the response range from component failure to collapse. Other strategies are discussed which emphasize quality assurance during design and in-service inspection for components whose behavior is critical to the system reliability. (Auth.)

  14. 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)

  15. 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

  16. Reliability analyses of safety systems for WWER-440 nuclear power plants

    International Nuclear Information System (INIS)

    Dusek, J.; Hojny, V.

    1985-01-01

    The UJV in Rez near Prague studied the reliability of the system of emergency core cooling and of the system for suppressing pressure in the sealed area of the nuclear power plant in the occurrence of a loss-of-coolant accident. The reliability of the systems was evaluated by failure tree analysis. Simulation and analytical calculation programs were developed and used for the reliability analysis. The results are briefly presented of the reliability analyses of the passive system for the immediate short-term flooding of the reactor core, of the active low-pressure system of emergency core cooling, the spray system, the bubble-vacuum system and the system of emergency supply of the steam generators. (E.S.)

  17. Reliability of operating WWER monitoring systems

    International Nuclear Information System (INIS)

    Yastrebenetsky, M.A.; Goldrin, V.M.; Garagulya, A.V.

    1996-01-01

    The elaboration of WWER monitoring systems reliability measures is described in this paper. The evaluation is based on the statistical data about failures what have collected at the Ukrainian operating nuclear power plants (NPP). The main attention is devoted to radiation safety monitoring system and unit information computer system, what collects information from different sensors and system of the unit. Reliability measures were used for decision the problems, connected with life extension of the instruments, and for other purposes. (author). 6 refs, 6 figs

  18. Reliability of operating WWER monitoring systems

    Energy Technology Data Exchange (ETDEWEB)

    Yastrebenetsky, M A; Goldrin, V M; Garagulya, A V [Ukrainian State Scientific Technical Center of Nuclear and Radiation Safety, Kharkov (Ukraine). Instrumentation and Control Systems Dept.

    1997-12-31

    The elaboration of WWER monitoring systems reliability measures is described in this paper. The evaluation is based on the statistical data about failures what have collected at the Ukrainian operating nuclear power plants (NPP). The main attention is devoted to radiation safety monitoring system and unit information computer system, what collects information from different sensors and system of the unit. Reliability measures were used for decision the problems, connected with life extension of the instruments, and for other purposes. (author). 6 refs, 6 figs.

  19. Use of F.M.E.A. for reliability analysis of safety systems in nuclear power plants

    International Nuclear Information System (INIS)

    Barbet, J.F.; Llory, M.; Villemeur, A.

    1982-01-01

    In the framework of the French nuclear power plant program, reliability studies of safety systems have been carried out at the Electricite de France since 1975. The main results of the studies are examined; about the methodological aspects it appears useful to develop an inductive approach such as the Failure Modes and Effects Analysis (F.M.E.A.). The method is described with its advantages and limitations; the possibilities of use of F.M.E.A. to solve specific safety problems are investigated. To conclude, the future trends of research and development in this field at Electricite de France are pointed out [fr

  20. CADRIGS--computer aided design reliability interactive graphics system

    International Nuclear Information System (INIS)

    Kwik, R.J.; Polizzi, L.M.; Sticco, S.; Gerrard, P.B.; Yeater, M.L.; Hockenbury, R.W.; Phillips, M.A.

    1982-01-01

    An integrated reliability analysis program combining graphic representation of fault trees, automated data base loadings and reference, and automated construction of reliability code input files was developed. The functional specifications for CADRIGS, the computer aided design reliability interactive graphics system, are presented. Previously developed fault tree segments used in auxiliary feedwater system safety analysis were constructed on CADRIGS and, when combined, yielded results identical to those resulting from manual input to the same reliability codes

  1. Reliability and protection against failure in computer systems

    International Nuclear Information System (INIS)

    Daniels, B.K.

    1979-01-01

    Computers are being increasingly integrated into the control and safety systems of large and potentially hazardous industrial processes. This development introduces problems which are particular to computer systems and opens the way to new techniques of solving conventional reliability and availability problems. References to the developing fields of software reliability, human factors and software design are given, and these subjects are related, where possible, to the quantified assessment of reliability. Original material is presented in the areas of reliability growth and computer hardware failure data. The report draws on the experience of the National Centre of Systems Reliability in assessing the capability and reliability of computer systems both within the nuclear industry, and from the work carried out in other industries by the Systems Reliability Service. (author)

  2. Failure Modes Effects and Criticality Analysis, an Underutilized Safety, Reliability, Project Management and Systems Engineering Tool

    Science.gov (United States)

    Mullin, Daniel Richard

    2013-09-01

    The majority of space programs whether manned or unmanned for science or exploration require that a Failure Modes Effects and Criticality Analysis (FMECA) be performed as part of their safety and reliability activities. This comes as no surprise given that FMECAs have been an integral part of the reliability engineer's toolkit since the 1950s. The reasons for performing a FMECA are well known including fleshing out system single point failures, system hazards and critical components and functions. However, in the author's ten years' experience as a space systems safety and reliability engineer, findings demonstrate that the FMECA is often performed as an afterthought, simply to meet contract deliverable requirements and is often started long after the system requirements allocation and preliminary design have been completed. There are also important qualitative and quantitative components often missing which can provide useful data to all of project stakeholders. These include; probability of occurrence, probability of detection, time to effect and time to detect and, finally, the Risk Priority Number. This is unfortunate as the FMECA is a powerful system design tool that when used effectively, can help optimize system function while minimizing the risk of failure. When performed as early as possible in conjunction with writing the top level system requirements, the FMECA can provide instant feedback on the viability of the requirements while providing a valuable sanity check early in the design process. It can indicate which areas of the system will require redundancy and which areas are inherently the most risky from the onset. Based on historical and practical examples, it is this author's contention that FMECAs are an immense source of important information for all involved stakeholders in a given project and can provide several benefits including, efficient project management with respect to cost and schedule, system engineering and requirements management

  3. 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.).

  4. 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.)

  5. 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

  6. 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

  7. Autonomous safety and reliability features of the K-1 avionics system

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, G.E.; Kohrs, D.; Bailey, R.; Lai, G. [Kistler Aerospace Corp., Kirkland, WA (United States)

    2004-03-01

    Kistler Aerospace Corporation is developing the K-1, a fully reusable, two-stage-to-orbit launch vehicle. Both stages return to the launch site using parachutes and airbags. Initial flight operations will occur from Woomera, Australia. K-1 guidance is performed autonomously. Each stage of the K- 1 employs a triplex, fault tolerant avionics architecture, including three fault tolerant computers and three radiation hardened Embedded GPS/INS units with a hardware voter. The K-1 has an Integrated Vehicle Health Management (IVHM) system on each stage residing in the three vehicle computers based on similar systems in commercial aircraft. During first-stage ascent, the IVHM system performs an Instantaneous Impact Prediction (IIP) calculation 25 times per second, initiating an abort in the event the vehicle is outside a predetermined safety corridor for at least three consecutive calculations. In this event, commands are issued to terminate thrust, separate the stages, dump all propellant in the first-stage, and initiate a normal landing sequence. The second-stage flight computer calculates its ability to reach orbit along its state vector, initiating an abort sequence similar to the first stage if it cannot. On a nominal mission, following separation, the second-stage also performs calculations to assure its impact point is within a safety corridor. The K-1's guidance and control design is being tested through simulation with hardware-in-the-loop at Draper Laboratory. Kistler's verification strategy assures reliable and safe operation of the K-1. (author)

  8. 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)

  9. 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.

  10. 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.

  11. 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

  12. Benefits of a systematic approach to maintenance for safety and safety related systems

    International Nuclear Information System (INIS)

    Dam, R.F.; Ayazzudin, S.; Nickerson, J.H.

    2003-01-01

    For safety and safety-related 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 results in system unavailability 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. In such cases, the value of the application of a Systematic Assessment of Maintenance (SAM) process, such as Reliability Centered Maintenance (RCM), is questioned. The special case of Standby-Safety systems was discussed in a previous paper, where it was demonstrated how SAM techniques provide useful insight into current system performance, the impact of testing on component and system reliability, and how PSA considerations can be integrated into a comprehensive Maintenance, Surveillance, and Inspection (MSI) strategy. Although the system reliability requirements are an important part of the strategy evaluation, SAM techniques provide a systematic assessment within a broader context. 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 known component degradation mechanisms. This thinking can be extended to safety and safety related systems in general. Over the past 6 years, AECL has been working with CANDU utilities in the development and implementation of a comprehensive and integrated Plant Life Management (PLiM) program. As part of developing a comprehensive plant asset management approach, SAM techniques are used to develop a technical basis that not only works towards ensuring reliable operation of plant systems, but also facilitates the optimization and

  13. Quantitative assessment of probability of failing safely for the safety instrumented system using reliability block diagram method

    International Nuclear Information System (INIS)

    Jin, Jianghong; Pang, Lei; Zhao, Shoutang; Hu, Bin

    2015-01-01

    Highlights: • Models of PFS for SIS were established by using the reliability block diagram. • The more accurate calculation of PFS for SIS can be acquired by using SL. • Degraded operation of complex SIS does not affect the availability of SIS. • The safe undetected failure is the largest contribution to the PFS of SIS. - Abstract: The spurious trip of safety instrumented system (SIS) brings great economic losses to production. How to ensure the safety instrumented system is reliable and available has been put on the schedule. But the existing models on spurious trip rate (STR) or probability of failing safely (PFS) are too simplified and not accurate, in-depth studies of availability to obtain more accurate PFS for SIS are required. Based on the analysis of factors that influence the PFS for the SIS, using reliability block diagram method (RBD), the quantitative study of PFS for the SIS is carried out, and gives some application examples. The results show that, the common cause failure will increase the PFS; degraded operation does not affect the availability of the SIS; if the equipment was tested and repaired one by one, the unavailability of the SIS can be ignored; the corresponding occurrence time of independent safe undetected failure should be the system lifecycle (SL) rather than the proof test interval and the independent safe undetected failure is the largest contribution to the PFS for the SIS

  14. 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.

  15. 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

  16. Improvement of standards on functional reliability of electric power systems

    International Nuclear Information System (INIS)

    Barinov, V.A.; Volkov, G.A.; Kalita, V.V.; Kogan, F.L.; Makarov, S.F.; Manevich, A.S.; Mogirev, V.V.; Sin'chugov, F.I.; Skopintsev, V.A.; Khvoshchinskaya, Z.G.

    1993-01-01

    Analysis of the most principal aspects of the existing standards and requirements on assuring safety and stability of electric power systems (EPS) and effective (reliable and economical) power supply of consumers is given. The reliability is determined as ability to accomplish the assigned functions. Basic recommendations on improving the standards regulating the safety and reliability of the NPP functioning are formulated

  17. 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

  18. 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

  19. Numerical methods for reliability and safety assessment multiscale and multiphysics systems

    CERN Document Server

    Hami, Abdelkhalak

    2015-01-01

    This book offers unique insight on structural safety and reliability by combining computational methods that address multiphysics problems, involving multiple equations describing different physical phenomena, and multiscale problems, involving discrete sub-problems that together  describe important aspects of a system at multiple scales. The book examines a range of engineering domains and problems using dynamic analysis, nonlinear methods, error estimation, finite element analysis, and other computational techniques. This book also: ·       Introduces novel numerical methods ·       Illustrates new practical applications ·       Examines recent engineering applications ·       Presents up-to-date theoretical results ·       Offers perspective relevant to a wide audience, including teaching faculty/graduate students, researchers, and practicing engineers

  20. Reliability analysis of shutdown system

    International Nuclear Information System (INIS)

    Kumar, C. Senthil; John Arul, A.; Pal Singh, Om; Suryaprakasa Rao, K.

    2005-01-01

    This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 x 10 -8 /de for failure of shutdown function in case of global faults and 4.4 x 10 -8 /de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 x 10 -6 /ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is -3 /ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability

  1. Developing safety performance functions incorporating reliability-based risk measures.

    Science.gov (United States)

    Ibrahim, Shewkar El-Bassiouni; Sayed, Tarek

    2011-11-01

    Current geometric design guides provide deterministic standards where the safety margin of the design output is generally unknown and there is little knowledge of the safety implications of deviating from these standards. Several studies have advocated probabilistic geometric design where reliability analysis can be used to account for the uncertainty in the design parameters and to provide a risk measure of the implication of deviation from design standards. However, there is currently no link between measures of design reliability and the quantification of safety using collision frequency. The analysis presented in this paper attempts to bridge this gap by incorporating a reliability-based quantitative risk measure such as the probability of non-compliance (P(nc)) in safety performance functions (SPFs). Establishing this link will allow admitting reliability-based design into traditional benefit-cost analysis and should lead to a wider application of the reliability technique in road design. The present application is concerned with the design of horizontal curves, where the limit state function is defined in terms of the available (supply) and stopping (demand) sight distances. A comprehensive collision and geometric design database of two-lane rural highways is used to investigate the effect of the probability of non-compliance on safety. The reliability analysis was carried out using the First Order Reliability Method (FORM). Two Negative Binomial (NB) SPFs were developed to compare models with and without the reliability-based risk measures. It was found that models incorporating the P(nc) provided a better fit to the data set than the traditional (without risk) NB SPFs for total, injury and fatality (I+F) and property damage only (PDO) collisions. Copyright © 2011 Elsevier Ltd. All rights reserved.

  2. Discrete event simulation versus conventional system reliability analysis approaches

    DEFF Research Database (Denmark)

    Kozine, Igor

    2010-01-01

    Discrete Event Simulation (DES) environments are rapidly developing and appear to be promising tools for building reliability and risk analysis models of safety-critical systems and human operators. If properly developed, they are an alternative to the conventional human reliability analysis models...... and systems analysis methods such as fault and event trees and Bayesian networks. As one part, the paper describes briefly the author’s experience in applying DES models to the analysis of safety-critical systems in different domains. The other part of the paper is devoted to comparing conventional approaches...

  3. Strategy to safety grade systems replacements

    International Nuclear Information System (INIS)

    Stimler, M.; Sullivan, K.E.; Trebincevic, I.

    1993-01-01

    The introduction of digital instrumentation and control systems in nuclear power plants is characterized by the need to satisfy the requirements of safety, reliability and man-machine ergonomics. Today digital instrumentation and control systems meet these requirements and the trend in Europe is towards full digital based nuclear power plant control systems. This paper describes Siemens (KWU) experience in nuclear power plants and development in trends within Europe. Topics which are the subject of major concern to NPP operators addressed in this paper are: human performance factors - man-machine interface; operating philosophy; safety, availability and reliability. Other aspects addressed are: Siemens open-quotes defense in depthclose quotes concept, description of Siemens digital I ampersand C systems, safety requirements and systems, I ampersand C qualification, control room ergonomics, information systems and retrofitting experience

  4. 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.

  5. Safety and reliability analysis based on nonprobabilistic methods

    International Nuclear Information System (INIS)

    Kozin, I.O.; Petersen, K.E.

    1996-01-01

    Imprecise probabilities, being developed during the last two decades, offer a considerably more general theory having many advantages which make it very promising for reliability and safety analysis. The objective of the paper is to argue that imprecise probabilities are more appropriate tool for reliability and safety analysis, that they allow to model the behavior of nuclear industry objects more comprehensively and give a possibility to solve some problems unsolved in the framework of conventional approach. Furthermore, some specific examples are given from which we can see the usefulness of the tool for solving some reliability tasks

  6. 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)

  7. Guidelines for implementation of RCM on safety systems

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Brijendra Singh.

    1996-04-01

    Reliability Centered Maintenance (RCM) methodology was originally developed by the commercial airlines industry in the early 1960s for identifying applicable and effective preventive maintenance tasks and as currently used in nuclear power industry. Effective maintenance of the systems at a nuclear power plant (NPP) is essential for its safe and reliable operation. Reliability Centered Maintenance at NPP is the program to assure that plant systems remain within an original design criteria and are not adversely affected during the plant life time. The aim of this report is to provide the guidelines to implement the RCM approach on NPP safety systems. Safety systems are usually standby and therefore, we need to periodically detect and repair failures that may have occurred since the previous activation or inspection the equipment. The RCM guidelines are intended to help identify the failure modes and related root causes and then decide the maintenance policies to achieve the high level of safety and reliability. The RCM is intended to improve or maintain high levels of system reliability and plant availability. Since the reliability of plant systems will be improved, the plant safety correspondingly will be increased. Another goal of RCM is to optimize the maintenance and surveillance tasks such that the overall level of resources required to accomplish essential tasks is kept to minimum. RCM also strives to eliminate unnecessary corrective maintenance and to select yet most cost-effective approach to maintenance, testing and inspection for system components. 9 refs. (Author) .new

  8. 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....

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  10. Sensitivity evaluation of human factors for reliability of the containment spray system

    International Nuclear Information System (INIS)

    Tsujimura, Yasuhiro; Suzuki, Eiji

    1988-01-01

    Evaluation of the human reliability is one of the most difficult problems that deal with the safety and reliability of large systems, especially of the Engineered Safety Features (ESF) of the nuclear power plant. Influences of human factors on the reliability of the Containment Spray System in the ESF were estimated by using the FTA method in this paper. As a result, the adequacy of the system structure and the effects of human factors on variations of the design of the system structure were explained. (author)

  11. Summary of the preparation of methodology for digital system reliability analysis for PSA purposes

    International Nuclear Information System (INIS)

    Hustak, S.; Babic, P.

    2001-12-01

    The report is structured as follows: Specific features of and requirements for the digital part of NPP Instrumentation and Control (I and C) systems (Computer-controlled digital technologies and systems of the NPP I and C system; Specific types of digital technology failures and preventive provisions; Reliability requirements for the digital parts of I and C systems; Safety requirements for the digital parts of I and C systems; Defence-in-depth). Qualitative analyses of NPP I and C system reliability and safety (Introductory system analysis; Qualitative requirements for and proof of NPP I and C system reliability and safety). Quantitative reliability analyses of the digital parts of I and C systems (Selection of a suitable quantitative measure of digital system reliability; Selected qualitative and quantitative findings regarding digital system reliability; Use of relations among the occurrences of the various types of failure). Mathematical section in support of the calculation of the various types of indices (Boolean reliability models, Markovian reliability models). Example of digital system analysis (Description of a selected protective function and the relevant digital part of the I and C system; Functional chain examined, its components and fault tree). (P.A.)

  12. Diagnostics and reliability of pipeline systems

    CERN Document Server

    Timashev, Sviatoslav

    2016-01-01

    The book contains solutions to fundamental problems which arise due to the logic of development of specific branches of science, which are related to pipeline safety, but mainly are subordinate to the needs of pipeline transportation.          The book deploys important but not yet solved aspects of reliability and safety assurance of pipeline systems, which are vital aspects not only for the oil and gas industry and, in general, fuel and energy industries , but also to virtually all contemporary industries and technologies. The volume will be useful to specialists and experts in the field of diagnostics/ inspection, monitoring, reliability and safety of critical infrastructures. First and foremost, it will be useful to the decision making persons —operators of different types of pipelines, pipeline diagnostics/inspection vendors, and designers of in-line –inspection (ILI) tools, industrial and ecological safety specialists, as well as to researchers and graduate students.

  13. 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.)

  14. Reliability evaluation of a natural circulation system

    International Nuclear Information System (INIS)

    Jafari, Jalil; D'Auria, Francesco; Kazeminejad, Hossein; Davilu, Hadi

    2003-01-01

    This paper discusses a reliability study performed with reference to a passive thermohydraulic natural circulation (NC) system, named TTL-1. A methodology based on probabilistic techniques has been applied with the main purpose to optimize the system design. The obtained results have been adopted to estimate the thermal-hydraulic reliability (TH-R) of the same system. A total of 29 relevant parameters (including nominal values and plausible ranges of variations) affecting the design and the NC performance of the TTL-1 loop are identified and a probability of occurrence is assigned for each value based on expert judgment. Following procedures established for the uncertainty evaluation of thermal-hydraulic system codes results, 137 system configurations have been selected and each configuration has been analyzed via the Relap5 best-estimate code. The reference system configuration and the failure criteria derived from the 'mission' of the passive system are adopted for the evaluation of the system TH-R. Four different definitions of a less-than-unity 'reliability-values' (where unity represents the maximum achievable reliability) are proposed for the performance of the selected passive system. This is normally considered fully reliable, i.e. reliability-value equal one, in typical Probabilistic Safety Assessment (PSA) applications in nuclear reactor safety. The two 'point' TH-R values for the considered NC system were found equal to 0.70 and 0.85, i.e. values comparable with the reliability of a pump installed in an 'equivalent' forced circulation (active) system having the same 'mission'. The design optimization study was completed by a regression analysis addressing the output of the 137 calculations: heat losses, undetected leakage, loop length, riser diameter, and equivalent diameter of the test section have been found as the most important parameters bringing to the optimal system design and affecting the TH-R. As added values for this work, the comparison has

  15. Safety regulations concerning instrumentation and control systems for research reactors

    International Nuclear Information System (INIS)

    El-Shanshoury, A.I.

    2009-01-01

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

  16. 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)

  17. 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

  18. Field reliability of electronic systems

    International Nuclear Information System (INIS)

    Elm, T.

    1984-02-01

    This report investigates, through several examples from the field, the reliability of electronic units in a broader sense. That is, it treats not just random parts failure, but also inadequate reliability design and (externally and internally) induced failures. The report is not meant to be merely an indication of the state of the art for the reliability prediction methods we know, but also as a contribution to the investigation of man-machine interplay in the operation and repair of electronic equipment. The report firmly links electronics reliability to safety and risk analyses approaches with a broader, system oriented view of reliability prediction and with postfailure stress analysis. It is intended to reveal, in a qualitative manner, the existence of symptom and cause patterns. It provides a background for further investigations to identify the detailed mechanisms of the faults and the remedical actions and precautions for achieving cost effective reliability. (author)

  19. [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.

  20. Safety design requirements for safety systems and components of JSFR

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

    2011-01-01

    Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

  1. An artificial intelligence system for reliability studies

    International Nuclear Information System (INIS)

    Llory, M.; Ancelin, C.; Bannelier, M.; Bouhadana, H.; Bouissou, M.; Lucas, J.Y.; Magne, L.; Villate, N.

    1990-01-01

    The EDF (French Electricity Company) software developed for computer aided reliability studies is considered. Such software tools were applied in the study of the safety requirements of the Paluel nuclear power plant. The reliability models, based on IF-THEN type rules, and the generation of models by the expert system are described. The models are then processed applying algorithm structures [fr

  2. Optimization of maintenance periodicity of complex of NPP safety systems

    International Nuclear Information System (INIS)

    Kolykhanov, V.; Skalozubov, V.; Kovrigkin, Y.

    2006-01-01

    The analysis of the positive and negative aspects connected to maintenance of the safety systems equipment which basically is in a standby state is executed. Tests of systems provide elimination of the latent failures and raise their reliability. Poor quality of carrying out the tests can be a source of the subsequent failures. Therefore excess frequency of tests can result in reducing reliability of safety systems. The method of optimization of maintenance periodicity of the equipment taking into account factors of its reliability and restoration procedures quality is submitted. The unavailability factor is used as a criterion of optimization of maintenance periodicity. It is offered to use parameters of reliability of the equipment and each of safety systems of NPPs received at developing PSA. And it is offered to carry out the concordance of maintenance periodicity of systems within the NPP maintenance program taking into account a significance factor of the system received on the basis of the contribution of system in CDF. Basing on the submitted method the small computer code is developed. This code allows to calculate reliability factors of a separate safety system and to determine optimum maintenance periodicity of its equipment. Optimization of maintenance periodicity of a complex of safety systems is stipulated also. As an example results of optimization of maintenance periodicity at Zaporizhzhya NPP are presented. (author)

  3. 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

  4. Research on the evaluation model of the software reliability in nuclear safety class digital instrumentation and control system

    International Nuclear Information System (INIS)

    Liu Ying; Yang Ming; Li Fengjun; Ma Zhanguo; Zeng Hai

    2014-01-01

    In order to analyze the software reliability (SR) in nuclear safety class digital instrumentation and control system (D-I and C), firstly, the international software design standards were analyzed, the standards' framework was built, and we found that the D-I and C software standards should follow the NUREG-0800 BTP7-14, according to the NRC NUREG-0800 review of requirements. Secondly, the quantitative evaluation model of SR using Bayesian Belief Network and thirteen sub-model frameworks were established. Thirdly, each sub-models and the weight of corresponding indexes in the evaluation model were analyzed. Finally, the safety case was introduced. The models lay a foundation for review and quantitative evaluation on the SR in nuclear safety class D-I and C. (authors)

  5. Development in structural systems reliability theory

    Energy Technology Data Exchange (ETDEWEB)

    Murotsu, Y

    1986-07-01

    This paper is concerned with two topics on structural systems reliability theory. One covers automatic generation of failure mode equations, identifications of stochastically dominant failure modes, and reliability assessment of redundant structures. Reduced stiffness matrixes and equivalent nodal forces representing the failed elements are introduced for expressing the safety of the elements, using a matrix method. Dominant failure modes are systematically selected by a branch-and-bound technique and heuristic operations. The other discusses the various optimum design problems based on reliability concept. Those problems are interpreted through a solution to a multi-objective optimization problem.

  6. Development in structural systems reliability theory

    International Nuclear Information System (INIS)

    Murotsu, Y.

    1986-01-01

    This paper is concerned with two topics on structural systems reliability theory. One covers automatic generation of failure mode equations, identifications of stochastically dominant failure modes, and reliability assessment of redundant structures. Reduced stiffness matrixes and equivalent nodal forces representing the failed elements are introduced for expressing the safety of the elements, using a matrix method. Dominant failure modes are systematically selected by a branch-and-bound technique and heuristic operations. The other discusses the various optimum design problems based on reliability concept. Those problems are interpreted through a solution to a multi-objective optimization problem. (orig.)

  7. Reliability Estimation for Digital Instrument/Control System

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yaguang; Sydnor, Russell [U.S. Nuclear Regulatory Commission, Washington, D.C. (United States)

    2011-08-15

    Digital instrumentation and controls (DI and C) systems are widely adopted in various industries because of their flexibility and ability to implement various functions that can be used to automatically monitor, analyze, and control complicated systems. It is anticipated that the DI and C will replace the traditional analog instrumentation and controls (AI and C) systems in all future nuclear reactor designs. There is an increasing interest for reliability and risk analyses for safety critical DI and C systems in regulatory organizations, such as The United States Nuclear Regulatory Commission. Developing reliability models and reliability estimation methods for digital reactor control and protection systems will involve every part of the DI and C system, such as sensors, signal conditioning and processing components, transmission lines and digital communication systems, D/A and A/D converters, computer system, signal processing software, control and protection software, power supply system, and actuators. Some of these components are hardware, such as sensors and actuators, their failure mechanisms are well understood, and the traditional reliability model and estimation methods can be directly applied. But many of these components are firmware which has software embedded in the hardware, and software needs special consideration because its failure mechanism is unique, and the reliability estimation method for a software system will be different from the ones used for hardware systems. In this paper, we will propose a reliability estimation method for the entire DI and C system reliability using a recently developed software reliability estimation method and a traditional hardware reliability estimation method.

  8. Reliability Estimation for Digital Instrument/Control System

    International Nuclear Information System (INIS)

    Yang, Yaguang; Sydnor, Russell

    2011-01-01

    Digital instrumentation and controls (DI and C) systems are widely adopted in various industries because of their flexibility and ability to implement various functions that can be used to automatically monitor, analyze, and control complicated systems. It is anticipated that the DI and C will replace the traditional analog instrumentation and controls (AI and C) systems in all future nuclear reactor designs. There is an increasing interest for reliability and risk analyses for safety critical DI and C systems in regulatory organizations, such as The United States Nuclear Regulatory Commission. Developing reliability models and reliability estimation methods for digital reactor control and protection systems will involve every part of the DI and C system, such as sensors, signal conditioning and processing components, transmission lines and digital communication systems, D/A and A/D converters, computer system, signal processing software, control and protection software, power supply system, and actuators. Some of these components are hardware, such as sensors and actuators, their failure mechanisms are well understood, and the traditional reliability model and estimation methods can be directly applied. But many of these components are firmware which has software embedded in the hardware, and software needs special consideration because its failure mechanism is unique, and the reliability estimation method for a software system will be different from the ones used for hardware systems. In this paper, we will propose a reliability estimation method for the entire DI and C system reliability using a recently developed software reliability estimation method and a traditional hardware reliability estimation method

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

    International Nuclear Information System (INIS)

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

    1995-04-01

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

  10. Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Muhammad, E-mail: hashimsajid@yahoo.com; Hidekazu, Yoshikawa, E-mail: yosikawa@kib.biglobe.ne.jp; Takeshi, Matsuoka, E-mail: mats@cc.utsunomiya-u.ac.jp; Ming, Yang, E-mail: myang.heu@gmail.com

    2014-10-15

    Highlights: • Discussion on reasons why AP1000 equipped with ADS system comparatively to PWR. • Clarification of full and partial depressurization of reactor coolant system by ADS system. • Application case study of four stages ADS system for reliability evaluation in LBLOCA. • GO-FLOW tool is capable to evaluate dynamic reliability of passive safety systems. • Calculated ADS reliability result significantly increased dynamic reliability of PXS. - Abstract: AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs. In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated

  11. Reliability engineering for nuclear and other high technology systems

    International Nuclear Information System (INIS)

    Lakner, A.A.; Anderson, R.T.

    1985-01-01

    This book is written for the reliability instructor, program manager, system engineer, design engineer, reliability engineer, nuclear regulator, probability risk assessment (PRA) analyst, general manager and others who are involved in system hardware acquisition, design and operation and are concerned with plant safety and operational cost-effectiveness. It provides criteria, guidelines and comprehensive engineering data affecting reliability; it covers the key aspects of system reliability as it relates to conceptual planning, cost tradeoff decisions, specification, contractor selection, design, test and plant acceptance and operation. It treats reliability as an integrated methodology, explicitly describing life cycle management techniques as well as the basic elements of a total hardware development program, including: reliability parameters and design improvement attributes, reliability testing, reliability engineering and control. It describes how these elements can be defined during procurement, and implemented during design and development to yield reliable equipment. (author)

  12. Mass and Reliability System (MaRS)

    Science.gov (United States)

    Barnes, Sarah

    2016-01-01

    The Safety and Mission Assurance (S&MA) Directorate is responsible for mitigating risk, providing system safety, and lowering risk for space programs from ground to space. The S&MA is divided into 4 divisions: The Space Exploration Division (NC), the International Space Station Division (NE), the Safety & Test Operations Division (NS), and the Quality and Flight Equipment Division (NT). The interns, myself and Arun Aruljothi, will be working with the Risk & Reliability Analysis Branch under the NC Division's. The mission of this division is to identify, characterize, diminish, and communicate risk by implementing an efficient and effective assurance model. The team utilizes Reliability and Maintainability (R&M) and Probabilistic Risk Assessment (PRA) to ensure decisions concerning risks are informed, vehicles are safe and reliable, and program/project requirements are realistic and realized. This project pertains to the Orion mission, so it is geared toward a long duration Human Space Flight Program(s). For space missions, payload is a critical concept; balancing what hardware can be replaced by components verse by Orbital Replacement Units (ORU) or subassemblies is key. For this effort a database was created that combines mass and reliability data, called Mass and Reliability System or MaRS. The U.S. International Space Station (ISS) components are used as reference parts in the MaRS database. Using ISS components as a platform is beneficial because of the historical context and the environment similarities to a space flight mission. MaRS uses a combination of systems: International Space Station PART for failure data, Vehicle Master Database (VMDB) for ORU & components, Maintenance & Analysis Data Set (MADS) for operation hours and other pertinent data, & Hardware History Retrieval System (HHRS) for unit weights. MaRS is populated using a Visual Basic Application. Once populated, the excel spreadsheet is comprised of information on ISS components including

  13. System reliability time-dependent models

    International Nuclear Information System (INIS)

    Debernardo, H.D.

    1991-06-01

    A probabilistic methodology for safety system technical specification evaluation was developed. The method for Surveillance Test Interval (S.T.I.) evaluation basically means an optimization of S.T.I. of most important system's periodically tested components. For Allowed Outage Time (A.O.T.) calculations, the method uses system reliability time-dependent models (A computer code called FRANTIC III). A new approximation, which was called Independent Minimal Cut Sets (A.C.I.), to compute system unavailability was also developed. This approximation is better than Rare Event Approximation (A.E.R.) and the extra computing cost is neglectible. A.C.I. was joined to FRANTIC III to replace A.E.R. on future applications. The case study evaluations verified that this methodology provides a useful probabilistic assessment of surveillance test intervals and allowed outage times for many plant components. The studied system is a typical configuration of nuclear power plant safety systems (two of three logic). Because of the good results, these procedures will be used by the Argentine nuclear regulatory authorities in evaluation of technical specification of Atucha I and Embalse nuclear power plant safety systems. (Author) [es

  14. 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

  15. Reliability analysis of service water system under earthquake

    International Nuclear Information System (INIS)

    Yu Yu; Qian Xiaoming; Lu Xuefeng; Wang Shengfei; Niu Fenglei

    2013-01-01

    Service water system is one of the important safety systems in nuclear power plant, whose failure probability is always gained by system reliability analysis. The probability of equipment failure under the earthquake is the function of the peak acceleration of earthquake motion, while the occurrence of earthquake is of randomicity, thus the traditional fault tree method in current probability safety assessment is not powerful enough to deal with such case of conditional probability problem. An analysis frame was put forward for system reliability evaluation in seismic condition in this paper, in which Monte Carlo simulation was used to deal with conditional probability problem. Annual failure probability of service water system was calculated, and failure probability of 1.46X10 -4 per year was obtained. The analysis result is in accordance with the data which indicate equipment seismic resistance capability, and the rationality of the model is validated. (authors)

  16. The reliability of the software of the digital control system Nuclear Advantage

    International Nuclear Information System (INIS)

    Graae, T.; Engdahl, L.

    1996-01-01

    The ABB nuclear power control system Nuclear Advantage is a truly integrated control system. The integration of process control and safety control aims at achieving a common operator interface in order to simplify and thus improve control room ergonomics. The challenge is to design an integrated control system and at the same time ensure the functional separation between the independent safety subsystems as well as between the safety and the conventional sections. Software reliability is discussed and illustrated by statistical test results. It has proved to be a hundred times better than the reliability of the high-quality hardware. (orig.) [de

  17. Integrating software reliability concepts into risk and reliability modeling of digital instrumentation and control systems used in nuclear power plants

    International Nuclear Information System (INIS)

    Arndt, S. A.

    2006-01-01

    As software-based digital systems are becoming more and more common in all aspects of industrial process control, including the nuclear power industry, it is vital that the current state of the art in quality, reliability, and safety analysis be advanced to support the quantitative review of these systems. Several research groups throughout the world are working on the development and assessment of software-based digital system reliability methods and their applications in the nuclear power, aerospace, transportation, and defense industries. However, these groups are hampered by the fact that software experts and probabilistic safety assessment experts view reliability engineering very differently. This paper discusses the characteristics of a common vocabulary and modeling framework. (authors)

  18. 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.

  19. Analysis of fault tolerance and reliability in distributed real-time system architectures

    International Nuclear Information System (INIS)

    Philippi, Stephan

    2003-01-01

    Safety critical real-time systems are becoming ubiquitous in many areas of our everyday life. Failures of such systems potentially have catastrophic consequences on different scales, in the worst case even the loss of human life. Therefore, safety critical systems have to meet maximum fault tolerance and reliability requirements. As the design of such systems is far from being trivial, this article focuses on concepts to specifically support the early architectural design. In detail, a simulation based approach for the analysis of fault tolerance and reliability in distributed real-time system architectures is presented. With this approach, safety related features can be evaluated in the early development stages and thus prevent costly redesigns in later ones

  20. Methodologies of the hardware reliability prediction for PSA of digital I and C systems

    International Nuclear Information System (INIS)

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

    2000-09-01

    Digital I and C systems are being used widely in the Non-safety systems of the NPP and they are expanding their applications to safety critical systems. The regulatory body shifts their policy to risk based and may require Probabilistic Safety Assessment for the digital I and C systems. But there is no established reliability prediction methodology for the digital I and C systems including both software and hardware yet. This survey report includes a lot of reliability prediction methods for electronic systems in view of hardware. Each method has both the strong and the weak points. This report provides the state-of-art of prediction methods and focus on Bellcore method and MIL-HDBK-217F method in deeply. The reliability analysis models are reviewed and discussed to help analysts. Also this report includes state-of-art of software tools that are supporting reliability prediction

  1. Methodologies of the hardware reliability prediction for PSA of digital I and C systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-01

    Digital I and C systems are being used widely in the Non-safety systems of the NPP and they are expanding their applications to safety critical systems. The regulatory body shifts their policy to risk based and may require Probabilistic Safety Assessment for the digital I and C systems. But there is no established reliability prediction methodology for the digital I and C systems including both software and hardware yet. This survey report includes a lot of reliability prediction methods for electronic systems in view of hardware. Each method has both the strong and the weak points. This report provides the state-of-art of prediction methods and focus on Bellcore method and MIL-HDBK-217F method in deeply. The reliability analysis models are reviewed and discussed to help analysts. Also this report includes state-of-art of software tools that are supporting reliability prediction.

  2. Reliability and validity of emergency department triage systems

    NARCIS (Netherlands)

    van der Wulp, I.

    2010-01-01

    Reliability and validity of triage systems is important because this can affect patient safety. In this thesis, these aspects of two emergency department (ED) triage systems were studied as well as methodological aspects in these types of studies. The consistency, reproducibility, and criterion

  3. Failure and Reliability Analysis for the Master Pump Shutdown System

    International Nuclear Information System (INIS)

    BEVINS, R.R.

    2000-01-01

    The Master Pump Shutdown System (MPSS) will be installed in the 200 Areas of the Hanford Site to monitor and control the transfer of liquid waste between tank farms and between the 200 West and 200 East areas through the Cross-Site Transfer Line. The Safety Function provided by the MPSS is to shutdown any waste transfer process within or between tank farms if a waste leak should occur along the selected transfer route. The MPSS, which provides this Safety Class Function, is composed of Programmable Logic Controllers (PLCs), interconnecting wires, relays, Human to Machine Interfaces (HMI), and software. These components are defined as providing a Safety Class Function and will be designated in this report as MPSS/PLC. Input signals to the MPSS/PLC are provided by leak detection systems from each of the tank farm leak detector locations along the waste transfer route. The combination of the MPSS/PLC, leak detection system, and transfer pump controller system will be referred to as MPSS/SYS. The components addressed in this analysis are associated with the MPSS/SYS. The purpose of this failure and reliability analysis is to address the following design issues of the Project Development Specification (PDS) for the MPSS/SYS (HNF 2000a): (1) Single Component Failure Criterion, (2) System Status Upon Loss of Electrical Power, (3) Physical Separation of Safety Class cables, (4) Physical Isolation of Safety Class Wiring from General Service Wiring, and (5) Meeting the MPSS/PLC Option 1b (RPP 1999) Reliability estimate. The failure and reliability analysis examined the system on a component level basis and identified any hardware or software elements that could fail and/or prevent the system from performing its intended safety function

  4. 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

  5. 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.

  6. RAVONSICS-challenging for assuring software reliability of nuclear I and C system

    International Nuclear Information System (INIS)

    Hai Zeng; Ming Yang; Yoshikawa, Hidekazu

    2015-01-01

    As the “central nerve system”, the highly reliable Instrumentation and Control (I and C) systems, which provide the right functions and functions correctly, are always desirable not only for the end users of NPPs but also the suppliers of I and C systems. The Digitalization of nuclear I and C system happened in recent years brought a lot of new features for nuclear I and C system. On one side digital technology provides more functionalities, and it should be more reliable and robust; on the other side, digital technology brings new challenge for nuclear I and C system, especially the software running in the hardware component. The software provides flexible functionalities for nuclear I and C system, but it also brings the difficulties to evaluate the reliability and safety of it because of the complexity of software. The reliability of software, which is indispensable part of I and C system, will have essential impact on the reliability of the whole system, and people definitely want to know what the reliability of this intangible part is. The methods used for the evaluation of reliability of system and hardware hardly work for software, because the inherent difference of failure mechanism exists between software and hardware. Failure in software is systematically induced by design error, but failure in hardware is randomly induced by material and production. To continue the effort on this hot topic and to try to achieve consensus on the potential methodology for software reliability evaluation, a cooperative research project called RAVONSICS (Reliability and Verification and Validation of Nuclear Safety I and C Software) is being carried on by 7 Chinese partners, which includes University, research institute, utility, vendor, and safety regulatory body. The objective of RAVONSICS is to bring forwards the methodology for the software reliability evaluation, and the software verification technique. RAVONSICS works cooperatively with its European sister project

  7. Development of digital safety system logic and control

    International Nuclear Information System (INIS)

    Nishikawa, H.; Sakamoto, H.

    1995-01-01

    Advanced-BWR (ABWR) uses total digital control and instrumentation (C and I) system. In particular, ABWR adopts a newly developed safety system using advanced digital technology. In the presentation the digital safety system design, manufacturing and factory validation test method are shortly overviewed. The digital safety system consists of micro-processor based digital controllers, data and information transmission by optical fibers and human-machine interface using color flat displays. This new developed safety system meet the nuclear safety requirements such as high reliability, independence of divisions, operability and maintainability. (2 refs., 4 figs., 1 tab.)

  8. System safety and reliability using object-oriented programming techniques

    International Nuclear Information System (INIS)

    Patterson-Hine, F.A.; Koen, B.V.

    1987-01-01

    Direct evaluation fault tree codes have been written in recursive, list-processing computer languages such as PL/1 (PATREC-I) and LISP (PATREC-L). The pattern-matching strategy implemented in these codes has been used extensively in France to evaluate system reliability. Recent reviews of the risk management process suggest that a data base containing plant-specific information be integrated with a package of codes used for probabilistic risk assessment (PRA) to alleviate some of the difficulties that make a PRA so costly and time-intensive. A new programming paradigm, object-oriented programming, is uniquely suited for the development of such a software system. A knowledge base and fault tree evaluation algorithm, based on previous experience with PATREC-L, have been implemented using object-oriented techniques, resulting in a reliability assessment environment that is easy to develop, modify, and extend

  9. Water chemistry data acquisition, processing, evaluation and diagnostic systems in Light Water Reactors: Future improvement of plant reliability and safety

    International Nuclear Information System (INIS)

    Uchida, S.; Takiguchi, H.; Ishigure, K.

    2006-01-01

    Data acquisition, processing and evaluation systems have been applied in major Japanese PWRs and BWRs to provide (1) reliable and quick data acquisition with manpower savings in plant chemical laboratories and (2) smooth and reliable information transfer among chemists, plant operators, and supervisors. Data acquisition systems in plants consist of automatic and semi-automatic instruments for chemical analyses, e. g., X-ray fluorescence analysis and ion chromatography, while data processing systems consist of PC base-sub-systems, e.g., data storage, reliability evaluation, clear display, and document preparation for understanding the plant own water chemistry trends. Precise and reliable evaluations of water chemistry data are required in order to improve plant reliability and safety. For this, quality assurance of the water chemistry data acquisition system is needed. At the same time, theoretical models are being applied to bridge the gaps between measured water chemistry data and the information desired to understand the interaction of materials and cooling water in plants. Major models which have already been applied for plant evaluation are: (1) water radiolysis models for BWRs and PWRs; (2) crevice radiolysis model for SCC in BWRs; and (3) crevice pH model for SG tubing in PWRs. High temperature water chemistry sensors and automatic plant diagnostic systems have been applied in only restricted areas. ECP sensors are gaining popularity as tools to determine the effects of hydrogen injection in BWR systems. Automatic plant diagnostic systems based on artificial intelligence will be more popular after having sufficient experience with off line diagnostic systems. (author)

  10. Design requirements of communication architecture of SMART safety system

    International Nuclear Information System (INIS)

    Park, H. Y.; Kim, D. H.; Sin, Y. C.; Lee, J. Y.

    2001-01-01

    To develop the communication network architecture of safety system of SMART, the evaluation elements for reliability and performance factors are extracted from commercial networks and classified the required-level by importance. A predictable determinacy, status and fixed based architecture, separation and isolation from other systems, high reliability, verification and validation are introduced as the essential requirements of safety system communication network. Based on the suggested requirements, optical cable, star topology, synchronous transmission, point-to-point physical link, connection-oriented logical link, MAC (medium access control) with fixed allocation are selected as the design elements. The proposed architecture will be applied as basic communication network architecture of SMART safety system

  11. 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)

  12. Safety of huge systems

    International Nuclear Information System (INIS)

    Kondo, Jiro.

    1995-01-01

    Recently accompanying the development of engineering technology, huge systems tend to be constructed. The disaster countermeasures of huge cities become large problems as the concentration of population into cities is conspicuous. To make the expected value of loss small, the knowledge of reliability engineering is applied. In reliability engineering, even if a part of structures fails, the safety as a whole system must be ensured, therefore, the design having margin is carried out. The degree of margin is called redundancy. However, such design concept makes the structure of a system complex, and as the structure is complex, the possibility of causing human errors becomes high. At the time of huge system design, the concept of fail-safe is effective, but simple design must be kept in mind. The accident in Mihama No. 2 plant of Kansai Electric Power Co. and the accident in Chernobyl nuclear power station, and the accident of Boeing B737 airliner and the fatigue breakdown are described. The importance of safety culture was emphasized as the method of preventing human errors. Man-system interface and management system are discussed. (K.I.)

  13. Preparation of methodology for reliability analysis of selected digital segments of the instrumentation and control systems of NPPs. Pt. 1

    International Nuclear Information System (INIS)

    Hustak, S.; Patrik, M.; Babic, P.

    2000-12-01

    The report is structured as follows: (i) Introduction; (ii) Important notions relating to the safety and dependability of software systems for nuclear power plants (selected notions from IAEA Technical Report No. 397; safety aspects of software application; reliability/dependability aspects of digital systems); (iii) Peculiarities of digital systems and ways to a dependable performance of the required function (failures in the system and principles of defence against them; ensuring resistance of digital systems against failures at various hardware and software levels); (iv) The issue of analytical procedures to assess the safety and reliability of safety-related digital systems (safety and reliability assessment at an early stage of the project; general framework of reliability analysis of complex systems; choice of an appropriate quantitative measure of software reliability); (v) Selected qualitative and quantitative information about the reliability of digital systems; the use of relations between the incidence of various types of faults); and (vi) Conclusions and recommendations. (P.A.)

  14. COMPRESS - a computerized reactor safety system

    International Nuclear Information System (INIS)

    Vegh, E.

    1986-01-01

    The computerized reactor safety system, called COMPRESS, provides the following services: scram initiation; safety interlockings; event recording. The paper describes the architecture of the system and deals with reliability problems. A self-testing unit checks permanently the correct operation of the independent decision units. Moreover the decision units are tested by short pulses whether they can initiate a scram. The self-testing is described in detail

  15. Qualitative analysis in reliability and safety studies

    International Nuclear Information System (INIS)

    Worrell, R.B.; Burdick, G.R.

    1976-01-01

    The qualitative evaluation of system logic models is described as it pertains to assessing the reliability and safety characteristics of nuclear systems. Qualitative analysis of system logic models, i.e., models couched in an event (Boolean) algebra, is defined, and the advantages inherent in qualitative analysis are explained. Certain qualitative procedures that were developed as a part of fault-tree analysis are presented for illustration. Five fault-tree analysis computer-programs that contain a qualitative procedure for determining minimal cut sets are surveyed. For each program the minimal cut-set algorithm and limitations on its use are described. The recently developed common-cause analysis for studying the effect of common-causes of failure on system behavior is explained. This qualitative procedure does not require altering the fault tree, but does use minimal cut sets from the fault tree as part of its input. The method is applied using two different computer programs. 25 refs

  16. Importance of independent and dependent human error to system reliability and plant safety

    International Nuclear Information System (INIS)

    Dach, K.

    1988-08-01

    Uncertainty analysis of the quantification of the unavailability for the emergency core cooling system was made. The reliability analysis of the low pressure injection system (LPIS) of the ECCS of WWER-440 reactor was also performed. Results of reliability analysis proved that LPIS reliability under normal conditions is sufficient and can be increased by two orders of magnitude. This increase in reliability can be achieved by means of simple changes such as securing an opening of the quick-acting fittings at LPIS discharge line. A method for analysis of systems uncertainty with periodic inspected components was elaborated and verified by performing an analysis of the medium size system. Refs, figs and tabs

  17. Quantitative risk assessment of digitalized safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sung Min; Lee, Sang Hun; Kang, Hym Gook [KAIST, Daejeon (Korea, Republic of); Lee, Seung Jun [UNIST, Ulasn (Korea, Republic of)

    2016-05-15

    A report published by the U.S. National Research Council indicates that appropriate methods for assessing reliability are key to establishing the acceptability of digital instrumentation and control (I and C) systems in safety-critical plants such as NPPs. Since the release of this issue, the methodology for the probabilistic safety assessment (PSA) of digital I and C systems has been studied. However, there is still no widely accepted method. Kang and Sung found three critical factors for safety assessment of digital systems: detection coverage of fault-tolerant techniques, software reliability quantification, and network communication risk. In reality the various factors composing digitalized I and C systems are not independent of each other but rather closely connected. Thus, from a macro point of view, a method that can integrate risk factors with different characteristics needs to be considered together with the micro approaches to address the challenges facing each factor.

  18. 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

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

    The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

  7. Insights from a reliability review of digital plant protection system

    International Nuclear Information System (INIS)

    Kim, I.S.; Hwang, S.W.; Kim, B.S.; Jeong, C.H.; Oh, S.H.

    2001-01-01

    The full text follows: As part of the design efforts for Ulchin nuclear power plant units 5 and 6 of Korea, a reliability analysis of digital plant protection system (DPPS) was performed by ABB-CE. An independent review of the DPPS reliability analysis was undertaken by Hanyang University to assist Korea Institute of Nuclear Safety (KINS), the nuclear regulatory body of Korea, in evaluating the design acceptability of the digital system. The DPPS is designed to encompass both reactor trip function and ESFAS (engineered safety feature actuation system) initiation function. The major methods used by the ABB-CE to assess the Ulchin 5-6 DPPS reliability are failure mode and effect analysis (FMEA) and fault tree analysis. Hence, our independent review was conducted focusing on: -) the establishment of review criteria based on various sources, such as the standard review plan of KINS, 10CFR50 Appendix A, IEEE standards 279, 577, and 603; -) the suitability of the FMEA and fault tree analysis for the Ulchin 5-6 DPPS, including the specific methods used (e.g., for human reliability analysis and common-cause failure analysis), the assumptions made (e.g., with respect to test frequency and watchdog timer coverage), and the data employed (e.g., CCF parameter, human error probability, and processor failure rate); and -) the design acceptability of the DPPS especially as compared to the analog plant protection system from a reliability and safety perspective. The paper shall also discuss key issues requiring further in-depth investigation, such as reliability of programmable logic controllers (PLCs), coverage factor of watchdog timers, and susceptibility of redundant digital units to common cause failure. Sensitivity analyses were carried out to investigate the impact of several parameters of special interest, like the coverage factor of watchdog timer and human error probability (e.g. an operator error to manually trip the reactor, or to mis-calibrate the trip parameters) on

  8. Reliability assessment of nuclear structural systems

    International Nuclear Information System (INIS)

    Reich, M.; Hwang, H.

    1983-01-01

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

  9. Reliable software systems via chains of object models with provably correct behavior

    International Nuclear Information System (INIS)

    Yakhnis, A.; Yakhnis, V.

    1996-01-01

    This work addresses specification and design of reliable safety-critical systems, such as nuclear reactor control systems. Reliability concerns are addressed in complimentary fashion by different fields. Reliability engineers build software reliability models, etc. Safety engineers focus on prevention of potential harmful effects of systems on environment. Software/hardware correctness engineers focus on production of reliable systems on the basis of mathematical proofs. The authors think that correctness may be a crucial guiding issue in the development of reliable safety-critical systems. However, purely formal approaches are not adequate for the task, because they neglect the connection with the informal customer requirements. They alleviate that as follows. First, on the basis of the requirements, they build a model of the system interactions with the environment, where the system is viewed as a black box. They will provide foundations for automated tools which will (a) demonstrate to the customer that all of the scenarios of system behavior are presented in the model, (b) uncover scenarios not present in the requirements, and (c) uncover inconsistent scenarios. The developers will work with the customer until the black box model will not possess scenarios (b) and (c) above. Second, the authors will build a chain of several increasingly detailed models, where the first model is the black box model and the last model serves to automatically generated proved executable code. The behavior of each model will be proved to conform to the behavior of the previous one. They build each model as a cluster of interactive concurrent objects, thus they allow both top-down and bottom-up development

  10. 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.)

  11. 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

  12. Research on conceptual design of simplified nuclear safety instrument and control system

    International Nuclear Information System (INIS)

    Huang Jie

    2015-01-01

    The Nuclear safety instrument and control system is directly related to the safety of the reactor. So redundant and diversity design is used to ensure the system's security and reliability. This make the traditional safety system large, more cabinets and wiring complexity. To solve these problem, we can adopt new technology to make the design more simple. The simplify conceptual design can make the system less cabinets, less wiring, but high security, strong reliability. (author)

  13. Hardware resilience: a way to achieve reliability and safety in new nuclear reactors I and C systems

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de, E-mail: msantana@ien.gov.br, E-mail: paulov@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Divisão de Engenharia Nuclear. Serviço de Instrumentação; Nedjah, Nadia, E-mail: nadia@eng.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia de Sistemas e Telecomunicações

    2017-07-01

    The idea that systems have a property called ‘resilience’ has emerged in the last decade [1]. In this paper we intend to bring the idea of resilient systems for the hardware applied in safety-critical systems, such as the new nuclear reactor instrumentation and control (I and C) systems. The new systems (based in hardware description language (HDL) programmable devices) have been developed in response to the obsolescence of old analog technologies and current microprocessor-based digital technologies. Although HDL programmable devices have been widely used in various other industries for decades, they are still very new in nuclear reactors systems, which can be seen as a challenge and risk in the safety analyses and licensing efforts for utilities and designers. The goal of this work is to develop and test hardware architectures to tolerate the occurrence of faults, including multiple faults, minimizing the impact of the recovery process on system availability. Basic concepts of resilience in complex systems, as 'return to equilibrium', 'robustness' and 'extra adaptive capacity' were analyzed from the point of view of hardware architectures, leading to linkages between concepts and methods for resilience using an approach that increases reliability and simplifies the licensing process of systems based in HDL programmable devices in nuclear plants. (author)

  14. Hardware resilience: a way to achieve reliability and safety in new nuclear reactors I and C systems

    International Nuclear Information System (INIS)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Nedjah, Nadia

    2017-01-01

    The idea that systems have a property called ‘resilience’ has emerged in the last decade [1]. In this paper we intend to bring the idea of resilient systems for the hardware applied in safety-critical systems, such as the new nuclear reactor instrumentation and control (I and C) systems. The new systems (based in hardware description language (HDL) programmable devices) have been developed in response to the obsolescence of old analog technologies and current microprocessor-based digital technologies. Although HDL programmable devices have been widely used in various other industries for decades, they are still very new in nuclear reactors systems, which can be seen as a challenge and risk in the safety analyses and licensing efforts for utilities and designers. The goal of this work is to develop and test hardware architectures to tolerate the occurrence of faults, including multiple faults, minimizing the impact of the recovery process on system availability. Basic concepts of resilience in complex systems, as 'return to equilibrium', 'robustness' and 'extra adaptive capacity' were analyzed from the point of view of hardware architectures, leading to linkages between concepts and methods for resilience using an approach that increases reliability and simplifies the licensing process of systems based in HDL programmable devices in nuclear plants. (author)

  15. Reliability of Cyber Physical Systems with Focus on Building Management Systems

    DEFF Research Database (Denmark)

    Lazarova-Molnar, Sanja; Shaker, Hamid Reza; Mohamed, Nader

    2016-01-01

    with our focus CPS, i.e. building management systems (BMS), which are not always safety critical per se, but under special circumstances they can become such. This certainly depends on the purpose of the building. We can easily imagine BMS of hospital buildings as safety-critical, but also BMS of buildings......Cyber-physical systems are slowly emerging to dominate our world. Cyber-physical systems (CPS) are systems that tightly integrates users, devices and software. Whereas many of these systems are obviously safety-critical systems, some of them become so under special circumstances. This is the case...... that store sensitive materials and equipment that could be of biological nature or encompassing sensitive technology that would need special temperature, humidity and light settings. For this reason, in this paper we would like to emphasize on the importance of reliability of CPS in general, with a special...

  16. Systems reliability/structural reliability

    International Nuclear Information System (INIS)

    Green, A.E.

    1980-01-01

    The question of reliability technology using quantified techniques is considered for systems and structures. Systems reliability analysis has progressed to a viable and proven methodology whereas this has yet to be fully achieved for large scale structures. Structural loading variants over the half-time of the plant are considered to be more difficult to analyse than for systems, even though a relatively crude model may be a necessary starting point. Various reliability characteristics and environmental conditions are considered which enter this problem. The rare event situation is briefly mentioned together with aspects of proof testing and normal and upset loading conditions. (orig.)

  17. Operational safety system reliability. Progress report, November 15, 1975--May 14, 1976

    International Nuclear Information System (INIS)

    Hockenbury, R.W.; Yeater, M.L.

    1976-05-01

    The report describes the objectives and present status of a study concerning the operational reliability of nuclear power plants. The purpose of the study is to develop utilitarian models for use with the Liquid-Metal-Cooled Fast Breeder Reactor; initial testing of the formalism can be carried out with LWR operational data. Methods are being directed towards (1) day-to-day operation of the nuclear plant protection system and (2) to better understand the protection system sensor characteristics in order to anticipate off-normal conditions. The initial models now underway are based on moment-matching, confidence bounding, and convolution methods in the case of the protection system reliability, and for the sensor response function, a convolution of component reliability probability distributions and noise signatures

  18. 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

  19. 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.

  20. Reliability Analysis of the CERN Radiation Monitoring Electronic System CROME

    CERN Document Server

    AUTHOR|(CDS)2126870

    For the new in-house developed CERN Radiation Monitoring Electronic System (CROME) a reliability analysis is necessary to ensure compliance with the statu-tory requirements regarding the Safety Integrity Level. The required Safety Integrity Level by IEC 60532 standard is SIL 2 (for the Safety Integrated Functions Measurement, Alarm Triggering and Interlock Triggering). The first step of the reliability analysis was a system and functional analysis which served as basis for the implementation of the CROME system in the software “Iso-graph”. In the “Prediction” module of Isograph the failure rates of all components were calculated. Failure rates for passive components were calculated by the Military Standard 217 and failure rates for active components were obtained from lifetime tests by the manufacturers. The FMEA was carried out together with the board designers and implemented in the “FMECA” module of Isograph. The FMEA served as basis for the Fault Tree Analysis and the detection of weak points...

  1. 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)

  2. 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)

  3. 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

  4. 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.

  5. Considerations on nuclear reactor passive safety systems

    International Nuclear Information System (INIS)

    2016-01-01

    After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

  6. Work Practice, Safety and Heedfulness. Studies of Organizational Reliability in Hospitals and Nuclear Power Plants

    International Nuclear Information System (INIS)

    Gauthereau, Vincent

    2003-01-01

    The study of safety in complex systems has focused on different issues over the past decades. This focus was often linked to the conclusions of previous accidents'/incidents' analyses. When accidents were attributed to technical causes, safety research focused on technical developments. When they were later attributed to 'human errors', safety research focused on this 'component'. And when, since the mid-eighties accidents have been attributed to 'organizational factors', safety research has focused on these very same 'organizational factors'. The present thesis argues for a 'practice view' over safety to be taken. This view is mainly drawn from the field of research on High Reliability Organizations (HRO). HRO theorists' point of view on safety is that we can operate complex systems safely despite the fact that we have made them so complex that they are prone to 'normal accidents'. Humans involved in the operation of our systems actually create safety. Safety is formed through the adaptation of work practice to local conditions, and this adaptation is part of safe operation. Safety is not only a substantial quality of our socio-technical systems: the discursive dimension of safety actually seems to be a central component of safety creation. However, the adaptive ability of HRO can sometimes become their downfall. Adaptation, which is the backbone of safety, can sometimes be a drawback as well. Consequently, the practice view of safety, proposed in the present work, argues that we need to further comprehend how work practice evolves over time, and more specifically what are the inherent characteristics of work practice that create this evolution. Empirical studies from health-care and nuclear power generation highlight different details about organizational reliability. For instance, one study of planning at a nuclear power plant draws our attention to the different roles of planning in the organization. Another study, within heath-care, underlines the evolution of

  7. Reliability analysis of digital I and C systems at KAERI

    International Nuclear Information System (INIS)

    Kim, Man Cheol

    2013-01-01

    This paper provides an overview of the ongoing research activities on a reliability analysis of digital instrumentation and control (I and C) systems of nuclear power plants (NPPs) performed by the Korea Atomic Energy Research Institute (KAERI). The research activities include the development of a new safety-critical software reliability analysis method by integrating the advantages of existing software reliability analysis methods, a fault coverage estimation method based on fault injection experiments, and a new human reliability analysis method for computer-based main control rooms (MCRs) based on human performance data from the APR-1400 full-scope simulator. The research results are expected to be used to address various issues such as the licensing issues related to digital I and C probabilistic safety assessment (PSA) for advanced digital-based NPPs. (author)

  8. 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

  9. Reliability model for helicopter main gearbox lubrication system using influence diagrams

    International Nuclear Information System (INIS)

    Rashid, H.S.J.; Place, C.S.; Mba, D.; Keong, R.L.C.; Healey, A.; Kleine-Beek, W.; Romano, M.

    2015-01-01

    The loss of oil from a helicopter main gearbox (MGB) leads to increased friction between components, a rise in component surface temperatures, and subsequent mechanical failure of gearbox components. A number of significant helicopter accidents have been caused due to such loss of lubrication. This paper presents a model to assess the reliability of helicopter MGB lubricating systems. Safety risk modeling was conducted for MGB oil system related accidents in order to analyse key failure mechanisms and the contributory factors. Thus, the dominant failure modes for lubrication systems and key contributing components were identified. The Influence Diagram (ID) approach was then employed to investigate reliability issues of the MGB lubrication systems at the level of primary causal factors, thus systematically investigating a complex context of events, conditions, and influences that are direct triggers of the helicopter MGB lubrication system failures. The interrelationships between MGB lubrication system failure types were thus identified, and the influence of each of these factors on the overall MGB lubrication system reliability was assessed. This paper highlights parts of the HELMGOP project, sponsored by the European Aviation Safety Agency to improve helicopter main gearbox reliability. - Highlights: • We investigated methods to optimize helicopter MGB oil system run-dry capability. • Used Influence Diagram to assess design and maintenance factors of MGB oil system. • Factors influencing overall MGB lubrication system reliability were identified. • This globally influences current and future helicopter MGB designs

  10. Reliability analysis of digital safety systems at nuclear power plants

    International Nuclear Information System (INIS)

    Sopira Vladimir; Kovacs, Zoltan

    2015-01-01

    Reliability analysis of digital reactor protection systems built on the basis of TELEPERM XS is described, and experience gained by the Slovak RELKO company during the past 20 years in this domain is highlighted. (orig.)

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

    International Nuclear Information System (INIS)

    Berning, A.; Spindler, H.

    1983-01-01

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

  12. Technical features of ABWR safety systems

    International Nuclear Information System (INIS)

    Sugisaki, Toshihiko; Tominaga, Kenji; Horiuchi, Tetsuo

    1986-01-01

    The engineering safety facilities of ABWRs have been disigned so as to have many excellent characteristics such as safety, reliability and economy, reflecting the merit of adopting new technology such as internal pumps and new control rod driving mechanism, and coupled with the safety peculiar to BWRs. In this paper, about ECCS, containment vessels and others which compose the engineering safety facilities of ABWRs, the characteristics related to the safety owing to the adoption of internal pumps and others, and the evaluation of the performance at the time of various accidents are discussed. As the results of safety evaluation, it was clarified that due to the safety peculiar to ABWRs and the characteristics of the safety facilities, the large increases of safety, reliability and economy have been planned in the ABWRs, and for example, core flooding can be maintained even at the time of a hypothetical loss of coolant accident. BWRs have the simple system constitution, good self controllability, large natural circulation ability, simple operation control method and excellent ability of confining heat and radioactivity. BWRs have three safety functions to stop reactors, to remove heat from reactors, and to confine radioactive substances. These functions of ABWRs were evaluated, and very high safety was confirmed. (Kako, I.)

  13. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

  14. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    International Nuclear Information System (INIS)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

  15. Research on Connection and Function Reliability of the Oil&Gas Pipeline System

    Directory of Open Access Journals (Sweden)

    Xu Bo

    2017-01-01

    Full Text Available Pipeline transportation is the optimal way for energy delivery in terms of safety, efficiency and environmental protection. Because of the complexity of pipeline external system including geological hazards, social and cultural influence, it is a great challenge to operate the pipeline safely and reliable. Therefore, the pipeline reliability becomes an important issue. Based on the classical reliability theory, the analysis of pipeline system is carried out, then the reliability model of the pipeline system is built, and the calculation is addressed thereafter. Further the connection and function reliability model is applied to a practical active pipeline system, with the use of the proposed methodology of the pipeline system; the connection reliability and function reliability are obtained. This paper firstly presented to considerate the connection and function reliability separately and obtain significant contribution to establish the mathematical reliability model of pipeline system, hence provide fundamental groundwork for the pipeline reliability research in the future.

  16. Experiences from maintaining the reliability of a nuclear standby diesel generator system

    International Nuclear Information System (INIS)

    Tammi, P.

    1982-01-01

    The nuclear standby diesel generator system is quite complicated comprising several mechanical and electrotechnical components, on which the reliability of the system is depending. It is an important support system of the plant safety system, and like the safety system it is composed of separate redundant units. The Loviisa nuclear power station has eight diesel generators. The first four of them were taken into operation in 1976. When the frequency of some mechanical failures showed increase, a project was started at the end of 1980 with the intention to find out potential failure possibilities and means for prevention of failures. The work has been mainly concentrated on improving the reliability of the diesel engines. (Auth.)

  17. Operation safety of complex industrial systems. Main concepts

    International Nuclear Information System (INIS)

    Zwingelstein, G.

    2009-01-01

    Operation safety consists in knowing, evaluating, foreseeing, measuring and mastering the technological system and human failures in order to avoid their impacts on health and people's safety, on productivity, and on the environment, and to preserve the Earth's resources. This article recalls the main concepts of operation safety: 1 - evolutions in the domain; 2 - failures, missions and functions of a system and of its components: functional failure, missions and functions, industrial processes, notions of probability; 3 - basic concepts and operation safety: reliability, unreliability, failure density, failure rate, relations between them, availability, maintainability, safety. (J.S.)

  18. Aspects of safety and reliability for fusion magnet systems first annual report

    International Nuclear Information System (INIS)

    Powell, J.

    1976-01-01

    General systems aspects of fusion magnet safety are examined first, followed by specific detailed analyses covering structural, thermal, electrical, and other aspects of fusion magnet safety. The design examples chosen for analysis are illustrative and are not intended to be definitive, since fusion magnet designs are rapidly evolving. Included is a comprehensive collection of design and operating data relating to the safety of existing superconducting magnet systems. The remainder of the overview lists the main conclusions developed from the work to date. These should be regarded as initial steps. Since this study has concentrated on examining potential safety concerns, it may tend to overemphasize the problems of fusion magnets. In fact, many aspects of fusion magnets are well developed and are consistent with good safety practice. A short summary of the findings of this study is given

  19. 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)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. An Introduction To Reliability

    International Nuclear Information System (INIS)

    Park, Kyoung Su

    1993-08-01

    This book introduces reliability with definition of reliability, requirement of reliability, system of life cycle and reliability, reliability and failure rate such as summary, reliability characteristic, chance failure, failure rate which changes over time, failure mode, replacement, reliability in engineering design, reliability test over assumption of failure rate, and drawing of reliability data, prediction of system reliability, conservation of system, failure such as summary and failure relay and analysis of system safety.

  2. ABWR (K-6/7) construction experience (computer-based safety system)

    International Nuclear Information System (INIS)

    Yokomura, T.

    1998-01-01

    TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)

  3. 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

  4. 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.

  5. Architecture Level Safety Analyses for Safety-Critical Systems

    Directory of Open Access Journals (Sweden)

    K. S. Kushal

    2017-01-01

    Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

  6. Feasibility study for the European Reliability Data System (ERDS)

    International Nuclear Information System (INIS)

    Mancini, G.

    1980-01-01

    In the framework of the Reactor Safety Programme of the Commission of the European Communities, the JRC - Ispra Establishment has performed a feasibility study for an integrated European Reliability Data System, the aim of which is the collection and organization of information related to the operation of LWRs with regard to component and systems behaviour, abnormal occurrences, outages, etc. Component Event Data Bank (CEGB), Abnormal Occurrences Reporting System, Generic Reliability Parameter Data Bank, Operating Unit Status Reports and the main activities carried out during the last two years are described. The most important achievements are briefly reported, such as: Reference Classification for Systems, Components and Failure Events, Informatic Structure of the Pilot Experiment of the CEDB, Information Retrieval System for Abnormal Occurrences Reports, Data Bank on Component Reliability Parameters, System on the Exchange of Operation Experience of LWRs, Statistical Data Treatment. Finally, the general conclusions of the feasibility study are summarized: the possibility and the usefulness for the creation of an integrated European Reliability Data System are outlined. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  8. Reliability And Maintenance Analysis Of CCTV Systems Used In Rail Transport

    Directory of Open Access Journals (Sweden)

    Siergiejczyk Mirosław

    2015-11-01

    Full Text Available CCTV systems are widely used across plethora of industrial areas including transport, where their function is to support transport telematics systems. Among others, they are used to ensure travel safety. This paper presented a reliability and maintenance analysis of CCTV. It led to building a relationships graph and then Chapman–Kolmogorov system of equations was derived to describe it. Drawing on those equations, relationships for calculating probability of system staying in state of full ability SPZ, state of the impendency over safety SZB1 as well as state of unreliability of safety SB were derived.

  9. 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

  10. Optimised and balanced structural and system reliability of offshore wind turbines. An account

    Energy Technology Data Exchange (ETDEWEB)

    Tarp-Johansen, N.J.; Kozine, I. (Risoe National Lab., DTU, Roskilde, (DK)); Rademarkers, L. (Netherlands Energy Research Foundation (NL)); Dalsgaard Soerensen, J. (Aalborg Univ. (DK)) Ronold, K. (Det Norske Veritas (DK))

    2005-04-15

    This report gives the results of the research project 'Optimised and Uniform Safety and Reliability of Offshore Wind Turbines (an account)'. The main subject of the project has been the account of the state-of-the art of knowledge about, and/or attempts to, harmonisation of the structural reliability of wind turbines, on the one hand, and the reliability of the wind turbine's control/safety system, on the other hand. Within the project some research pointing ahead has also been conducted. (au)

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

    International Nuclear Information System (INIS)

    2000-01-01

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

  12. Reliable computer systems.

    Science.gov (United States)

    Wear, L L; Pinkert, J R

    1993-11-01

    In this article, we looked at some decisions that apply to the design of reliable computer systems. We began with a discussion of several terms such as testability, then described some systems that call for highly reliable hardware and software. The article concluded with a discussion of methods that can be used to achieve higher reliability in computer systems. Reliability and fault tolerance in computers probably will continue to grow in importance. As more and more systems are computerized, people will want assurances about the reliability of these systems, and their ability to work properly even when sub-systems fail.

  13. 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...

  14. Systems reliability Benchmark exercise part 1-Description and results

    International Nuclear Information System (INIS)

    Amendola, A.

    1986-01-01

    The report describes aims, rules and results of the Systems Reliability Benchmark Exercise, which has been performed in order to assess methods and procedures for reliability analysis of complex systems and involved a large number of European organizations active in NPP safety evaluation. The exercise included both qualitative and quantitative methods and was structured in such a way that separation of the effects of uncertainties in modelling and in data on the overall spread was made possible. Part I describes the way in which RBE has been performed, its main results and conclusions

  15. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

    A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-26

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

  18. 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.

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

    International Nuclear Information System (INIS)

    Sitorus Pane, Jupiter

    2001-01-01

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

  20. 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.

  1. Assessment of Safety Standards for Automotive Electronic Control Systems

    Science.gov (United States)

    2016-06-01

    This report summarizes the results of a study that assessed and compared six industry and government safety standards relevant to the safety and reliability of automotive electronic control systems. These standards include ISO 26262 (Road Vehicles - ...

  2. 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

  3. Programmable Electronic Safety Systems

    International Nuclear Information System (INIS)

    Parry, R.

    1993-05-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement failsafe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL

  4. A critical evaluation of deterministic methods in size optimisation of reliable and cost effective standalone hybrid renewable energy systems

    International Nuclear Information System (INIS)

    Maheri, Alireza

    2014-01-01

    Reliability of a hybrid renewable energy system (HRES) strongly depends on various uncertainties affecting the amount of power produced by the system. In the design of systems subject to uncertainties, both deterministic and nondeterministic design approaches can be adopted. In a deterministic design approach, the designer considers the presence of uncertainties and incorporates them indirectly into the design by applying safety factors. It is assumed that, by employing suitable safety factors and considering worst-case-scenarios, reliable systems can be designed. In fact, the multi-objective optimisation problem with two objectives of reliability and cost is reduced to a single-objective optimisation problem with the objective of cost only. In this paper the competence of deterministic design methods in size optimisation of reliable standalone wind–PV–battery, wind–PV–diesel and wind–PV–battery–diesel configurations is examined. For each configuration, first, using different values of safety factors, the optimal size of the system components which minimises the system cost is found deterministically. Then, for each case, using a Monte Carlo simulation, the effect of safety factors on the reliability and the cost are investigated. In performing reliability analysis, several reliability measures, namely, unmet load, blackout durations (total, maximum and average) and mean time between failures are considered. It is shown that the traditional methods of considering the effect of uncertainties in deterministic designs such as design for an autonomy period and employing safety factors have either little or unpredictable impact on the actual reliability of the designed wind–PV–battery configuration. In the case of wind–PV–diesel and wind–PV–battery–diesel configurations it is shown that, while using a high-enough margin of safety in sizing diesel generator leads to reliable systems, the optimum value for this margin of safety leading to a

  5. 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)

  6. 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

  7. 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

  8. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

    Yu, J.; Kim, J. Y. [Chungnam Nat. Univ., Daejeon (Korea, Republic of)

    2006-06-15

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system.

  9. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

    Yu, J.; Kim, J. Y.

    2006-06-01

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system

  10. The selection of field component reliability data for use in nuclear safety studies

    International Nuclear Information System (INIS)

    Coxson, B.A.; Tabaie, Mansour

    1990-01-01

    The paper reviews the user requirements for field component failure data in nuclear safety studies, and the capability of various data sources to satisfy these requirements. Aspects such as estimating the population of items exposed to failure, incompleteness, and under-reporting problems are discussed. The paper takes as an example the selection of component reliability data for use in the Pre-Operational Safety Report (POSR) for Sizewell 'B' Power Station, where field data has in many cases been derived from equipment other than that to be procured and operated on site. The paper concludes that the main quality sought in the available data sources for such studies is the ability to examine failure narratives in component reliability data systems for equipment performing comparable duties to the intended plant application. The main benefit brought about in the last decade is the interactive access to data systems which are adequately structured with regard to the equipment covered, and also provide a text-searching capability of quality-controlled event narratives. (author)

  11. 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

  12. The micro-processor controlled process radiation monitoring system for reactor safety systems

    International Nuclear Information System (INIS)

    Mizuno, K.; Noguchi, A.; Kumagami, S.; Gotoh, Y.; Kumahara, T.; Arita, S.

    1986-01-01

    Digital computers are soon expected to be applied to various real-time safety and safety-related systems in nuclear power plants. Hitachi is now engaged in the development of a micro-processor controlled process radiation monitoring system, which operates on digital processing methods employed with a log ratemeter. A newly defined methodology of design and test procedures is being applied as a means of software program verification for these safety systems. Recently implemented micro-processor technology will help to achieve an advanced man-machine interface and highly reliable performance. (author)

  13. On the safeness of examinees and the reliability of system

    International Nuclear Information System (INIS)

    Kudo, Kazumi; Kanda, Kosuke; Saito, Kazuhiko; Maesawa, Tsuneharu; Idekami, Tomio

    1979-01-01

    The control technique of the reliability of examination system was investigated from the viewpoint of patient safety and image information, based on the prevention of microshock owing to circulatory organ checking system. As for the equipments in hospitals, the size of rooms, air conditioning system, power source installation, earth and piping arrangements should be fully discussed at the planning stage. EPR system must be introduced for the prevention for microshock. Intensive education and training are required for operators to secure safeness in operation. Thorough care should be taken to prevent bacilli infection. Further examinations were made on the control technique of the reliability of photographing system from viewpoint of image information, and it is necessary to study the factors for obtaining the reliability of compound machinery components and the devices of generating radiation. (Kobatake, H.)

  14. Operation safety of complex industrial systems

    International Nuclear Information System (INIS)

    Zwingelstein, G.

    1999-01-01

    Zero fault or zero risk is an unreachable goal in industrial activities like nuclear activities. However, methods and techniques exist to reduce the risks to the lowest possible and acceptable level. The operation safety consists in the recognition, evaluation, prediction, measurement and mastery of technological and human faults. This paper analyses each of these points successively: 1 - evolution of operation safety; 2 - definitions and basic concepts: failure, missions and functions of a system and of its components, basic concepts and operation safety; 3 - forecasting analysis of operation safety: reliability data, data-banks, precautions for the use of experience feedback data; realization of an operation safety study: management of operation safety, quality assurance, critical review and audit of operation safety studies; 6 - conclusions. (J.S.)

  15. 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

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

    International Nuclear Information System (INIS)

    Melis, M.; Mancini, G.

    1982-01-01

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

  17. CERN safety system monitoring - SSM

    International Nuclear Information System (INIS)

    Hakulinen, T.; Ninin, P.; Valentini, F.; Gonzalez, J.; Salatko-Petryszcze, C.

    2012-01-01

    CERN SSM (Safety System Monitoring) is a system for monitoring state-of-health of the various access and safety systems of the CERN site and accelerator infrastructure. The emphasis of SSM is on the needs of maintenance and system operation with the aim of providing an independent and reliable verification path of the basic operational parameters of each system. Included are all network-connected devices, such as PLCs (local purpose control unit), servers, panel displays, operator posts, etc. The basic monitoring engine of SSM is a freely available system-monitoring framework Zabbix, on top of which a simplified traffic-light-type web-interface has been built. The web-interface of SSM is designed to be ultra-light to facilitate access from hand-held devices over slow connections. The underlying Zabbix system offers history and notification mechanisms typical of advanced monitoring systems. (authors)

  18. Reliability of redundant structures of nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Vojnovic, B.

    1983-01-01

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

  19. Some aspects of the interaction between systems- and structural reliability

    International Nuclear Information System (INIS)

    Schueller, G.K.; Schmitt, W.

    1979-01-01

    The purpose of this paper is to study the interaction between systems- and structural reliability analysis with reference to the design of structural components of LWR. Presently the evaluation of systems reliability is carried out apart from structural reliability analysis. Moreover, two basically different methodologies are used for analysis. While in systems analysis the simplified binary approach is still generally accepted, in structural reliability one has to resort to more sophisticated procedures to obtain realistic results. The interactive effect may be illustrated as follows: For example, the integrity of the primary circuit interacts with the integrity of the containment structure. This means that the probability of occurrence of the pipe rupture which may cause a LOCA and consequently leads to a build-up of temperature and pressure within the containment affects directly its structural reliability. The piping system, particularly the primary piping, in turn interacts with the protective system, which is part of the safety system. This piping structure is also subjected to various operational loading conditions. In a numerical example dealing with leakage probabilities of pipes it is shown how methods of structural reliability may be used to gain more insight in the estimation of failure rates of system components. (orig.)

  20. Development of Network Protocol for the Integrated Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. W.; Baek, J. I.; Lee, S. H.; Park, C. S.; Park, K. H.; Shin, J. M. [Hannam Univ., Daejeon (Korea, Republic of)

    2007-06-15

    Communication devices in the safety system of nuclear power plants are distinguished from those developed for commercial purposes in terms of a strict requirement of safety. The concept of safety covers the determinability, the reliability, and the separation/isolation to prevent the undesirable interactions among devices. The safety also requires that these properties be never proof less. Most of the current commercialized communication products rarely have the safety properties. Moreover, they can be neither verified nor validated to satisfy the safety property of implementation process. This research proposes the novel architecture and protocol of a data communication network for the safety system in nuclear power plants.

  1. Development of Network Protocol for the Integrated Safety System

    International Nuclear Information System (INIS)

    Park, S. W.; Baek, J. I.; Lee, S. H.; Park, C. S.; Park, K. H.; Shin, J. M.

    2007-06-01

    Communication devices in the safety system of nuclear power plants are distinguished from those developed for commercial purposes in terms of a strict requirement of safety. The concept of safety covers the determinability, the reliability, and the separation/isolation to prevent the undesirable interactions among devices. The safety also requires that these properties be never proof less. Most of the current commercialized communication products rarely have the safety properties. Moreover, they can be neither verified nor validated to satisfy the safety property of implementation process. This research proposes the novel architecture and protocol of a data communication network for the safety system in nuclear power plants

  2. 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.

  3. 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

  4. Programmable electronic safety systems

    International Nuclear Information System (INIS)

    Parry, R.R.

    1993-01-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement fail-safe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL

  5. Autonomous system for launch vehicle range safety

    Science.gov (United States)

    Ferrell, Bob; Haley, Sam

    2001-02-01

    The Autonomous Flight Safety System (AFSS) is a launch vehicle subsystem whose ultimate goal is an autonomous capability to assure range safety (people and valuable resources), flight personnel safety, flight assets safety (recovery of valuable vehicles and cargo), and global coverage with a dramatic simplification of range infrastructure. The AFSS is capable of determining current vehicle position and predicting the impact point with respect to flight restriction zones. Additionally, it is able to discern whether or not the launch vehicle is an immediate threat to public safety, and initiate the appropriate range safety response. These features provide for a dramatic cost reduction in range operations and improved reliability of mission success. .

  6. European Workshop Industrical Computer Science Systems approach to design for safety

    Science.gov (United States)

    Zalewski, Janusz

    1992-01-01

    This paper presents guidelines on designing systems for safety, developed by the Technical Committee 7 on Reliability and Safety of the European Workshop on Industrial Computer Systems. The focus is on complementing the traditional development process by adding the following four steps: (1) overall safety analysis; (2) analysis of the functional specifications; (3) designing for safety; (4) validation of design. Quantitative assessment of safety is possible by means of a modular questionnaire covering various aspects of the major stages of system development.

  7. 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

  8. Manufacture of Platform Prototype for Digital Safety System

    International Nuclear Information System (INIS)

    Lee, S. Y.; Kim, J. S.; Kim, J. M.

    2010-01-01

    Unit controller is a basic unit of digital safety system platform prototype. The typical unit controller is comprised of CPB(CPU board), CMB(communication board), AIB(Analog input board), AOB(Analog output board), CIB(contact input board), COB(contact output board), and a subrack. It is developed according to H/W development procedure and S/W development life cycle. A digital safety system(for example, plant protection system) is the assemblies of unit controllers. CPB performs the function of each system. DSP(digital signal processor) is built in CPB. CMB is responsible for communication between unit controllers. NSD(Network Switching Device) exchanges data between the unit controllers. Each unit controller of the platform are connected to NSD through CMB. Reliability analyses on unit controller and NSD are performed. These reliability data are used as input of technical validation

  9. 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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  11. Design Information from the PSA for Digital Safety-Critical Systems

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung Cheol

    2005-01-01

    Many safety-critical applications such as nuclear field application usually adopt a similar design strategy for digital safety-critical systems. Their differences from the normal design for the non-safety-critical applications could be summarized as: multiple-redundancy, highly reliable components, strengthened monitoring mechanism, verified software, and automated test procedure. These items are focusing on maintaining the capability to perform the given safety function when it is requested. For the past several decades, probabilistic safety assessment (PSA) techniques are used in the nuclear industry to assess the relative effects of contributing events on plant risk and system reliability. They provide a unifying means of assessing physical faults, recovery processes, contributing effects, human actions, and other events that have a high degree of uncertainty. The applications of PSA provide not only the analysis results of already installed system but also the useful information for the system under design. The information could be derived from the PSA experience of the various safety-critical systems. Thanks to the design flexibility, the digital system is one of the most suitable candidates for risk-informed design (RID). In this article, we will describe the feedbacks for system design and try to develop a procedure for RID. Even though the procedure is not sophisticated enough now, it could be the start point of the further investigation for developing more complete and practical methodology

  12. Preliminary safety evaluation for CSR1000 with passive safety system

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

    2014-01-01

    Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

  13. 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

  14. 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

  15. 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

  16. Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

    International Nuclear Information System (INIS)

    Wetzel, N.; Scharfe, A.

    1998-01-01

    The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-28

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

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

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

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

  19. Operational safety system performance alternative to the WANO's indicator

    International Nuclear Information System (INIS)

    Lyra, Moacir

    2002-01-01

    One of the operational safety performance indicators recommended by the World Association of Nuclear Operators (WANO) and adopted by Electronuclear is the reliability of the safety systems. The parameter selected to represent this indicator is the average unavailability of the trains of the concerned system. This parameter would be universally representative of the reliability for comparison purpose only if all nuclear power plants were designed within the same redundancy criteria. Considering the diversity of design criteria of the power plants in operation and based on a probabilistic approach, this paper proposes new performance indicators which are comparable regardless the redundancy criteria of the system. A case example applied to a system of the Angra 2 nuclear power plant shows that, even though with the plant in the infancy phase, the performance of the system in the period is very good. (author)

  20. Improvement of the reliability graph with general gates to analyze the reliability of dynamic systems that have various operation modes

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Seung Ki [Div. of Research Reactor System Design, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); No, Young Gyu; Seong, Poong Hyun [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2016-04-15

    The safety of nuclear power plants is analyzed by a probabilistic risk assessment, and the fault tree analysis is the most widely used method for a risk assessment with the event tree analysis. One of the well-known disadvantages of the fault tree is that drawing a fault tree for a complex system is a very cumbersome task. Thus, several graphical modeling methods have been proposed for the convenient and intuitive modeling of complex systems. In this paper, the reliability graph with general gates (RGGG) method, one of the intuitive graphical modeling methods based on Bayesian networks, is improved for the reliability analyses of dynamic systems that have various operation modes with time. A reliability matrix is proposed and it is explained how to utilize the reliability matrix in the RGGG for various cases of operation mode changes. The proposed RGGG with a reliability matrix provides a convenient and intuitive modeling of various operation modes of complex systems, and can also be utilized with dynamic nodes that analyze the failure sequences of subcomponents. The combinatorial use of a reliability matrix with dynamic nodes is illustrated through an application to a shutdown cooling system in a nuclear power plant.

  1. Improvement of the reliability graph with general gates to analyze the reliability of dynamic systems that have various operation modes

    International Nuclear Information System (INIS)

    Shin, Seung Ki; No, Young Gyu; Seong, Poong Hyun

    2016-01-01

    The safety of nuclear power plants is analyzed by a probabilistic risk assessment, and the fault tree analysis is the most widely used method for a risk assessment with the event tree analysis. One of the well-known disadvantages of the fault tree is that drawing a fault tree for a complex system is a very cumbersome task. Thus, several graphical modeling methods have been proposed for the convenient and intuitive modeling of complex systems. In this paper, the reliability graph with general gates (RGGG) method, one of the intuitive graphical modeling methods based on Bayesian networks, is improved for the reliability analyses of dynamic systems that have various operation modes with time. A reliability matrix is proposed and it is explained how to utilize the reliability matrix in the RGGG for various cases of operation mode changes. The proposed RGGG with a reliability matrix provides a convenient and intuitive modeling of various operation modes of complex systems, and can also be utilized with dynamic nodes that analyze the failure sequences of subcomponents. The combinatorial use of a reliability matrix with dynamic nodes is illustrated through an application to a shutdown cooling system in a nuclear power plant

  2. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  3. Reliability analysis and initial requirements for FC systems and stacks

    Science.gov (United States)

    Åström, K.; Fontell, E.; Virtanen, S.

    In the year 2000 Wärtsilä Corporation started an R&D program to develop SOFC systems for CHP applications. The program aims to bring to the market highly efficient, clean and cost competitive fuel cell systems with rated power output in the range of 50-250 kW for distributed generation and marine applications. In the program Wärtsilä focuses on system integration and development. System reliability and availability are key issues determining the competitiveness of the SOFC technology. In Wärtsilä, methods have been implemented for analysing the system in respect to reliability and safety as well as for defining reliability requirements for system components. A fault tree representation is used as the basis for reliability prediction analysis. A dynamic simulation technique has been developed to allow for non-static properties in the fault tree logic modelling. Special emphasis has been placed on reliability analysis of the fuel cell stacks in the system. A method for assessing reliability and critical failure predictability requirements for fuel cell stacks in a system consisting of several stacks has been developed. The method is based on a qualitative model of the stack configuration where each stack can be in a functional, partially failed or critically failed state, each of the states having different failure rates and effects on the system behaviour. The main purpose of the method is to understand the effect of stack reliability, critical failure predictability and operating strategy on the system reliability and availability. An example configuration, consisting of 5 × 5 stacks (series of 5 sets of 5 parallel stacks) is analysed in respect to stack reliability requirements as a function of predictability of critical failures and Weibull shape factor of failure rate distributions.

  4. Development and applications of a safety assessment system for promoting safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Takano, Ken-ichi; Hasegawa, Naoko; Hirose, Ayako; Hayase, Ken-ichi

    2004-01-01

    For past five years, CRIEPI has been continuing efforts to develop and make applications of a 'safety assessment system' which enable to measure the safety level of organization. This report describe about frame of the system, assessment results and its reliability, and relation between labor accident rate in the site and total safety index (TSI), which can be obtained by the principal factors analysis. The safety assessment in this report is based on questionnaire survey of employee. The format and concrete questionnaires were developed using existing literatures including organizational assessment tools. The tailored questionnaire format involved 124 questionnaire items. The assessment results could be considered as a well indicator of the safety level of organization, safety management, and safety awareness of employee. (author)

  5. Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

    International Nuclear Information System (INIS)

    Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

    1996-01-01

    Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits

  6. 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

  7. 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

  8. 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

  9. Collection of methods for reliability and safety engineering

    International Nuclear Information System (INIS)

    Fussell, J.B.; Rasmuson, D.M.; Wilson, J.R.; Burdick, G.R.; Zipperer, J.C.

    1976-04-01

    The document presented contains five reports each describing a method of reliability and safety engineering. Report I provides a conceptual framework for the study of component malfunctions during system evaluations. Report II provides methods for locating groups of critical component failures such that all the component failures in a given group can be caused to occur by the occurrence of a single separate event. These groups of component failures are called common cause candidates. Report III provides a method for acquiring and storing system-independent component failure logic information. The information stored is influenced by the concepts presented in Report I and also includes information useful in locating common cause candidates. Report IV puts forth methods for analyzing situations that involve systems which change character in a predetermined time sequence. These phased missions techniques are applicable to the hypothetical ''accident chains'' frequently analyzed for nuclear power plants. Report V presents a unified approach to cause-consequence analysis, a method of analysis useful during risk assessments. This approach, as developed by the Danish Atomic Energy Commission, is modified to reflect the format and symbology conventionally used for other types of analysis of nuclear reactor systems

  10. Systems Reliability Framework for Surface Water Sustainability and Risk Management

    Science.gov (United States)

    Myers, J. R.; Yeghiazarian, L.

    2016-12-01

    With microbial contamination posing a serious threat to the availability of clean water across the world, it is necessary to develop a framework that evaluates the safety and sustainability of water systems in respect to non-point source fecal microbial contamination. The concept of water safety is closely related to the concept of failure in reliability theory. In water quality problems, the event of failure can be defined as the concentration of microbial contamination exceeding a certain standard for usability of water. It is pertinent in watershed management to know the likelihood of such an event of failure occurring at a particular point in space and time. Microbial fate and transport are driven by environmental processes taking place in complex, multi-component, interdependent environmental systems that are dynamic and spatially heterogeneous, which means these processes and therefore their influences upon microbial transport must be considered stochastic and variable through space and time. A physics-based stochastic model of microbial dynamics is presented that propagates uncertainty using a unique sampling method based on artificial neural networks to produce a correlation between watershed characteristics and spatial-temporal probabilistic patterns of microbial contamination. These results are used to address the question of water safety through several sustainability metrics: reliability, vulnerability, resilience and a composite sustainability index. System reliability is described uniquely though the temporal evolution of risk along watershed points or pathways. Probabilistic resilience describes how long the system is above a certain probability of failure, and the vulnerability metric describes how the temporal evolution of risk changes throughout a hierarchy of failure levels. Additionally our approach allows for the identification of contributions in microbial contamination and uncertainty from specific pathways and sources. We expect that this

  11. An application of the fault tree analysis for the power system reliability estimation

    International Nuclear Information System (INIS)

    Volkanovski, A.; Cepin, M.; Mavko, B.

    2007-01-01

    The power system is a complex system with its main function to produce, transfer and provide consumers with electrical energy. Combinations of failures of components in the system can result in a failure of power delivery to certain load points and in some cases in a full blackout of power system. The power system reliability directly affects safe and reliable operation of nuclear power plants because the loss of offsite power is a significant contributor to the core damage frequency in probabilistic safety assessments of nuclear power plants. The method, which is based on the integration of the fault tree analysis with the analysis of the power flows in the power system, was developed and implemented for power system reliability assessment. The main contributors to the power system reliability are identified, both quantitatively and qualitatively. (author)

  12. Validation study on reliability analysis of main safety system in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Cho, Chang Keun; Kim, Yong Hui; Kim, Tae Hyeong; Hong, Seo Kee; Park, Keon Woo; Park, Chang Jea [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Cheong, Woo Sik [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Moon Kyu [KEPRI, Taejon (Korea, Republic of)

    1993-12-15

    The scope and contents of this validation study are to review the design changes of the four main safety systems in Wolsong 2/3/4 Nuclear Power Plants, to review the consideration of the above design changes in the AECL reports, the structure of fault trees, and the data base used in the quantification of the fault trees, to quantify the unavailabilities of main safety systems and check them if they meet the requirements, and to recommend desirable design changes in the emergency core cooling system to reduce the unavailability.

  13. System Reliability Engineering

    International Nuclear Information System (INIS)

    Lim, Tae Jin

    2005-02-01

    This book tells of reliability engineering, which includes quality and reliability, reliability data, importance of reliability engineering, reliability and measure, the poisson process like goodness of fit test and the poisson arrival model, reliability estimation like exponential distribution, reliability of systems, availability, preventive maintenance such as replacement policies, minimal repair policy, shock models, spares, group maintenance and periodic inspection, analysis of common cause failure, and analysis model of repair effect.

  14. 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

  15. Reliability analysis of neutron flux monitoring system for PFBR

    International Nuclear Information System (INIS)

    Rajesh, M.G.; Bhatnagar, P.V.; Das, D.; Pithawa, C.K.; Vinod, Gopika; Rao, V.V.S.S.

    2010-01-01

    The Neutron Flux Monitoring System (NFMS) measures reactor power, rate of change of power and reactivity changes in the core in all states of operation and shutdown. The system consists of instrument channels that are designed and built to have high reliability. All channels are required to have a Mean Time Between Failures (MTBF) of 150000 hours minimum. Failure Mode and Effects Analysis (FMEA) and failure rate estimation of NFMS channels has been carried out. FMEA is carried out in compliance with MIL-STD-338B. Reliability estimation of the channels is done according to MIL-HDBK-217FN2. Paper discusses the methodology followed for FMEA and failure rate estimation of two safety channels and results. (author)

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

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi; Kobayashi, Michiyuki

    1994-01-01

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

  17. Optimization of redundancy by using genetic algorithm for reliability of plant protection system

    International Nuclear Information System (INIS)

    Yoo, D. W.; Seong, S. H.; Kim, D. H.; Park, H. Y.; Gu, I. S.

    2000-01-01

    The design and development of a reliable protection system has been becoming a key issue in industry field because the reliability of system is considered as an important factor to perform the system's function successfully. Plant Protection System(PPS) guarantees the safety of plant by accident detection and control action against the transient conditions of plant. This paper presents the analysis of PPS reliability and the formal problem statement about optimal redundancy based on the reliability of PPS. And the optimization problem is solved by genetic algorithm. The genetic algorithm is a useful tool to solve the problems, in the case of large searching, complex gradient, existence local minimum. The effectiveness of the proposed optimization technique is proved by the target reliability of one channel of PPS, using the failure rate based on the MIL-HDBK-217

  18. Dependent systems reliability estimation by structural reliability approach

    DEFF Research Database (Denmark)

    Kostandyan, Erik; Sørensen, John Dalsgaard

    2014-01-01

    Estimation of system reliability by classical system reliability methods generally assumes that the components are statistically independent, thus limiting its applicability in many practical situations. A method is proposed for estimation of the system reliability with dependent components, where...... the leading failure mechanism(s) is described by physics of failure model(s). The proposed method is based on structural reliability techniques and accounts for both statistical and failure effect correlations. It is assumed that failure of any component is due to increasing damage (fatigue phenomena...... identification. Application of the proposed method can be found in many real world systems....

  19. 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

  20. Electrical system design and reliability at Ontario Hydro nuclear generating stations

    Energy Technology Data Exchange (ETDEWEB)

    Royce, C. J. [Ontario Hydro, 700 University Avenue, Toronto, Ontario M5G 1X6 (Canada)

    1986-02-15

    This paper provides an overview of design practice and the predicted and actual reliability of electrical station service Systems at Ontario Nuclear Generating Stations. Operational experience and licensing changes have indicated the desirability of improving reliability in certain instances. For example, the requirement to start large emergency coolant injection pumps resulted in the turbine generator units in a multi-unit station being used as a back-up power supply. Results of reliability analyses are discussed. To mitigate the effects of common mode events Ontario Hydro adopted a 'two group' approach to the design of safety related Systems. This 'two group' approach is reviewed and a single fully environmentally qualified standby power supply is proposed for future use. (author)

  1. 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

  2. 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

  3. 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)

  4. Three suggestions on the definition of terms for the safety and reliability analysis of digital systems

    International Nuclear Information System (INIS)

    Kim, Man Cheol; Smidts, Carol S.

    2015-01-01

    As digital instrumentation and control systems are being progressively introduced into nuclear power plants, a growing number of related technical issues are coming to light needing to be resolved. As a result, an understanding of relevant terms and basic concepts becomes increasingly important. Under the framework of the OECD/NEA WGRISK DIGREL Task Group, the authors were involved in reviewing definitions of terms forming the supporting vocabulary for addressing issues related to the safety and reliability analysis of digital instrumentation and control (SRA of DI and C). These definitions were extracted from various standards regulating the disciplines that form the technical and scientific basis of SRA DI and C. The authors discovered that different definitions are provided by different standards within a common discipline and used differently across various disciplines. This paper raises the concern that a common understanding of terms and basic concepts has not yet been established to address the very specific technical issues facing SRA DI and C. Based on the lessons learned from the review of the definitions of interest and the analysis of dependency relationships existing between these definitions, this paper establishes a set of recommendations for the development of a consistent terminology for SRA DI and C. - Highlights: ●We reviewed definitions of terms used in reliability analysis of digital systems. ●Different definitions are provided by different standards within a common discipline. ●Acyclic and cyclic structures of dependency in defining terms are compared. ●Three recommendations for the development of a consistent terminology provided

  5. LED system reliability

    NARCIS (Netherlands)

    Driel, W.D. van; Yuan, C.A.; Koh, S.; Zhang, G.Q.

    2011-01-01

    This paper presents our effort to predict the system reliability of Solid State Lighting (SSL) applications. A SSL system is composed of a LED engine with micro-electronic driver(s) that supplies power to the optic design. Knowledge of system level reliability is not only a challenging scientific

  6. A high reliability oxygen deficiency monitoring system

    International Nuclear Information System (INIS)

    Parry, R.; Claborn, G.; Haas, A.; Landis, R.; Page, W.; Smith, J.

    1993-05-01

    The escalating use of cryogens at national laboratories in general and accelerators in particular, along with the increased emphasis placed on personnel safety, mandates the development and installation of oxygen monitoring systems to insure personnel safety in the event of a cryogenic leak. Numerous vendors offer oxygen deficiency monitoring systems but fail to provide important features and/or flexibility. This paper describes a unique oxygen monitoring system developed for the Magnet Test Laboratory (MTL) at the Superconducting Super Collider Laboratory (SSCL). Features include: high reliability, oxygen cell redundancy, sensor longevity, simple calibration, multiple trip points, offending sensor audio and visual indication, global alarms for building evacuation, local and remote analog readout, event and analog data logging, EMAIL event notification, phone line voice status system, and multi-drop communications network capability for reduced cable runs. Of particular importance is the distributed topology of the system which allows it to operate in a stand-alone configuration or to communicate with a host computer. This flexibility makes it ideal for small applications such as a small room containing a cryogenic dewar, as well as larger systems which monitor many offices and labs in several buildings

  7. A high reliability oxygen deficiency monitoring system

    International Nuclear Information System (INIS)

    Parry, R.; Claborn, G.; Haas, A.; Landis, R.; Page, W.; Smith, J.

    1993-01-01

    The escalating use of cryogens at national laboratories in general and accelerators in particular, along with the increased emphasis placed on personnel safety, mandates the development and installation of oxygen monitoring systems to insure personnel safety in the event of a cryogenic leak. Numerous vendors offer oxygen deficiency monitoring systems but fail to provide important features and/or flexibility. This paper describes a unique oxygen monitoring system developed for the Magnet Test Laboratory (MTL) at the Superconducting Super Collider Laboratory (SSCL). Features include: high reliability, oxygen cell redundancy, sensor longevity, simple calibration, multiple trip points, offending sensor audio and visual indication, global alarms for building evacuation, local and remote analog readout, event and analog data logging, EMAIL event notification, phone line voice status system, and multi-drop communications network capability for reduced cable runs. Of particular importance is the distributed topology of the system which allows it to operate in a stand-alone configuration or to communicate with a host computer. This flexibility makes it ideal for small applications such as a small room containing a cryogenic dewar, as well as larger systems which monitor many offices and labs in several buildings

  8. Automated Flight Safety Inference Engine (AFSIE) System, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to develop an innovative Autonomous Flight Safety Inference Engine (AFSIE) system to autonomously and reliably terminate the flight of an errant launch...

  9. Technical reliability of geological disposal for high-level radioactive wastes in Japan. The second progress report. Part 3. Safety assessment for geological disposal systems

    International Nuclear Information System (INIS)

    1999-11-01

    Based on the Advisory Committee Report on Nuclear Fuel Cycle Backend Policy submitted to the Japanese Government in 1997, JNC documents the progress of research and development program in the form of the second progress report (the first one published in 1992). It summarizes an evaluation of the technical reliability and safety of the geological disposal concept for high-level radioactive wastes (HLW) in Japan. The present document, the part 3 of the progress report, concerns safety assessment for geological disposal systems definitely introduced in part 1 and 2 of this series and consists of 9 chapters. Chapter I concerns the methodology for safety assessment while Chapter II deals with diversity and uncertainty about the scenario, the adequate model and the required data of the systems above. Chapter III summarizes the components of the geological disposal system. Chapter IV refers to the relationship between radioactive wastes and human life through groundwater, i.e. nuclide migration. In Chapter V is made a reference case which characterizes the geological environmental data using artificial barrier specifications. (Ohno. S.)

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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)

  15. Structural reliability of atomic power plant

    International Nuclear Information System (INIS)

    Klemin, A.I.; Polyakov, E.F.

    1980-01-01

    In 1978 the first specialized technical manual ''Technique of Calculating the Structural Reliability of an Atomic Power Plant and Its Systems in the Design Stage'' was developed. The present article contains information about the main characteristics and capabilities of the manual. The manual gives recommendations concerning the calculations of the reliability of such specific systems as the reactor control and safety system, the system of instrumentation and automatic control, and safety systems. 2 refs

  16. A study on maintenance reliability allocation of urban transit brake system using hybrid neuro-genetic technique

    International Nuclear Information System (INIS)

    Bae, Chul Ho; Kim, Hyun Jun; Lee, Jung Hwan; Suh, Myung Won; Chu, Yul

    2007-01-01

    For reasonable establishing of maintenance strategies, safety security and cost limitation must be considered at the same time. In this paper, the concept of system reliability introduces and optimizes as the key of reasonable maintenance strategies. This study aims at optimizing component's reliability that satisfies the target reliability of brake system in the urban transit. First of all, constructed reliability evaluation system is used to predict and analyze reliability. This data is used for the optimization. To identify component reliability in a system, a method is presented in this paper which uses hybrid neuro-genetic technique. Feed-forward multi-layer neural networks trained by back propagation are used to find out the relationship between component reliability (input) and system reliability (output) of a structural system. The inverse problem can be formulated by using neural network. Genetic algorithm is used to find the minimum square error. Finally, this paper presents reasonable maintenance cycle of urban transit brake system by using optimal system reliability

  17. Determination of performance criteria of safety systems in a nuclear power plant via simulated annealing optimization method

    International Nuclear Information System (INIS)

    Jung, Woo Sik

    1993-02-01

    This study presents and efficient methodology that derives design alternatives and performance criteria of safety functions/systems in commercial nuclear power plants. Determination of design alternatives and intermediate-level performance criteria is posed as a reliability allocation problem. The reliability allocation is performed for determination of reliabilities of safety functions/systems from top-level performance criteria. The reliability allocation is a very difficult multi objective optimization problem (MOP) as well as a global optimization problem with many local minima. The weighted Chebyshev norm (WCN) approach in combination with an improved Metropolis algorithm of simulated annealing is developed and applied to the reliability allocation problem. The hierarchy of probabilistic safety criteria (PSC) may consist of three levels, which ranges from the overall top level (e.g., core damage frequency, acute fatality and latent cancer fatality) through the interlnediate level (e.g., unavailiability of safety system/function) to the low level (e.g., unavailability of components, component specifications or human error). In order to determine design alternatives of safety functions/systems and the intermediate-level PSC, the reliability allocation is performed from the top-level PSC. The intermediated level corresponds to an objective space and the top level is related to a risk space. The reliability allocation is performed by means of a concept of two-tier noninferior solutions in the objective and risk spaces within the top-level PSC. In this study, two kinds of towtier noninferior solutions are defined: intolerable intermediate-level PSC and desirable design alternatives of safety functions/systems that are determined from Sets 1 and 2, respectively. Set 1 is obtained by maximizing simultaneously not only safety function/system unavailabilities but also risks. Set 1 reflects safety function/system unavailabilities in the worst case. Hence, the

  18. Evaluating software for safety systems in nuclear power plants

    International Nuclear Information System (INIS)

    Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.

    1994-01-01

    In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  20. 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

  1. 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....

  2. The improvement of the connections reliability in a local communication system

    International Nuclear Information System (INIS)

    Buisson, J.; Sanchis, P.

    1986-01-01

    The local communication systems used in the management of the industrial processes must fulfil more severe conditions concerning the reliability and safety than those used in bureaucratics. After mentioning the specific requirements the solutions for the realization of a local industrial communication system are presented. Studies to develop an optical fibers coupling system for a linear network used for the transmission in a hostile medium are mentioned

  3. Reliability analysis of emergency decay heat removal system of nuclear ship under various accident conditions

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi

    1984-01-01

    A reliability analysis is given for the emergency decay heat removal system of the Nuclear Ship ''Mutsu'' and the emergency sea water cooling system of the Nuclear Ship ''Savannah'', under ten typical nuclear ship accident conditions. Basic event probabilities under these accident conditions are estimated from literature survey. These systems of Mutsu and Savannah have almost the same reliability under the normal condition. The dispersive arrangement of a system is useful to prevent the reduction of the system reliability under the condition of an accident restricted in one room. As for the reliability of these two systems under various accident conditions, it is seen that the configuration and the environmental condition of a system are two main factors which determine the reliability of the system. Furthermore, it was found that, for the evaluation of the effectiveness of safety system of a nuclear ship, it is necessary to evaluate its reliability under various accident conditions. (author)

  4. 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.)

  5. Reliability and availability requirements analysis for DEMO: fuel cycle system

    International Nuclear Information System (INIS)

    Pinna, T.; Borgognoni, F.

    2015-01-01

    The Demonstration Power Plant (DEMO) will be a fusion reactor prototype designed to demonstrate the capability to produce electrical power in a commercially acceptable way. Two of the key elements of the engineering development of the DEMO reactor are the definitions of reliability and availability requirements (or targets). The availability target for a hypothesized Fuel Cycle has been analysed as a test case. The analysis has been done on the basis of the experience gained in operating existing tokamak fusion reactors and developing the ITER design. Plant Breakdown Structure (PBS) and Functional Breakdown Structure (FBS) related to the DEMO Fuel Cycle and correlations between PBS and FBS have been identified. At first, a set of availability targets has been allocated to the various systems on the basis of their operating, protection and safety functions. 75% and 85% of availability has been allocated to the operating functions of fuelling system and tritium plant respectively. 99% of availability has been allocated to the overall systems in executing their safety functions. The chances of the systems to achieve the allocated targets have then been investigated through a Failure Mode and Effect Analysis and Reliability Block Diagram analysis. The following results have been obtained: 1) the target of 75% for the operations of the fuelling system looks reasonable, while the target of 85% for the operations of the whole tritium plant should be reduced to 80%, even though all the tritium plant systems can individually reach quite high availability targets, over 90% - 95%; 2) all the DEMO Fuel Cycle systems can reach the target of 99% in accomplishing their safety functions. (authors)

  6. 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)

  7. Multidisciplinary System Reliability Analysis

    Science.gov (United States)

    Mahadevan, Sankaran; Han, Song; Chamis, Christos C. (Technical Monitor)

    2001-01-01

    The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code, developed under the leadership of NASA Glenn Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multidisciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

  8. The Intelligent Safety System: could it introduce complex computing into CANDU shutdown systems

    International Nuclear Information System (INIS)

    Hall, J.A.; Hinds, H.W.; Pensom, C.F.; Barker, C.J.; Jobse, A.H.

    1984-07-01

    The Intelligent Safety System is a computerized shutdown system being developed at the Chalk River Nuclear Laboratories (CRNL) for future CANDU nuclear reactors. It differs from current CANDU shutdown systems in both the algorithm used and the size and complexity of computers required to implement the concept. This paper provides an overview of the project, with emphasis on the computing aspects. Early in the project several needs leading to an introduction of computing complexity were identified, and a computing system that met these needs was conceived. The current work at CRNL centers on building a laboratory demonstration of the Intelligent Safety System, and evaluating the reliability and testability of the concept. Some fundamental problems must still be addressed for the Intelligent Safety System to be acceptable to a CANDU owner and to the regulatory authorities. These are also discussed along with a description of how the Intelligent Safety System might solve these problems

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-12-15

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

  10. Guidelines for reliability analysis of digital systems in PSA context. Phase 1 status report

    International Nuclear Information System (INIS)

    Authen, S.; Larsson, J.; Bjoerkman, K.; Holmberg, J.-E.

    2010-12-01

    Digital protection and control systems are appearing as upgrades in older nuclear power plants (NPPs) and are commonplace in new NPPs. To assess the risk of NPP operation and to determine the risk impact of digital system upgrades on NPPs, quantitative reliability models are needed for digital systems. Due to the many unique attributes of these systems, challenges exist in systems analysis, modeling and in data collection. Currently there is no consensus on reliability analysis approaches. Traditional methods have clearly limitations, but more dynamic approaches are still in trial stage and can be difficult to apply in full scale probabilistic safety assessments (PSA). The number of PSAs worldwide including reliability models of digital I and C systems are few. A comparison of Nordic experiences and a literature review on main international references have been performed in this pre-study project. The study shows a wide range of approaches, and also indicates that no state-of-the-art currently exists. The study shows areas where the different PSAs agree and gives the basis for development of a common taxonomy for reliability analysis of digital systems. It is still an open matter whether software reliability needs to be explicitly modelled in the PSA. The most important issue concerning software reliability is proper descriptions of the impact that software-based systems has on the dependence between the safety functions and the structure of accident sequences. In general the conventional fault tree approach seems to be sufficient for modelling reactor protection system kind of functions. The following focus areas have been identified for further activities: 1. Common taxonomy of hardware and software failure modes of digital components for common use 2. Guidelines regarding level of detail in system analysis and screening of components, failure modes and dependencies 3. Approach for modelling of CCF between components (including software). (Author)

  11. Guidelines for reliability analysis of digital systems in PSA context. Phase 1 status report

    Energy Technology Data Exchange (ETDEWEB)

    Authen, S.; Larsson, J. (Risk Pilot AB, Stockholm (Sweden)); Bjoerkman, K.; Holmberg, J.-E. (VTT, Helsingfors (Finland))

    2010-12-15

    Digital protection and control systems are appearing as upgrades in older nuclear power plants (NPPs) and are commonplace in new NPPs. To assess the risk of NPP operation and to determine the risk impact of digital system upgrades on NPPs, quantitative reliability models are needed for digital systems. Due to the many unique attributes of these systems, challenges exist in systems analysis, modeling and in data collection. Currently there is no consensus on reliability analysis approaches. Traditional methods have clearly limitations, but more dynamic approaches are still in trial stage and can be difficult to apply in full scale probabilistic safety assessments (PSA). The number of PSAs worldwide including reliability models of digital I and C systems are few. A comparison of Nordic experiences and a literature review on main international references have been performed in this pre-study project. The study shows a wide range of approaches, and also indicates that no state-of-the-art currently exists. The study shows areas where the different PSAs agree and gives the basis for development of a common taxonomy for reliability analysis of digital systems. It is still an open matter whether software reliability needs to be explicitly modelled in the PSA. The most important issue concerning software reliability is proper descriptions of the impact that software-based systems has on the dependence between the safety functions and the structure of accident sequences. In general the conventional fault tree approach seems to be sufficient for modelling reactor protection system kind of functions. The following focus areas have been identified for further activities: 1. Common taxonomy of hardware and software failure modes of digital components for common use 2. Guidelines regarding level of detail in system analysis and screening of components, failure modes and dependencies 3. Approach for modelling of CCF between components (including software). (Author)

  12. Adoption of digital safety protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Z.

    1998-01-01

    The application of micro-processor-based digital controllers has been widely propagated among various industries in recent years. While in the nuclear power plant industry, the application of them has also been expanding gradually starting from non-safety related systems, taking advantage of their reliability and maintainability over the conventional analog devices. Based on the careful study of the feasibility of digital controllers to the safety protection system, the Tokyo Electric Power Company proposed on May 1989 the adoption of digital controllers to the safety protection system in the Application for Permission of Establishment of Kashiwazaki-Kariwa units 6 and 7 (ABWR-1350Mwe each). MITI, Ministry of International Trade and Industry, the Japanese regulatory body for electric power generating facilities, had approved this application after careful review. This paper describes a series of supporting activities leading to the MITI's approval of the digital safety protection system and the MITI's licensing activities. (author)

  13. Reliability and Failure in NASA Missions: Blunders, Normal Accidents, High Reliability, Bad Luck

    Science.gov (United States)

    Jones, Harry W.

    2015-01-01

    NASA emphasizes crew safety and system reliability but several unfortunate failures have occurred. The Apollo 1 fire was mistakenly unanticipated. After that tragedy, the Apollo program gave much more attention to safety. The Challenger accident revealed that NASA had neglected safety and that management underestimated the high risk of shuttle. Probabilistic Risk Assessment was adopted to provide more accurate failure probabilities for shuttle and other missions. NASA's "faster, better, cheaper" initiative and government procurement reform led to deliberately dismantling traditional reliability engineering. The Columbia tragedy and Mars mission failures followed. Failures can be attributed to blunders, normal accidents, or bad luck. Achieving high reliability is difficult but possible.

  14. Hawaii Electric System Reliability

    Energy Technology Data Exchange (ETDEWEB)

    Loose, Verne William [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Silva Monroy, Cesar Augusto [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2012-08-01

    This report addresses Hawaii electric system reliability issues; greater emphasis is placed on short-term reliability but resource adequacy is reviewed in reference to electric consumers’ views of reliability “worth” and the reserve capacity required to deliver that value. The report begins with a description of the Hawaii electric system to the extent permitted by publicly available data. Electrical engineering literature in the area of electric reliability is researched and briefly reviewed. North American Electric Reliability Corporation standards and measures for generation and transmission are reviewed and identified as to their appropriateness for various portions of the electric grid and for application in Hawaii. Analysis of frequency data supplied by the State of Hawaii Public Utilities Commission is presented together with comparison and contrast of performance of each of the systems for two years, 2010 and 2011. Literature tracing the development of reliability economics is reviewed and referenced. A method is explained for integrating system cost with outage cost to determine the optimal resource adequacy given customers’ views of the value contributed by reliable electric supply. The report concludes with findings and recommendations for reliability in the State of Hawaii.

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

    International Nuclear Information System (INIS)

    Snaith, E.R.

    1975-01-01

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

  16. 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

  17. 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.

  18. Systems reliability Benchmark exercise part 2-Contributions by the participants

    International Nuclear Information System (INIS)

    Amendola, A.

    1986-01-01

    The report describes aims, rules and results of the Systems Reliability Benchmark Exercise, which has been performed in order to assess methods and procedures for reliability analysis of complex systems and involved a large number of European organizations active in NPP safety evaluation. The exercise included both qualitative and quantitative methods and was structured in such a way that separation of the effects of uncertainties in modelling and in data on the overall spread was made possible. This second part of the report is devoted to the documentation of the single contributions by the participant teams (Swedish, GRS, ENEA, NIRA and ENEL, EWE, EdF, Risoe, KWU/IA, ECN, KEMA/KUL, and Framatome contributions)

  19. Reliability And Maintenance Analysis Of CCTV Systems Used In Rail Transport

    OpenAIRE

    Siergiejczyk Mirosław; Paś Jacek; Rosiński Adam

    2015-01-01

    CCTV systems are widely used across plethora of industrial areas including transport, where their function is to support transport telematics systems. Among others, they are used to ensure travel safety. This paper presented a reliability and maintenance analysis of CCTV. It led to building a relationships graph and then Chapman–Kolmogorov system of equations was derived to describe it. Drawing on those equations, relationships for calculating probability of system staying in state of full ab...

  20. Evaluation of implementation an Integrated Safety and Preventive Maintenance System for Improving of Safety Indexes

    Directory of Open Access Journals (Sweden)

    I mohammadfam

    2014-03-01

    Full Text Available Accident analysis shows that one of the main reasons for accidents is non-integration of maintenance units with safety. Merging these two processes through an integrated system can reduce and or eliminate accidents, diseases, and environmental pollution. These issues lead to improvement in organizational performance, as well. The aim of this study is to design and establish an integrated system for obtaining the aforementioned goal. Integration was carried out at Nirou Moharreke Machine Tools Company via Structured System Analysis & Design Method (SSADM. In order to measure the effectiveness of the system, selected indexes were compared using statistical methods prior and after system establishment. Results show that the accident severity index reduced from 135.46 in 2010, to 43.85 in 2012. Moreover, system effectiveness improved equipment reliability and availability (e.g. reliability of the Pfeiffer Milling machine (P (t>50 increased from 0.89 in 2010, to 0.9 in 2012. This system by forecasting various failures, and planning and designing the required operations for preventing occurrence of these failures, plays an important role in improving safety conditions of equipment, and increasing organizational performance, and is capable of presenting an excellent accident prevention program.

  1. 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.

  2. Passive safety systems for integral reactors

    International Nuclear Information System (INIS)

    Kuul, V.S.; Samoilov, O.B.

    1996-01-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs

  3. Passive safety systems for integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)

    1996-12-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.

  4. High-Reliable PLC RTOS Development and RPS Structure Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, H. S.; Song, D. Y.; Sohn, D. S.; Kim, J. H. [Enersys Co., Daejeon (Korea, Republic of)

    2008-04-15

    One of the KNICS objectives is to develop a platform for Nuclear Power Plant(NPP) I and C(Instrumentation and Control) system, especially plant protection system. The developed platform is POSAFE-Q and this work supports the development of POSAFE-Q with the development of high-reliable real-time operating system(RTOS) and programmable logic device(PLD) software. Another KNICS objective is to develop safety I and C systems, such as Reactor Protection System(RPS) and Engineered Safety Feature-Component Control System(ESF-CCS). This work plays an important role in the structure analysis for RPS. Validation and verification(V and V) of the safety critical software is an essential work to make digital plant protection system highly reliable and safe. Generally, the reliability and safety of software based system can be improved by strict quality assurance framework including the software development itself. In other words, through V and V, the reliability and safety of a system can be improved and the development activities like software requirement specification, software design specification, component tests, integration tests, and system tests shall be appropriately documented for V and V.

  5. High-Reliable PLC RTOS Development and RPS Structure Analysis

    International Nuclear Information System (INIS)

    Sohn, H. S.; Song, D. Y.; Sohn, D. S.; Kim, J. H.

    2008-04-01

    One of the KNICS objectives is to develop a platform for Nuclear Power Plant(NPP) I and C(Instrumentation and Control) system, especially plant protection system. The developed platform is POSAFE-Q and this work supports the development of POSAFE-Q with the development of high-reliable real-time operating system(RTOS) and programmable logic device(PLD) software. Another KNICS objective is to develop safety I and C systems, such as Reactor Protection System(RPS) and Engineered Safety Feature-Component Control System(ESF-CCS). This work plays an important role in the structure analysis for RPS. Validation and verification(V and V) of the safety critical software is an essential work to make digital plant protection system highly reliable and safe. Generally, the reliability and safety of software based system can be improved by strict quality assurance framework including the software development itself. In other words, through V and V, the reliability and safety of a system can be improved and the development activities like software requirement specification, software design specification, component tests, integration tests, and system tests shall be appropriately documented for V and V.

  6. Hawaii electric system reliability.

    Energy Technology Data Exchange (ETDEWEB)

    Silva Monroy, Cesar Augusto; Loose, Verne William

    2012-09-01

    This report addresses Hawaii electric system reliability issues; greater emphasis is placed on short-term reliability but resource adequacy is reviewed in reference to electric consumers' views of reliability %E2%80%9Cworth%E2%80%9D and the reserve capacity required to deliver that value. The report begins with a description of the Hawaii electric system to the extent permitted by publicly available data. Electrical engineering literature in the area of electric reliability is researched and briefly reviewed. North American Electric Reliability Corporation standards and measures for generation and transmission are reviewed and identified as to their appropriateness for various portions of the electric grid and for application in Hawaii. Analysis of frequency data supplied by the State of Hawaii Public Utilities Commission is presented together with comparison and contrast of performance of each of the systems for two years, 2010 and 2011. Literature tracing the development of reliability economics is reviewed and referenced. A method is explained for integrating system cost with outage cost to determine the optimal resource adequacy given customers' views of the value contributed by reliable electric supply. The report concludes with findings and recommendations for reliability in the State of Hawaii.

  7. A novel random-pulser concept for empirical reliability studies of complex systems

    International Nuclear Information System (INIS)

    Priesmeyer, H.G.

    1985-01-01

    The concept of a computer-controlled pseudo-random pulser is described, which is able to produce pulse sequences obeying statistical distributions, used in probability assessments of safety technology. It shall be used in empirical investigations of the reliability of complex systems. (orig.) [de

  8. Calculating system reliability with SRFYDO

    Energy Technology Data Exchange (ETDEWEB)

    Morzinski, Jerome [Los Alamos National Laboratory; Anderson - Cook, Christine M [Los Alamos National Laboratory; Klamann, Richard M [Los Alamos National Laboratory

    2010-01-01

    SRFYDO is a process for estimating reliability of complex systems. Using information from all applicable sources, including full-system (flight) data, component test data, and expert (engineering) judgment, SRFYDO produces reliability estimates and predictions. It is appropriate for series systems with possibly several versions of the system which share some common components. It models reliability as a function of age and up to 2 other lifecycle (usage) covariates. Initial output from its Exploratory Data Analysis mode consists of plots and numerical summaries so that the user can check data entry and model assumptions, and help determine a final form for the system model. The System Reliability mode runs a complete reliability calculation using Bayesian methodology. This mode produces results that estimate reliability at the component, sub-system, and system level. The results include estimates of uncertainty, and can predict reliability at some not-too-distant time in the future. This paper presents an overview of the underlying statistical model for the analysis, discusses model assumptions, and demonstrates usage of SRFYDO.

  9. 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

  10. The socio-technical system and nuclear safety

    International Nuclear Information System (INIS)

    Stefanescu, Petre; Mihailescu, Nicolae; Dragusin, Octavian

    1999-01-01

    In the field of nuclear safety there have been defined notions like 'technical factors' and 'human factors'. The technical factors depend on designing and manufacturing of components/equipment, actually depend on the people's work. The study of human factors consists in analyzing and recommending the terms that allow an individual to be a reliable and safety agent. Accordingly, he/she is placed in working conditions corresponding to human abilities, associating the means of three levels: - designing, i.e. the action upon the technical system and upon work organization; - correction, i.e. the action upon the evolution of the technical system and organizing; - formation/training, i.e. action upon operators. The paper presents a characterization of the socio-technical system and on this basis discusses the issue of individual adjustment to the socio-technical system and reciprocally, the issue of the socio-technical system adjustment to the individual. Concepts as: ergonomics, physical medium, man/machine interface and support of the operator, man/machine task sharing, the work organizing are put in relation with the central subject, the nuclear safety

  11. Spallation Neutron Source Accelerator Facility Target Safety and Non-safety Control Systems

    International Nuclear Information System (INIS)

    Battle, Ronald E.; DeVan, B.; Munro, John K. Jr.

    2006-01-01

    The Spallation Neutron Source (SNS) is a proton accelerator facility that generates neutrons for scientific researchers by spallation of neutrons from a mercury target. The SNS became operational on April 28, 2006, with first beam on target at approximately 200 W. The SNS accelerator, target, and conventional facilities controls are integrated by standardized hardware and software throughout the facility and were designed and fabricated to SNS conventions to ensure compatibility of systems with Experimental Physics Integrated Control System (EPICS). ControlLogix Programmable Logic Controllers (PLCs) interface to instruments and actuators, and EPICS performs the high-level integration of the PLCs such that all operator control can be accomplished from the Central Control room using EPICS graphical screens that pass process variables to and from the PLCs. Three active safety systems were designed to industry standards ISA S84.01 and IEEE 603 to meet the desired reliability for these safety systems. The safety systems protect facility workers and the environment from mercury vapor, mercury radiation, and proton beam radiation. The facility operators operated many of the systems prior to beam on target and developed the operating procedures. The safety and non-safety control systems were tested extensively prior to beam on target. This testing was crucial to identify wiring and software errors and failed components, the result of which was few problems during operation with beam on target. The SNS has continued beam on target since April to increase beam power, check out the scientific instruments, and continue testing the operation of facility subsystems

  12. Failure database and tools for wind turbine availability and reliability analyses. The application of reliability data for selected wind turbines

    DEFF Research Database (Denmark)

    Kozine, Igor; Christensen, P.; Winther-Jensen, M.

    2000-01-01

    The objective of this project was to develop and establish a database for collecting reliability and reliability-related data, for assessing the reliability of wind turbine components and subsystems and wind turbines as a whole, as well as for assessingwind turbine availability while ranking the ...... similar safety systems. The database was established with Microsoft Access DatabaseManagement System, the software for reliability and availability assessments was created with Visual Basic....... the contributions at both the component and system levels. The project resulted in a software package combining a failure database with programs for predicting WTB availability and the reliability of all thecomponents and systems, especially the safety system. The report consists of a description of the theoretical......The objective of this project was to develop and establish a database for collecting reliability and reliability-related data, for assessing the reliability of wind turbine components and subsystems and wind turbines as a whole, as well as for assessingwind turbine availability while ranking...

  13. State-of-art report on digital I and C system reliability issues for nuclear power plants

    International Nuclear Information System (INIS)

    Hwang, In Koo; Lee, Dong Gyoung; Cha, Kyung Ho; Kwon, Kee Choon; Wood, Richard T.

    2000-01-01

    As the instrumentation and control (Iand C) equipment suppliers tend to provide digital components rather than conventional analog type components for instrument and control systems of nuclear power plants(NPPs), it is unavoidable to adopt digital equipment for safety I and C systems as well as non-safety systems. However, the full introduction of digital equipment to I and C systems of nuclear power plants raises several concerns which have not been considered for conventional analog I and C systems. The two major examples of the issues of digital systems are electromagnetic compatibility (EMC) and software reliability. KAERI invited a technical expert, Dr. Richard T. Wood, from Oak Ridge National Laboratory (ORNL) in Unites States and held seminars to recognize the state-of-art of the above issues and to share the information on techniques dealing with the problems. Dr. Wood has been working on the development of EMC guidelines and technical basis in using digital equipment for safety systems in nuclear power plants on the sponsorship of US Nuclear Regulatory Commission (NRC). Being based on his statements and discussions during his visit, this report describes technical considerations and issues on digital safety I and C system application in NPPs, EMC methods, environmental effects vulnerable to digital components, reliability assurance methods, etc. (author)

  14. The European reliability data system - ERDS: a state of the art and future developments

    International Nuclear Information System (INIS)

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

    1982-01-01

    In the frame of the Multiannual Nuclear Safety Programme of the Joint Research Centre of the Commisson of the European Communities, a project is being carried out aiming at the creation of a centralized data system collecting and organizing, at European level, information related to the operation of LWRs. The European Reliability Data System ERDS will exploit information already collected in national data systems and information deriving from single reactor sources. The paper describes the development of the four data systems constituting the ERDS: Component Event Data Bank; Abnormal Occurrences Reporting System; Operating Unit Status Report; Generic Reliability Parameter Data Bank

  15. Fundamental study on applicability of resilience index for system safety assessment

    International Nuclear Information System (INIS)

    Suzuki, Masaaki; Demachi, Kazuyuki; Murakami, Kenta

    2015-01-01

    We have developed a new index called Resilience index, which evaluate the reliability of system safety of nuclear power plant under severe accident by considering the capability to recover from the situation the system safety function was lost. In this paper, a detailed evaluation procedure for the Resilience index was described. System safety of a PWR plant under severe accident was then assessed according to the Resilience index concept to discuss applicability of the index. We found that the Resilience index successfully visualize the management capability, and therefore, resilience capability of a nuclear power plant. (author)

  16. Design reliability engineering

    International Nuclear Information System (INIS)

    Buden, D.; Hunt, R.N.M.

    1989-01-01

    Improved design techniques are needed to achieve high reliability at minimum cost. This is especially true of space systems where lifetimes of many years without maintenance are needed and severe mass limitations exist. Reliability must be designed into these systems from the start. Techniques are now being explored to structure a formal design process that will be more complete and less expensive. The intent is to integrate the best features of design, reliability analysis, and expert systems to design highly reliable systems to meet stressing needs. Taken into account are the large uncertainties that exist in materials, design models, and fabrication techniques. Expert systems are a convenient method to integrate into the design process a complete definition of all elements that should be considered and an opportunity to integrate the design process with reliability, safety, test engineering, maintenance and operator training. 1 fig

  17. 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

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

    International Nuclear Information System (INIS)

    Locks, M.O.

    1978-01-01

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

  19. Nuclear safety: operational aspects. 3. Hazard Analysis of Passive Systems

    International Nuclear Information System (INIS)

    Burgazzi, Luciano

    2001-01-01

    Interest has been aroused in recent years regarding the reliability assessment of passive systems being developed by suppliers, industries, utilities, and research organizations that aim at plant safety improvement and substantial simplification in its implementation. The approach to passive systems reliability assessment entails first a detailed system and safety analysis, and failure mode and effect analysis (FMEA) methodology has been chosen to perform the safety analysis at the system level. The FMEA technique allows identification of all potential failure modes in a system to evaluate their effects on the system and to classify them according to their severity; this technique identifies the reliability-critical areas in the system where modifications to the design are required to reduce the probability of failure. The present study concerns passive systems designed for decay heat removal relying on natural circulation that foresee, for the most part, a condenser immersed in a cooling pool. This is to identify and rank by importance the potential hazards related to passive-system equipment and operation that may critically affect the safety or availability of the plant. More specifically, the content of the paper analyzes the isolation condenser (IC) system foreseen for advanced boiling water reactors for removal of excess sensible and core decay heat by natural circulation during isolation transients. This FMEA analysis is the initial step to be accomplished as support for the development of a methodology aimed at the reliability assessment of thermal-hydraulic passive safety systems, providing important input to more detailed quantitative studies employing, for instance, event trees and fault trees or other reliability/availability models. Main purposes of the work are to identify important accident initiators, find out the possible consequences on the plant deriving from component failures, individuate possible causes, identify mitigating features and

  20. Passive components of NPP safety-related systems

    International Nuclear Information System (INIS)

    Ionaytis Romuald, R.; Bubnova Tatyana, A.

    2005-01-01

    This paper presents a new passive components with having drives: fast-response cutoff valves; modular actuators with opposite cocking pneumatic drives and actuation spring drives; voting electromagnetic valve units for control of pneumatic drives; passive initiators of actuation; visual diagnostics . All these devices have been developed and tested at mock-ups. This paper presents also the following direct-action passive safety components: modular pressure-relief safety valves; pilot safety valves with passive action; check valves with remote position indicator and after-tightening; modular inserts for limiting emergency coolant flow; vortex rectifier; critical weld fasteners; gas-liquid valves; fast-removable seal assembly; seal spring loaders; grooves for increasing hydraulic resistance. Replacement of active safety system components for passive ones improves the general reliability NPP by 1.5 or 2 orders of magnitudes. (authors)

  1. 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

  2. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  3. The computer vision in the service of safety and reliability in steam generators inspection services

    International Nuclear Information System (INIS)

    Pineiro Fernandez, P.; Garcia Bueno, A.; Cabrera Jordan, E.

    2012-01-01

    The actual computational vision has matured very quickly in the last ten years by facilitating new developments in various areas of nuclear application allowing to automate and simplify processes and tasks, instead or in collaboration with the people and equipment efficiently. The current computer vision (more appropriate than the artificial vision concept) provides great possibilities of also improving in terms of the reliability and safety of NPPS inspection systems.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  6. 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)

  7. Designing high availability systems DFSS and classical reliability techniques with practical real life examples

    CERN Document Server

    Taylor, Zachary

    2014-01-01

    A practical, step-by-step guide to designing world-class, high availability systems using both classical and DFSS reliability techniques Whether designing telecom, aerospace, automotive, medical, financial, or public safety systems, every engineer aims for the utmost reliability and availability in the systems he, or she, designs. But between the dream of world-class performance and reality falls the shadow of complexities that can bedevil even the most rigorous design process. While there are an array of robust predictive engineering tools, there has been no single-source guide to understan

  8. 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

  9. 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

  10. Flood risk and economically optimal safety targets for coastal flood defense systems

    NARCIS (Netherlands)

    Dupuits, E.J.C.; Schweckendiek, T.

    2015-01-01

    A front defense can improve the reliability of a rear defense in a coastal flood defense system. The influence of this interdependency on the accompanying economically optimal safety targets of both front and rear defense is investigated. The results preliminary suggest that the optimal safety level

  11. An Integrated Approach of Model checking and Temporal Fault Tree for System Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Koh, Kwang Yong; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2009-10-15

    Digitalization of instruments and control systems in nuclear power plants offers the potential to improve plant safety and reliability through features such as increased hardware reliability and stability, and improved failure detection capability. It however makes the systems and their safety analysis more complex. Originally, safety analysis was applied to hardware system components and formal methods mainly to software. For software-controlled or digitalized systems, it is necessary to integrate both. Fault tree analysis (FTA) which has been one of the most widely used safety analysis technique in nuclear industry suffers from several drawbacks as described in. In this work, to resolve the problems, FTA and model checking are integrated to provide formal, automated and qualitative assistance to informal and/or quantitative safety analysis. Our approach proposes to build a formal model of the system together with fault trees. We introduce several temporal gates based on timed computational tree logic (TCTL) to capture absolute time behaviors of the system and to give concrete semantics to fault tree gates to reduce errors during the analysis, and use model checking technique to automate the reasoning process of FTA.

  12. Rapid Prototyping of the Central Safety System for Nuclear Risk in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Scibile, L. [ITER Organization, 13 - St. Paul lez Durance (France); Ambrosino, G.; De Tommasi, G.; Pironti, A. [Euratom-ENEA-CREATE, Universita di Napoli Federico II, Napoli (Italy)

    2009-07-01

    Full text of publication follows: In the current ITER Baseline design, the Central Safety System for Nuclear Risk (CSS-N) is the safety control system in charge to assure nuclear safety for the plant, personnel and environment. In particular it is envisaged that the CSS shall interface to the plant safety systems for nuclear risk and shall coordinate the individual protection provided by the intervention of these systems by the activation, where required, of additional protections. The design of such a system, together with its implementation, strongly depends on the requirements, particularly in terms of reliability. The CSS-N is a safety critical system, thus its validation and commissioning play a very important role, since the required level of reliability must be demonstrated. In such a scenario, where a new and non-conventional system has to be deployed, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the system requirements, and they will be used to test and validate the control logic. Furthermore these tools can be used to rapid design the safety system and to carry out hardware-in-the-loop (HIL) simulations, which permit to assess the performance of the control hardware against a plant simulator. Both a control system prototype and a safety system oriented plant simulator have been developed to assess first the requirements and then the performance of the CSS-N. In particular the presented SW/HW framework permits to design and verify the CSS protection logics and to test and validate these logics by means of HIL simulations. This work introduces both the prototype and plant simulator architectures, together with the methodology adopted to design and implement these validation tools. (authors)

  13. Analysing supercritical water reactor's (SCWR's) special safety systems using probabilistic tools

    International Nuclear Information System (INIS)

    Ituen, I.; Novog, D.R.

    2011-01-01

    The next generation of reactors, termed Generation IV, has very attractive features -- its superior safety characteristics, high thermal efficiency, and fuel cycle sustainability. A key element of the Generation IV designs is the improvement in safety, which in turn requires improvements in safety system performance and reliability, as well as a reduction in initiating event frequencies. This study compares the response of the systems important to safety in the CANDU-Supercritical Water Reactor to those of the generic CANDU under a main steamline break accident and loss of forced circulation events -- to quantify the improvements in safety for the pre-conceptual CANDU SCWR design. Probabilistic safety analysis is the tool used in this study to test the behavior of the pre- conceptual design during these events. (author)

  14. 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

  15. The use of reliability analysis techniques applied to nuclear power station emergency core cooling systems

    International Nuclear Information System (INIS)

    Danielsen, A.; Snaith, E.R.

    1975-01-01

    A reliability investigation carried out by the Safety and Reliability Services of the UKAEA, and the SSEB, of the essential system/reactor coolant system for a large nuclear power station is described. In AGR type reactors, after all reactor shutdown conditions, it is necessary to restore forced gas circulation and sufficient boiler feed to maintain the heat removal capacity of the boilers. The coolant requirements are provided by several independent mechanical systems of primary coolant fans, feedwater pumps, and valves integrated with electrical power sources, switchgear, and automatic control equipment. Reliability is treated as one aspect of system performance and quantified in terms of failure to meet a specific objective. Based on the reliability performance of the constituent components the optimum system configuration is determined together with the preferred plant operating procedures and maintenance requirements. (author)

  16. Software for computers in the safety systems of nuclear power stations

    International Nuclear Information System (INIS)

    1987-08-01

    This standard includes the safety actuation systems, the safety system support features and the protection systems. The standard provides requirements for each stage of software generation, including design, development, qualification and operation as well as the documentation for each stage of the software generation for the purpose of achieving highly reliable software. The principles applied in developing these requirements include: Best available practice; top-down design methods; modularity; verification of each phase; clear documentation; auditable documents and validation testing. (orig./HP)

  17. RICIS Symposium 1992: Mission and Safety Critical Systems Research and Applications

    Science.gov (United States)

    1992-01-01

    This conference deals with computer systems which control systems whose failure to operate correctly could produce the loss of life and or property, mission and safety critical systems. Topics covered are: the work of standards groups, computer systems design and architecture, software reliability, process control systems, knowledge based expert systems, and computer and telecommunication protocols.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  19. Implementation of a patient safety program at a tertiary health system: A longitudinal analysis of interventions and serious safety events.

    Science.gov (United States)

    Cropper, Douglas P; Harb, Nidal H; Said, Patricia A; Lemke, Jon H; Shammas, Nicolas W

    2018-04-01

    We hypothesize that implementation of a safety program based on high reliability organization principles will reduce serious safety events (SSE). The safety program focused on 7 essential elements: (a) safety rounding, (b) safety oversight teams, (c) safety huddles, (d) safety coaches, (e) good catches/safety heroes, (f) safety education, and (g) red rule. An educational curriculum was implemented focusing on changing high-risk behaviors and implementing critical safety policies. All unusual occurrences were captured in the Midas system and investigated by risk specialists, the safety officer, and the chief medical officer. A multidepartmental committee evaluated these events, and a root cause analysis (RCA) was performed. Events were tabulated and serious safety event (SSE) recorded and plotted over time. Safety success stories (SSSs) were also evaluated over time. A steady drop in SSEs was seen over 9 years. Also a rise in SSSs was evident, reflecting on staff engagement in the program. The parallel change in SSEs, SSSs, and the implementation of various safety interventions highly suggest that the program was successful in achieving its goals. A safety program based on high-reliability organization principles and made a core value of the institution can have a significant positive impact on reducing SSEs. © 2018 American Society for Healthcare Risk Management of the American Hospital Association.

  20. Reliability of Power Electronic Converter Systems

    DEFF Research Database (Denmark)

    -link capacitance in power electronic converter systems; wind turbine systems; smart control strategies for improved reliability of power electronics system; lifetime modelling; power module lifetime test and state monitoring; tools for performance and reliability analysis of power electronics systems; fault...... for advancing the reliability, availability, system robustness, and maintainability of PECS at different levels of complexity. Drawing on the experience of an international team of experts, this book explores the reliability of PECS covering topics including an introduction to reliability engineering in power...... electronic converter systems; anomaly detection and remaining-life prediction for power electronics; reliability of DC-link capacitors in power electronic converters; reliability of power electronics packaging; modeling for life-time prediction of power semiconductor modules; minimization of DC...

  1. Qualification of safety-critical software for digital reactor safety system in nuclear power plants

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo

    2013-01-01

    This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)

  2. Reliability research based experience with systems and events at the Kozloduy NPP units 1-4

    Energy Technology Data Exchange (ETDEWEB)

    Khristova, R; Kaltchev, B; Dimitrov, B [Energoproekt, Sofia (Bulgaria); Nedyalkova, D; Sonev, A [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    An overview of equipment reliability based on operational data of selected safety systems at the Kozloduy NPP is presented. Conclusions are drawn on reliability of the service water system, feed water system, emergency power supply - category 2, emergency high pressure ejection system and spray system. For the units 1-4 all recorded accident protocols in the period 1974-1993 have been processed and the main initiators identified. A list with 39 most frequent initiators of accidents/incidents is compiled. The human-caused errors account for 27% of all events. The reliability characteristics and frequencies have been calculated for all initiating events. It is concluded that there have not been any accidents with consequences for fuel integrity or radioactive release. 14 refs.

  3. Reliability research based experience with systems and events at the Kozloduy NPP units 1-4

    International Nuclear Information System (INIS)

    Khristova, R.; Kaltchev, B.; Dimitrov, B.; Nedyalkova, D.; Sonev, A.

    1995-01-01

    An overview of equipment reliability based on operational data of selected safety systems at the Kozloduy NPP is presented. Conclusions are drawn on reliability of the service water system, feed water system, emergency power supply - category 2, emergency high pressure ejection system and spray system. For the units 1-4 all recorded accident protocols in the period 1974-1993 have been processed and the main initiators identified. A list with 39 most frequent initiators of accidents/incidents is compiled. The human-caused errors account for 27% of all events. The reliability characteristics and frequencies have been calculated for all initiating events. It is concluded that there have not been any accidents with consequences for fuel integrity or radioactive release. 14 refs

  4. Study and application of human reliability analysis for digital human-system interface

    International Nuclear Information System (INIS)

    Jia Ming; Liu Yanzi; Zhang Jianbo

    2014-01-01

    The knowledge of human-orientated abilities and limitations could be used to digital human-system interface (HSI) design by human reliability analysis (HRA) technology. Further, control room system design could achieve the perfect match of man-machine-environment. This research was conducted to establish an integrated HRA method. This method identified HSI potential design flaws which may affect human performance and cause human error. Then a systematic approach was adopted to optimize HSI. It turns out that this method is practical and objective, and effectively improves the safety, reliability and economy of nuclear power plant. This method was applied to CRP1000 projects under construction successfully with great potential. (authors)

  5. Multi-Disciplinary System Reliability Analysis

    Science.gov (United States)

    Mahadevan, Sankaran; Han, Song

    1997-01-01

    The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code developed under the leadership of NASA Lewis Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multi-disciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

  6. Review report: safety and reliability issues on digital instrumentation and control systems in nuclear power plants and United States Nuclear Regulatory Commission's dispositions

    International Nuclear Information System (INIS)

    Watanabe, Norio; Suzudo, Tomoaki

    1998-09-01

    Recently, digital instrumentation and control (I and C) systems have been applied to nuclear power plants (NPPs) in various countries. Introduction of digital I and C systems, however, raises special issues on design, implementation, safety and licensing. Since FY 1997, the Japan Atomic Energy Research Institute (JAERI) has been carrying out a project, Study on Reliability of Digital I and C Systems, which includes extensive reviews of design approaches, technical standards, regulatory processes, especially, in the United States. This report summarizes the results from the study of National Research Council (NRC) and the U.S. Nuclear Regulatory Commission's (USNRC's) responses to the recommendations made by the NRC's study. That study identified six technical key issues (system aspects of digital I and C technology, software quality assurance, common-mode software failure potential, safety and reliability assessment methods, human factors and man-machine interface, dedication of commercial off-the-shelf hardware and software) and two strategic key issues (case-by-case licensing process, adequacy of technical infrastructure) that arise from the introduction of digital I and C technology and then, made recommendations to the USNRC for coping with digital I and C applications. The USNRC responded to each recommendation and showed their own dispositions in which the USNRC agreed with most of the recommendations. In Japan, it is expected that introduction of digital I and C technology is inevitable in NPPs because the vendors are gradually discontinuing support and stocking of analog components. To cope with such situations, there is a need to develop and update the standards and guidelines applicable to digital I and C technology. The key issues and the USNRC's dispositions provided in this report is believed to be useful for developing and updating them. (J.P.N.)

  7. Digital Processor Module Reliability Analysis of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lee, Sang Yong; Jung, Jae Hyun; Kim, Jae Ho; Kim, Sung Hun

    2005-01-01

    The system used in plant, military equipment, satellite, etc. consists of many electronic parts as control module, which requires relatively high reliability than other commercial electronic products. Specially, Nuclear power plant related to the radiation safety requires high safety and reliability, so most parts apply to Military-Standard level. Reliability prediction method provides the rational basis of system designs and also provides the safety significance of system operations. Thus various reliability prediction tools have been developed in recent decades, among of them, the MI-HDBK-217 method has been widely used as a powerful tool for the prediction. In this work, It is explained that reliability analysis work for Digital Processor Module (DPM, control module of SMART) is performed by Parts Stress Method based on MIL-HDBK-217F NOTICE2. We are using the Relex 7.6 of Relex software corporation, because reliability analysis process requires enormous part libraries and data for failure rate calculation

  8. 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

  9. Engaging Employees: The Importance of High-Performance Work Systems for Patient Safety.

    Science.gov (United States)

    Etchegaray, Jason M; Thomas, Eric J

    2015-12-01

    To develop and test survey items that measure high-performance work systems (HPWSs), report psychometric characteristics of the survey, and examine associations between HPWSs and teamwork culture, safety culture, and overall patient safety grade. We reviewed literature to determine dimensions of HPWSs and then asked executives to tell us which dimensions they viewed as most important for safety and quality. We then created a HPWSs survey to measure the most important HPWSs dimensions. We administered an anonymous, electronic survey to employees with direct patient care working at a large hospital system in the Southern United States and looked for linkages between HPWSs, culture, and outcomes. Similarities existed for the HPWS practices viewed as most important by previous researchers and health-care executives. The HPWSs survey was found to be reliable, distinct from safety culture and teamwork culture based on a confirmatory factor analysis, and was the strongest predictor of the extent to which employees felt comfortable speaking up about patient safety problems as well as patient safety grade. We used information from a literature review and executive input to create a reliable and valid HPWSs survey. Future research needs to examine whether HPWSs is associated with additional safety and quality outcomes.

  10. 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)

  11. Fatigue Reliability of Offshore Wind Turbine Systems

    DEFF Research Database (Denmark)

    Marquez-Dominguez, Sergio; Sørensen, John Dalsgaard

    2012-01-01

    of appropriate partial safety factors / fatigue design factors (FDF) for steel substructures of offshore wind turbines (OWTs). The fatigue life is modeled by the SN approach. Design and limit state equations are established based on the accumulated fatigue damage. The acceptable reliability level for optimal...... fatigue design of OWTs is discussed and results for reliability assessment of typical fatigue critical design of offshore steel support structures are presented....

  12. Safety aspect of digital reactor protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Zen-Ichi

    1998-01-01

    It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

  13. Resilience Engineering in Critical Long Term Aerospace Software Systems: A New Approach to Spacecraft Software Safety

    Science.gov (United States)

    Dulo, D. A.

    Safety critical software systems permeate spacecraft, and in a long term venture like a starship would be pervasive in every system of the spacecraft. Yet software failure today continues to plague both the systems and the organizations that develop them resulting in the loss of life, time, money, and valuable system platforms. A starship cannot afford this type of software failure in long journeys away from home. A single software failure could have catastrophic results for the spaceship and the crew onboard. This paper will offer a new approach to developing safe reliable software systems through focusing not on the traditional safety/reliability engineering paradigms but rather by focusing on a new paradigm: Resilience and Failure Obviation Engineering. The foremost objective of this approach is the obviation of failure, coupled with the ability of a software system to prevent or adapt to complex changing conditions in real time as a safety valve should failure occur to ensure safe system continuity. Through this approach, safety is ensured through foresight to anticipate failure and to adapt to risk in real time before failure occurs. In a starship, this type of software engineering is vital. Through software developed in a resilient manner, a starship would have reduced or eliminated software failure, and would have the ability to rapidly adapt should a software system become unstable or unsafe. As a result, long term software safety, reliability, and resilience would be present for a successful long term starship mission.

  14. System Statement of Tasks of Calculating and Providing the Reliability of Heating Cogeneration Plants in Power Systems

    Science.gov (United States)

    Biryuk, V. V.; Tsapkova, A. B.; Larin, E. A.; Livshiz, M. Y.; Sheludko, L. P.

    2018-01-01

    A set of mathematical models for calculating the reliability indexes of structurally complex multifunctional combined installations in heat and power supply systems was developed. Reliability of energy supply is considered as required condition for the creation and operation of heat and power supply systems. The optimal value of the power supply system coefficient F is based on an economic assessment of the consumers’ loss caused by the under-supply of electric power and additional system expences for the creation and operation of an emergency capacity reserve. Rationing of RI of the industrial heat supply is based on the use of concept of technological margin of safety of technological processes. The definition of rationed RI values of heat supply of communal consumers is based on the air temperature level iside the heated premises. The complex allows solving a number of practical tasks for providing reliability of heat supply for consumers. A probabilistic model is developed for calculating the reliability indexes of combined multipurpose heat and power plants in heat-and-power supply systems. The complex of models and calculation programs can be used to solve a wide range of specific tasks of optimization of schemes and parameters of combined heat and power plants and systems, as well as determining the efficiency of various redundance methods to ensure specified reliability of power supply.

  15. Management systems for high reliability organizations. Integration and effectiveness; Managementsysteme fuer Hochzuverlaessigkeitsorganisationen. Integration und Wirksamkeit

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, Michael

    2015-03-09

    The scope of the thesis is the development of a method for improvement of efficient integrated management systems for high reliability organizations (HRO). A comprehensive analysis of severe accident prevention is performed. Severe accident management, mitigation measures and business continuity management are not included. High reliability organizations are complex and potentially dynamic organization forms that can be inherently dangerous like nuclear power plants, offshore platforms, chemical facilities, large ships or large aircrafts. A recursive generic management system model (RGM) was development based on the following factors: systemic and cybernetic Asepcts; integration of different management fields, high decision quality, integration of efficient methods of safety and risk analysis, integration of human reliability aspects, effectiveness evaluation and improvement.

  16. Department of Defense need for a micro-electromechanical systems (MEMS) reliability assessment program

    Science.gov (United States)

    Zunino, James L., III; Skelton, Donald

    2005-01-01

    As the United States (U.S.) Army transforms into a lighter, more lethal, and more agile force, the technologies that support both legacy and emerging weapon systems must decrease in size while increasing in intelligence. Micro-electromechanical systems (MEMS) are one such technology that the Army as well as entire DOD will heavily rely on in achieving these objectives. Current and future military applications of MEMS devices include safety and arming devices, guidance systems, sensors/detectors, inertial measurement units, tracking devices, radio frequency devices, wireless radio frequency identification (RFID), etc. Even though the reliance on MEMS devices has been increasing, there have been no studies performed to determine their reliability and failure mechanisms. Furthermore, no standardized test protocols exist for assessing reliability. Accordingly, the U.S. Army Corrosion Office at Picatinny, NJ has initiated the MEMS Reliability Assessment Program to address this issue.

  17. Research on integrated managing system based on CIMS for nuclear power plant safety

    International Nuclear Information System (INIS)

    Zhou Gang

    2006-01-01

    In order to improve safety, economy and reliability of operation for nuclear power plant (NPP), a novel integrated managing method was proposed based on the ideas of computer and contemporary integrated manufacturing system (CIMS). The application of CIMS to nuclear power plant safety management was researched. In order to design an integrated managing system to meet the needs of NPP safety management, all work related to nuclear safety is divided into different category according to its characters. On basis of this work, general integrated managing system was designed at first. Then subsystems were designed and every subsystem implements a category of nuclear safety management work. All subsystems are independent relatively on the one hand and are interrelated on other hand by global information system. (authors)

  18. Interactive reliability analysis project. FY 80 progress report

    International Nuclear Information System (INIS)

    Rasmuson, D.M.; Shepherd, J.C.

    1981-03-01

    This report summarizes the progress to date in the interactive reliability analysis project. Purpose is to develop and demonstrate a reliability and safety technique that can be incorporated early in the design process. Details are illustrated in a simple example of a reactor safety system

  19. Reliability analysis of nuclear component cooling water system using semi-Markov process model

    International Nuclear Information System (INIS)

    Veeramany, Arun; Pandey, Mahesh D.

    2011-01-01

    Research highlights: → Semi-Markov process (SMP) model is used to evaluate system failure probability of the nuclear component cooling water (NCCW) system. → SMP is used because it can solve reliability block diagram with a mixture of redundant repairable and non-repairable components. → The primary objective is to demonstrate that SMP can consider Weibull failure time distribution for components while a Markov model cannot → Result: the variability in component failure time is directly proportional to the NCCW system failure probability. → The result can be utilized as an initiating event probability in probabilistic safety assessment projects. - Abstract: A reliability analysis of nuclear component cooling water (NCCW) system is carried out. Semi-Markov process model is used in the analysis because it has potential to solve a reliability block diagram with a mixture of repairable and non-repairable components. With Markov models it is only possible to assume an exponential profile for component failure times. An advantage of the proposed model is the ability to assume Weibull distribution for the failure time of components. In an attempt to reduce the number of states in the model, it is shown that usage of poly-Weibull distribution arises. The objective of the paper is to determine system failure probability under these assumptions. Monte Carlo simulation is used to validate the model result. This result can be utilized as an initiating event probability in probabilistic safety assessment projects.

  20. Maintaining scale as a realiable computational system for criticality safety analysis

    International Nuclear Information System (INIS)

    Bowmann, S.M.; Parks, C.V.; Martin, S.K.

    1995-01-01

    Accurate and reliable computational methods are essential for nuclear criticality safety analyses. The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer code system was originally developed at Oak Ridge National Laboratory (ORNL) to enable users to easily set up and perform criticality safety analyses, as well as shielding, depletion, and heat transfer analyses. Over the fifteen-year life of SCALE, the mainstay of the system has been the criticality safety analysis sequences that have featured the KENO-IV and KENO-V.A Monte Carlo codes and the XSDRNPM one-dimensional discrete-ordinates code. The criticality safety analysis sequences provide automated material and problem-dependent resonance processing for each criticality calculation. This report details configuration management which is essential because SCALE consists of more than 25 computer codes (referred to as modules) that share libraries of commonly used subroutines. Changes to a single subroutine in some cases affect almost every module in SCALE exclamation point Controlled access to program source and executables and accurate documentation of modifications are essential to maintaining SCALE as a reliable code system. The modules and subroutine libraries in SCALE are programmed by a staff of approximately ten Code Managers. The SCALE Software Coordinator maintains the SCALE system and is the only person who modifies the production source, executables, and data libraries. All modifications must be authorized by the SCALE Project Leader prior to implementation

  1. Reliability of Beam Loss Monitors System for the Large Hadron Collider

    Science.gov (United States)

    Guaglio, G.; Dehning, B.; Santoni, C.

    2004-11-01

    The employment of superconducting magnets in high energy colliders opens challenging failure scenarios and brings new criticalities for the whole system protection. For the LHC beam loss protection system, the failure rate and the availability requirements have been evaluated using the Safety Integrity Level (SIL) approach. A downtime cost evaluation is used as input for the SIL approach. The most critical systems, which contribute to the final SIL value, are the dump system, the interlock system, the beam loss monitors system and the energy monitor system. The Beam Loss Monitors System (BLMS) is critical for short and intense particle losses, while at medium and higher loss time it is assisted by other systems, such as the quench protection system and the cryogenic system. For BLMS, hardware and software have been evaluated in detail. The reliability input figures have been collected using historical data from the SPS, using temperature and radiation damage experimental data as well as using standard databases. All the data have been processed by reliability software (Isograph). The analysis ranges from the components data to the system configuration.

  2. Reliability of Beam Loss Monitors System for the Large Hadron Collider

    International Nuclear Information System (INIS)

    Guaglio, G.; Dehning, B.; Santoni, C.

    2004-01-01

    The employment of superconducting magnets in high energy colliders opens challenging failure scenarios and brings new criticalities for the whole system protection. For the LHC beam loss protection system, the failure rate and the availability requirements have been evaluated using the Safety Integrity Level (SIL) approach. A downtime cost evaluation is used as input for the SIL approach. The most critical systems, which contribute to the final SIL value, are the dump system, the interlock system, the beam loss monitors system and the energy monitor system. The Beam Loss Monitors System (BLMS) is critical for short and intense particle losses, while at medium and higher loss time it is assisted by other systems, such as the quench protection system and the cryogenic system. For BLMS, hardware and software have been evaluated in detail. The reliability input figures have been collected using historical data from the SPS, using temperature and radiation damage experimental data as well as using standard databases. All the data have been processed by reliability software (Isograph). The analysis ranges from the components data to the system configuration

  3. The achievement and assessment of safety in systems containing software

    International Nuclear Information System (INIS)

    Ball, A.; Dale, C.J.; Butterfield, M.H.

    1986-01-01

    In order to establish confidence in the safe operation of a reactor protection system, there is a need to establish, as far as it is possible, that: (i) the algorithms used are correct; (ii) the system is a correct implementation of the algorithms; and (iii) the hardware is sufficiently reliable. This paper concentrates principally on the second of these, as it applies to the software aspect of the more accurate and complex trip functions to be performed by modern reactor protection systems. In order to engineer safety into software, there is a need to use a development strategy which will stand a high chance of achieving a correct implementation of the trip algorithms. This paper describes three broad methodologies by which it is possible to enhance the integrity of software: fault avoidance, fault tolerance and fault removal. Fault avoidance is concerned with making the software as fault free as possible by appropriate choice of specification, design and implementation methods. A fault tolerant strategy may be advisable in many safety critical applications, in order to guard against residual faults present in the software of the installed system. Fault detection and removal techniques are used to remove as many faults as possible of those introduced during software development. The paper also discusses safety and reliability assessment as it applies to software, outlining the various approaches available. Finally, there is an outline of a research project underway in the UKAEA which is intended to assess methods for developing and testing safety and protection systems involving software. (author)

  4. Technical feasibility and reliability of passive safety systems for nuclear power plants. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1996-12-01

    The meeting provided an overview of the key issues on passive safety. Technical problems which may affect future deployment, and the operating experience of passive systems and components, as well as, definitions of passive safety terms, were discussed. Advantages and disadvantages of passive systems were also highlighted. The philosophy behind different passive safety systems was presented and the range of possibility between fully passive and fully active systems was discussed. Refs, figs, tabs

  5. Technical feasibility and reliability of passive safety systems for nuclear power plants. Proceedings of an advisory group meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    The meeting provided an overview of the key issues on passive safety. Technical problems which may affect future deployment, and the operating experience of passive systems and components, as well as, definitions of passive safety terms, were discussed. Advantages and disadvantages of passive systems were also highlighted. The philosophy behind different passive safety systems was presented and the range of possibility between fully passive and fully active systems was discussed. Refs, figs, tabs.

  6. 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.

  7. Reliability and optimization of structural systems

    International Nuclear Information System (INIS)

    Thoft-Christensen, P.

    1987-01-01

    The proceedings contain 28 papers presented at the 1st working conference. The working conference was organized by the IFIP Working Group 7.5. The proceedings also include 4 papers which were submitted, but for various reasons not presented at the working conference. The working conference was attended by 50 participants from 18 countries. The conference was the first scientific meeting of the new IFIP Working Group 7.5 on 'Reliability and Optimization of Structural Systems'. The purpose of the Working Group 7.5 is to promote modern structural system optimization and reliability theory, to advance international cooperation in the field of structural system optimization and reliability theory, to stimulate research, development and application of structural system optimization and reliability theory, to further the dissemination and exchange of information on reliability and optimization of structural system optimization and reliability theory, and to encourage education in structural system optimization and reliability theory. (orig./HP)

  8. Use of digital computing devices in systems important to safety

    International Nuclear Information System (INIS)

    1986-01-01

    The incorporation of digital computing devices in systems important to safety now is progressing fast in several countries, including Canada, France, Federal Republic of Germany, Japan, USA. There are now reactors with microprocessors in some trip systems. The major functions of those systems are: reactor trip initiation, display, monitoring, testing, re-calibration of detectors. The benefits of moving to a fully computerized shut-down system should be improved reliability, greater flexibility, better man-machine interface, improved testing, higher reactor output and lower overall cost. With the introduction of computer devices in systems important to safety, plant availability and safety are improved because disturbances are treated before they lead to safety action, in this way helping the operator to avoid errors. The Meeting presentations were divided into sessions devoted to the following topics: Needs for the use of digital devices (DCD) in safety important systems (SIS) (5 papers); Problems raised by the integration SIS in the NPP control (7 papers); Description and presentation of DCD of SIS (6 papers); Results of experiences in engineering, manufacture, qualification operation of DCD hardware and software (5 papers). A separate abstract was prepared for each of these papers

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

    International Nuclear Information System (INIS)

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

    1984-10-01

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

  10. Study concerning the power plant control and safety equipment by integrated distributed systems

    International Nuclear Information System (INIS)

    Optea, I.; Oprea, M.; Stanescu, P.

    1995-01-01

    The paper deals with the trends existing in the field of nuclear control and safety equipment and systems, proposing a high-efficiency integrated system. In order to enhance the safety of the plant and reliability of the structure system and components, we present a concept based on the latest computer technology with an open, distributed system, connected by a local area network with high redundancy. A modern conception for the control and safety system is to integrate all the information related to the reactor protection, active engineered safeguard and auxiliary systems parameters, offering a fast flow of information between all the agencies concerned so that situations can be quickly assessed. The integrated distributed control is based on a high performance operating system for realtime applications, flexible enough for transparent networking and modular for demanding configurations. The general design considerations for nuclear reactors instrumentation reliability and testing methods for real-time functions under dynamic regime are presented. Taking into account the fast progress in information technology, we consider the replacement of the old instrumentation of Cernavoda-1 NPP by a modern integrated system as an economical and efficient solution for the next units. (Author) 20 Refs

  11. Reliability-Based Optimization of Series Systems of Parallel Systems

    DEFF Research Database (Denmark)

    Enevoldsen, I.; Sørensen, John Dalsgaard

    1993-01-01

    Reliability-based design of structural systems is considered. In particular, systems where the reliability model is a series system of parallel systems are treated. A sensitivity analysis for this class of problems is presented. Optimization problems with series systems of parallel systems...... optimization of series systems of parallel systems, but it is also efficient in reliability-based optimization of series systems in general....

  12. French power system reliability report 2008

    International Nuclear Information System (INIS)

    Tesseron, J.M.

    2009-06-01

    The reliability of the French power system was fully under control in 2008, despite the power outage in the eastern part of the Provence-Alpes-Cote d'Azur region on November 3, which had been dreaded for several years, since it had not been possible to set up a structurally adequate network. Pursuant to a consultation meeting, the reinforcement solution proposed by RTE was approved by the Minister of Energy, boding well for greater reliability in future. Based on the observations presented in this 2008 Report, RTE's Power System Reliability Audit Mission considers that no new recommendations are needed beyond those expressed in previous reliability reports and during reliability audits. The publication of this yearly report is in keeping with RTE's goal to promote the follow-up over time of the evolution of reliability in its various aspects. RTE thus aims to contribute to the development of reliability culture, by encouraging an improved assessment by the different players (both RTE and network users) of the role they play in building reliability, and by advocating the taking into account of reliability and benchmarking in the European organisations of Transmission System Operators. Contents: 1 - Brief overview of the evolution of the internal and external environment; 2 - Operating situations encountered: climatic conditions, supply / demand balance management, operation of interconnections, management of internal congestion, contingencies affecting the transmission facilities; 3 - Evolution of the reliability reference guide: external reference guide: directives, laws, decrees, etc, ETSO, UCTE, ENTSO-E, contracting contributing to reliability, RTE internal reference guide; 4 - Evolution of measures contributing to reliability in the equipment field: intrinsic performances of components (generating sets, protection systems, operation PLC's, instrumentation and control, automatic frequency and voltage controls, transmission facilities, control systems, load

  13. Human-system safety methods for development of advanced air traffic management systems

    International Nuclear Information System (INIS)

    Nelson, William R.

    1999-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is supporting the National Aeronautics and Space Administration in the development of advanced air traffic management (ATM) systems as part of the Advanced Air Transportation Technologies program. As part of this program INEEL conducted a survey of human-system safety methods that have been applied to complex technical systems, to identify lessons learned from these applications and provide recommendations for the development of advanced ATM systems. The domains that were surveyed included offshore oil and gas, commercial nuclear power, commercial aviation, and military. The survey showed that widely different approaches are used in these industries, and that the methods used range from very high-level, qualitative approaches to very detailed quantitative methods such as human reliability analysis (HRA) and probabilistic safety assessment (PSA). In addition, the industries varied widely in how effectively they incorporate human-system safety assessment in the design, development, and testing of complex technical systems. In spite of the lack of uniformity in the approaches and methods used, it was found that methods are available that can be combined and adapted to support the development of advanced air traffic management systems (author) (ml)

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

    Science.gov (United States)

    Bavuso, S. J.

    1984-01-01

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

  15. Rapid prototyping of the Central Safety System for Nuclear Risk in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Scibile, L. [ITER Organization, CS 90 046, St. Paul-lez-Durance, Cedex (France); Ambrosino, G. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); De Tommasi, G., E-mail: detommas@unina.i [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); Pironti, A. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy)

    2010-07-15

    The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.

  16. Rapid prototyping of the Central Safety System for Nuclear Risk in ITER

    International Nuclear Information System (INIS)

    Scibile, L.; Ambrosino, G.; De Tommasi, G.; Pironti, A.

    2010-01-01

    The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.

  17. Uncertainty analysis methods for estimation of reliability of passive system of VHTR

    International Nuclear Information System (INIS)

    Han, S.J.

    2012-01-01

    An estimation of reliability of passive system for the probabilistic safety assessment (PSA) of a very high temperature reactor (VHTR) is under development in Korea. The essential approach of this estimation is to measure the uncertainty of the system performance under a specific accident condition. The uncertainty propagation approach according to the simulation of phenomenological models (computer codes) is adopted as a typical method to estimate the uncertainty for this purpose. This presentation introduced the uncertainty propagation and discussed the related issues focusing on the propagation object and its surrogates. To achieve a sufficient level of depth of uncertainty results, the applicability of the propagation should be carefully reviewed. For an example study, Latin-hypercube sampling (LHS) method as a direct propagation was tested for a specific accident sequence of VHTR. The reactor cavity cooling system (RCCS) developed by KAERI was considered for this example study. This is an air-cooled type passive system that has no active components for its operation. The accident sequence is a low pressure conduction cooling (LPCC) accident that is considered as a design basis accident for the safety design of VHTR. This sequence is due to a large failure of the pressure boundary of the reactor system such as a guillotine break of coolant pipe lines. The presentation discussed the obtained insights (benefit and weakness) to apply an estimation of reliability of passive system

  18. An Online Risk Monitor System (ORMS) to Increase Safety and Security Levels in Industry

    International Nuclear Information System (INIS)

    Zubair, M; Ur Rahman, Khalil; Ul Hassan, Mehmood

    2013-01-01

    The main idea of this research is to develop an Online Risk Monitor System (ORMS) based on Living Probabilistic Safety Assessment (LPSA). The article highlights the essential features and functions of ORMS. The basic models and modules such as, Reliability Data Update Model (RDUM), running time update, redundant system unavailability update, Engineered Safety Features (ESF) unavailability update and general system update have been described in this study. ORMS not only provides quantitative analysis but also highlights qualitative aspects of risk measures. ORMS is capable of automatically updating the online risk models and reliability parameters of equipment. ORMS can support in the decision making process of operators and managers in Nuclear Power Plants

  19. An Online Risk Monitor System (ORMS) to Increase Safety and Security Levels in Industry

    Science.gov (United States)

    Zubair, M.; Rahman, Khalil Ur; Hassan, Mehmood Ul

    2013-12-01

    The main idea of this research is to develop an Online Risk Monitor System (ORMS) based on Living Probabilistic Safety Assessment (LPSA). The article highlights the essential features and functions of ORMS. The basic models and modules such as, Reliability Data Update Model (RDUM), running time update, redundant system unavailability update, Engineered Safety Features (ESF) unavailability update and general system update have been described in this study. ORMS not only provides quantitative analysis but also highlights qualitative aspects of risk measures. ORMS is capable of automatically updating the online risk models and reliability parameters of equipment. ORMS can support in the decision making process of operators and managers in Nuclear Power Plants.

  20. 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

  1. Review report: safety and reliability issues on digital instrumentation and control systems in nuclear power plants and United States Nuclear Regulatory Commission`s dispositions

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio; Suzudo, Tomoaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-09-01

    Recently, digital instrumentation and control (I and C) systems have been applied to nuclear power plants (NPPs) in various countries. Introduction of digital I and C systems, however, raises special issues on design, implementation, safety and licensing. Since FY 1997, the Japan Atomic Energy Research Institute (JAERI) has been carrying out a project, Study on Reliability of Digital I and C Systems, which includes extensive reviews of design approaches, technical standards, regulatory processes, especially, in the United States. This report summarizes the results from the study of National Research Council (NRC) and the U.S. Nuclear Regulatory Commission`s (USNRC`s) responses to the recommendations made by the NRC`s study. That study identified six technical key issues (system aspects of digital I and C technology, software quality assurance, common-mode software failure potential, safety and reliability assessment methods, human factors and man-machine interface, dedication of commercial off-the-shelf hardware and software) and two strategic key issues (case-by-case licensing process, adequacy of technical infrastructure) that arise from the introduction of digital I and C technology and then, made recommendations to the USNRC for coping with digital I and C applications. The USNRC responded to each recommendation and showed their own dispositions in which the USNRC agreed with most of the recommendations. In Japan, it is expected that introduction of digital I and C technology is inevitable in NPPs because the vendors are gradually discontinuing support and stocking of analog components. To cope with such situations, there is a need to develop and update the standards and guidelines applicable to digital I and C technology. The key issues and the USNRC`s dispositions provided in this report is believed to be useful for developing and updating them. (J.P.N.)

  2. Advances in safety related maintenance

    International Nuclear Information System (INIS)

    2000-03-01

    The maintenance of systems, structures and components in nuclear power plants (NPPs) plays an important role in assuring their safe and reliable operation. Worldwide, NPP maintenance managers are seeking to reduce overall maintenance costs while maintaining or improving the levels of safety and reliability. Thus, the issue of NPP maintenance is one of the most challenging aspects of nuclear power generation. There is a direct relation between safety and maintenance. While maintenance alone (apart from modifications) will not make a plant safer than its original design, deficient maintenance may result in either an increased number of transients and challenges to safety systems or reduced reliability and availability of safety systems. The confidence that NPP structures, systems and components will function as designed is ultimately based on programmes which monitor both their reliability and availability to perform their intended safety function. Because of this, approaches to monitor the effectiveness of maintenance are also necessary. An effective maintenance programme ensures that there is a balance between the improvement in component reliability to be achieved and the loss of component function due to maintenance downtime. This implies that the safety level of an NPP should not be adversely affected by maintenance performed during operation. The nuclear industry widely acknowledges the importance of maintenance in NPP safety and operation and therefore devotes great efforts to develop techniques, methods and tools to aid in maintenance planning, follow-up and optimization, and in assuring the effectiveness of maintenance

  3. Passive safety systems and natural circulation in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2009-11-01

    Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems can eliminate the costs associated with the installation, maintenance, and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

  4. 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

  5. Reliability of electronic systems

    International Nuclear Information System (INIS)

    Roca, Jose L.

    2001-01-01

    Reliability techniques have been developed subsequently as a need of the diverse engineering disciplines, nevertheless they are not few those that think they have been work a lot on reliability before the same word was used in the current context. Military, space and nuclear industries were the first ones that have been involved in this topic, however not only in these environments it is that it has been carried out this small great revolution in benefit of the increase of the reliability figures of the products of those industries, but rather it has extended to the whole industry. The fact of the massive production, characteristic of the current industries, drove four decades ago, to the fall of the reliability of its products, on one hand, because the massively itself and, for other, to the recently discovered and even not stabilized industrial techniques. Industry should be changed according to those two new requirements, creating products of medium complexity and assuring an enough reliability appropriated to production costs and controls. Reliability began to be integral part of the manufactured product. Facing this philosophy, the book describes reliability techniques applied to electronics systems and provides a coherent and rigorous framework for these diverse activities providing a unifying scientific basis for the entire subject. It consists of eight chapters plus a lot of statistical tables and an extensive annotated bibliography. Chapters embrace the following topics: 1- Introduction to Reliability; 2- Basic Mathematical Concepts; 3- Catastrophic Failure Models; 4-Parametric Failure Models; 5- Systems Reliability; 6- Reliability in Design and Project; 7- Reliability Tests; 8- Software Reliability. This book is in Spanish language and has a potentially diverse audience as a text book from academic to industrial courses. (author)

  6. A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods; Um estudo de arquiteturas de hardware para aplicacao em sistemas digitais de protecao de reatores nucleares - metodos de analise de confiabilidade e seguranca

    Energy Technology Data Exchange (ETDEWEB)

    Benko, Pedro Luiz

    1997-07-01

    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)

  7. Software reliability evaluation of digital plant protection system development process using V and V

    International Nuclear Information System (INIS)

    Lee, Na Young; Hwang, Il Soon; Seong, Seung Hwan; Oh, Seung Rok

    2001-01-01

    In the nuclear power industry, digital technology has been introduced recently for the Instrumentation and Control (I and C) of reactor systems. For its application to the safety critical system such as Reactor Protection System(RPS), a reliability assessment is indispensable. Unlike traditional reliability models, software reliability is hard to evaluate, and should be evaluated throughout development lifecycle. In the development process of Digital Plant Protection System(DPPS), the concept of verification and validation (V and V) was introduced to assure the quality of the product. Also, test should be performed to assure the reliability. Verification procedure with model checking is relatively well defined, however, test is labor intensive and not well organized. In this paper, we developed the methodological process of combining the verification with validation test case generation. For this, we used PVS for the table specification and for the theorem proving. As a result, we could not only save time to design test case but also get more effective and complete verification related test case set. Add to this, we could extract some meaningful factors useful for the reliability evaluation both from the V and V and verification combined tests

  8. System 80+ Design and Licensing : Improving Plant Reliability

    International Nuclear Information System (INIS)

    Newman, Robert E.

    1989-01-01

    The U. S. nuclear industry is striving to improve plant reliability and availability through improved plant design, component designs and plant maintenance. In an effort to improve safety and to demonstrate that commercial nuclear power is economically competitive with other energy sources, the utilities, nuclear vendors, architect engineers and constructors, and component suppliers are all participating in an industry-wide effort to develop improved Light Water Reactor (LWR) designs that are based upon the many years of successful LWR operation. In an age when the world faces the environmental pressures of the greenhouse effect and acid rain, electricity generated from nuclear energy must play an increasing role in the energy picture of Korea, the United States and the rest of the world. This paper discusses the plant availability requirement that has been established by the industry-wide effort mentioned above. After briefly describing Combustion Engineering's program for development of the System 80 Plus standard design and the participation of the Korea Advanced Energy Research Institute (KAERI) in the program, the paper then describes the design features that are being incorporated into System 80+. The industry ALRR Program has established a very ambitious criterion of 87% for the plant availability of future nuclear units. To satisfy such a requirement, the next generation of nuclear plants will include a great many design improvements that reflect the hundreds of years of operating experience that we have accrued. C-ESA's System 80+ will include a number of design changes that improve operating margins and make the plant easier to operate and maintain. Not surprisingly, there is a great deal of overlap between improved safety and improved reliability. In the end, our design will satisfy the future needs of the utilities, the regulators, and the public. C-E is very pleased that KAERI is working with US to achieve these important goals

  9. 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

  10. Reliability modeling of safety-critical network communication in a digitalized nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Kim, Hee Eun; Son, Kwang Seop; Shin, Sung Min; Lee, Seung Jun; Kang, Hyun Gook

    2015-01-01

    The Engineered Safety Feature-Component Control System (ESF-CCS), which uses a network communication system for the transmission of safety-critical information from group controllers (GCs) to loop controllers (LCs), was recently developed. However, the ESF-CCS has not been applied to nuclear power plants (NPPs) because the network communication failure risk in the ESF-CCS has yet to be fully quantified. Therefore, this study was performed to identify the potential hazardous states for network communication between GCs and LCs and to develop quantification schemes for various network failure causes. To estimate the risk effects of network communication failures in the ESF-CCS, a fault-tree model of an ESF-CCS signal failure in the containment spray actuation signal condition was developed for the case study. Based on a specified range of periodic inspection periods for network modules and the baseline probability of software failure, a sensitivity study was conducted to analyze the risk effect of network failure between GCs and LCs on ESF-CCS signal failure. This study is expected to provide insight into the development of a fault-tree model for network failures in digital I&C systems and the quantification of the risk effects of network failures for safety-critical information transmission in NPPs. - Highlights: • Network reliability modeling framework for digital I&C system in NPP is proposed. • Hazardous states of network protocol between GC and LC in ESF-CCS are identified. • Fault-tree model of ESF-CCS signal failure in ESF actuation condition is developed. • Risk effect of network failure on ESF-CCS signal failure is analyzed.

  11. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  12. Concept of safety subsystem for RF system for the VINCY Cyclotron; Koncept sigurnosnog podsistema radiofrekventnog sistema ciklotrona VINCY

    Energy Technology Data Exchange (ETDEWEB)

    Spasojevic, S; Djuric, D [Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia)

    1996-07-01

    The concept of the safety subsystem of the RF system of cyclotron VINCY is described. By applying the principle of separation of the control and safety functions and the fail-safe concept, an autonomous and reliable safety subsystem has been designed. A combination of the traditional relay technology, often applied in safety systems, and a modern, industrial PC based, acquisition system resulted into a solution meeting all design requirements. (author)

  13. Integrating reliability analysis and design

    International Nuclear Information System (INIS)

    Rasmuson, D.M.

    1980-10-01

    This report describes the Interactive Reliability Analysis Project and demonstrates the advantages of using computer-aided design systems (CADS) in reliability analysis. Common cause failure problems require presentations of systems, analysis of fault trees, and evaluation of solutions to these. Results have to be communicated between the reliability analyst and the system designer. Using a computer-aided design system saves time and money in the analysis of design. Computer-aided design systems lend themselves to cable routing, valve and switch lists, pipe routing, and other component studies. At EG and G Idaho, Inc., the Applicon CADS is being applied to the study of water reactor safety systems

  14. Methodology for identifying boundaries of systems important to safety in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Therrien, S.; Komljenovic, D.; Therrien, P.; Ruest, C.; Prevost, P.; Vaillancourt, R.

    2007-01-01

    This paper presents a methodology developed to identify the boundaries of the systems important to safety (SIS) at the Gentilly-2 Nuclear Power Plant (NPP), Hydro-Quebec. The SIS boundaries identification considers nuclear safety only. Components that are not identified as important to safety are systematically identified as related to safety. A global assessment process such as WANO/INPO AP-913 'Equipment Reliability Process' will be needed to implement adequate changes in the management rules of those components. The paper depicts results in applying the methodology to the Shutdown Systems 1 and 2 (SDS 1, 2), and to the Emergency Core Cooling System (ECCS). This validation process enabled fine tuning the methodology, performing a better estimate of the effort required to evaluate a system, and identifying components important to safety of these systems. (author)

  15. Natural Circulation in Water Cooled Nuclear Power Plants Phenomena, models, and methodology for system reliability assessments

    International Nuclear Information System (INIS)

    Jose Reyes

    2005-01-01

    In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''

  16. ICAROUS - Integrated Configurable Algorithms for Reliable Operations Of Unmanned Systems

    Science.gov (United States)

    Consiglio, María; Muñoz, César; Hagen, George; Narkawicz, Anthony; Balachandran, Swee

    2016-01-01

    NASA's Unmanned Aerial System (UAS) Traffic Management (UTM) project aims at enabling near-term, safe operations of small UAS vehicles in uncontrolled airspace, i.e., Class G airspace. A far-term goal of UTM research and development is to accommodate the expected rise in small UAS traffic density throughout the National Airspace System (NAS) at low altitudes for beyond visual line-of-sight operations. This paper describes a new capability referred to as ICAROUS (Integrated Configurable Algorithms for Reliable Operations of Unmanned Systems), which is being developed under the UTM project. ICAROUS is a software architecture comprised of highly assured algorithms for building safety-centric, autonomous, unmanned aircraft applications. Central to the development of the ICAROUS algorithms is the use of well-established formal methods to guarantee higher levels of safety assurance by monitoring and bounding the behavior of autonomous systems. The core autonomy-enabling capabilities in ICAROUS include constraint conformance monitoring and contingency control functions. ICAROUS also provides a highly configurable user interface that enables the modular integration of mission-specific software components.

  17. RIO: a program to determine reliability importance and allocate optimal reliability goals

    International Nuclear Information System (INIS)

    Poloski, J.P.

    1978-09-01

    The designer of a nuclear plant must know the plant's associated risk limitations so that he can design the plant accordingly. To design a safety system, he must understand its importance and how it relates to the overall plant risk. The computer program RIO can aid the designer to understand a system's contribution to the plant's overall risk. The methodology developed and presented was sponsored by the Nuclear Research Applications Division of the Department of Energy for use in the Gas Cooled Fast Breeder Reactor (GCFR) Program. The principal motivation behind its development was the need to translate nuclear plants safety goals into reliability goals for systems which make up that plant. The method described herein will make use of the GCFR Accident Initiation and Progression Analyses (AIPA) event trees and other models in order to determine these reliability goals

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

    International Nuclear Information System (INIS)

    Touati, Said; Chennai, Salim; Souli, Aissa

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  20. Safety and security analysis for distributed control system in nuclear power plants

    International Nuclear Information System (INIS)

    Lu Zhigang; Liu Baoxu

    2011-01-01

    The Digital Distributed Control System (DCS) is the core that manages all monitoring and operation tasks in a Nuclear Power Plant (NPP). So, Digital Distributed Control System in Nuclear Power Plant has strict requirements for control and automation device safety and security due to many factors. In this article, factors of safety are analyzed firstly, while placing top priority on reliability, quality of supply and stability have also been carefully considered. In particular, advanced digital and electronic technologies are adopted to maintain sufficient reliability and supervisory capabilities in nuclear power plants. Then, security of networking and information technology have been remarked, several design methodologies considering the security characteristics are suggested. Methods and technologies of this article are being used in testing and evaluation for a real implement of a nuclear power plant in China. (author)

  1. Exploitation examination of reliability of coal dust systems

    International Nuclear Information System (INIS)

    Dojchinovski, Ilija; Trajkovski, Kole

    1997-01-01

    Designers and operators wish is, long, failure free operation at designed parameters of every system. Always we know the system start up time, but we don't know how long this system will operate successfully. Because of that in this article is given a method how, step by step, to determine the reliability of the system. Reliability parameters are obtained from experimental and operational data. When reliability parameters are determined then it is very easy to compare reliability of similar systems, for example excavators, or different systems, such as truck and rubber band transport system. Practical use of the theory of reliability is by purchasing of the systems when manufacturers have to have and present reliability parameters and on this way we can decide which system satisfies our needs regarding the quality-price-reliability. Reliability can be practically used in system operation where: 1) system reliability is maintained with proper start, use and shutdown of the system; 2) a system reliability is maintained with good maintenance organization; 3) a system reliability is maintained with innovations and improvements with final purpose removing of the imperfections experienced through the operation. Reliability is very important parameter in power generation plants. (Author)

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

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor

    2017-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  5. Effect Analysis of Digital I and C Systems on Plant Safety based on Fault-Tree Analysis

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Jung, Wondea

    2014-01-01

    Deterioration and an inadequate supply of components of analog I and C systems have led to inefficient and costly maintenance. Moreover, since the fast evolution of digital technology has enabled more reliable functions to be designed for NPP safety, the transition from analog to digital has been accelerated. Owing to the distinguishable characteristics of digital I and C systems, a reliability analysis of digital systems has become an important element of a probabilistic safety assessment (PSA). Digital I and C systems have unique characteristics such as fault-tolerant techniques and software. However, these features have not been properly considered yet in most NPP PSA models. The effect of digital I and C systems should be evaluated by comparing them to that of analog I and C systems. Before installing a digital I and C system, even though it is expected that the plant safety can be improved through the advantageous features of digital I and C systems, it should be validated whether the total NPP safety is better than analog systems or is the same at least. In this work, the fault-tree (FT) technique, which is most widely used in a PSA, was used to compare the effects of analog and digital I and C systems. From a case study, the results of plant safety were compared. In this work, the effect of a digital RPS was evaluated by comparing it to that of an analog RPS based on the FT models. In the evaluation results, it was observed that digital RPS has a positive effect on reducing the system unavailability. The analysis results can be used for the development of a guide for evaluating digital I and C systems and reliability requirements

  6. Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Son, Kwang Seop; Jung, Wondea; Kang, Hyun Gook

    2017-01-01

    Highlights: • Safety data communication risk assessment framework and quantitative scheme were proposed. • Fault-tree model of ESFAS unavailability due to safety data communication failure was developed. • Safety data link and network risk were assessed based on various ESF-CCS design specifications. • The effect of fault-tolerant algorithm reliability of safety data network on ESFAS unavailability was assessed. - Abstract: As one of the safety-critical systems in nuclear power plants (NPPs), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for the transmission of safety component actuation signals from the group controllers to loop controllers to effectively accommodate various safety-critical field controllers. Since data communication failure risk in the ESF-CCS has yet to be fully quantified, the ESF-CCS employing data communication systems have not been applied in NPPs. This study therefore developed a fault tree model to assess the data link and data network failure-induced unavailability of a system function used to generate an automated control signal for accident mitigation equipment. The current aim is to provide risk information regarding data communication failure in a digital safety feature control system in consideration of interconnection between controllers and the fault-tolerant algorithm implemented in the target system. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to data link and network failure and its risk effect on safety signal generation failure. This study is expected to provide insight into the risk assessment of safety-critical data communication in a digitalized NPP instrumentation and control system.

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

    International Nuclear Information System (INIS)

    Carlson, D.D.

    1983-01-01

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

  8. PV Systems Reliability Final Technical Report.

    Energy Technology Data Exchange (ETDEWEB)

    Lavrova, Olga [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flicker, Jack David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Johnson, Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Armijo, Kenneth Miguel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gonzalez, Sigifredo [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorensen, Neil R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Yang, Benjamin Bing-Yeh [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    The continued exponential growth of photovoltaic technologies paves a path to a solar-powered world, but requires continued progress toward low-cost, high-reliability, high-performance photovoltaic (PV) systems. High reliability is an essential element in achieving low-cost solar electricity by reducing operation and maintenance (O&M) costs and extending system lifetime and availability, but these attributes are difficult to verify at the time of installation. Utilities, financiers, homeowners, and planners are demanding this information in order to evaluate their financial risk as a prerequisite to large investments. Reliability research and development (R&D) is needed to build market confidence by improving product reliability and by improving predictions of system availability, O&M cost, and lifetime. This project is focused on understanding, predicting, and improving the reliability of PV systems. The two areas being pursued include PV arc-fault and ground fault issues, and inverter reliability.

  9. 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

  10. Application of systems engineering techniques (reliability, availability, maintainability, and dollars) to the Gas Centrifuge Enrichment Plant

    International Nuclear Information System (INIS)

    Boylan, J.G.; DeLozier, R.C.

    1982-01-01

    The systems engineering function for the Gas Centrifuge Enrichment Plant (GCEP) covers system requirements definition, analyses, verification, technical reviews, and other system efforts necessary to assure good balance of performance, safety, cost, and scheduling. The systems engineering function will support the design, installation, start-up, and operational phases of GCEP. The principal objectives of the systems engineering function are to: assure that the system requirements of the GCEP process are adequately specified and documented and that due consideration and emphasis are given to all aspects of the project; provide system analyses of the designs as they progress to assure that system requirements are met and that GCEP interfaces are compatible; assist in the definition of programs for the necessary and sufficient verification of GCEP systems; and integrate reliability, maintainability, logistics, safety, producibility, and other related specialties into a total system effort. This paper addresses the GCEP reliability, availability, maintainability, and dollars (RAM dollars) analyses which are the primary systems engineering tools for the development and implementation of trade-off studies. These studies are basic to reaching cost-effective project decisions. The steps necessary to achieve optimum cost-effective design are shown

  11. System safety education focused on flight safety

    Science.gov (United States)

    Holt, E.

    1971-01-01

    The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.

  12. Power system reliability memento; Memento de la surete du systeme electrique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    The reliability memento of the French power system (national power transmission grid) is an educational document which purpose is to point out the role of each one as regards power system operating reliability. This memento was first published in 1999. Extensive changes have taken place since then. The new 2002 edition shows that system operating reliability is as an important subject as ever: 1 - foreword; 2 - system reliability: the basics; 3 - equipment measures taken in order to guarantee the reliability of the system; 4 - organisational and human measures taken to guarantee the reliability of the system; appendix 1 - system operation: basic concepts; appendix 2 - guiding principles governing the reliability of the power system; appendix 3 - international associations of transmission system operators; appendix 4 - description of major incidents.

  13. Reliability evaluation of power systems

    CERN Document Server

    Billinton, Roy

    1996-01-01

    The Second Edition of this well-received textbook presents over a decade of new research in power system reliability-while maintaining the general concept, structure, and style of the original volume. This edition features new chapters on the growing areas of Monte Carlo simulation and reliability economics. In addition, chapters cover the latest developments in techniques and their application to real problems. The text also explores the progress occurring in the structure, planning, and operation of real power systems due to changing ownership, regulation, and access. This work serves as a companion volume to Reliability Evaluation of Engineering Systems: Second Edition (1992).

  14. Integrated system reliability analysis

    DEFF Research Database (Denmark)

    Gintautas, Tomas; Sørensen, John Dalsgaard

    Specific targets: 1) The report shall describe the state of the art of reliability and risk-based assessment of wind turbine components. 2) Development of methodology for reliability and risk-based assessment of the wind turbine at system level. 3) Describe quantitative and qualitative measures...

  15. Reliability Based Optimization of Structural Systems

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard

    1987-01-01

    The optimization problem to design structural systems such that the reliability is satisfactory during the whole lifetime of the structure is considered in this paper. Some of the quantities modelling the loads and the strength of the structure are modelled as random variables. The reliability...... is estimated using first. order reliability methods ( FORM ). The design problem is formulated as the optimization problem to minimize a given cost function such that the reliability of the single elements satisfies given requirements or such that the systems reliability satisfies a given requirement....... For these optimization problems it is described how a sensitivity analysis can be performed. Next, new optimization procedures to solve the optimization problems are presented. Two of these procedures solve the system reliability based optimization problem sequentially using quasi-analytical derivatives. Finally...

  16. 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.

  17. The possibilities of applying a risk-oriented approach to the NPP reliability and safety enhancement problem

    Science.gov (United States)

    Komarov, Yu. A.

    2014-10-01

    An analysis and some generalizations of approaches to risk assessments are presented. Interconnection between different interpretations of the "risk" notion is shown, and the possibility of applying the fuzzy set theory to risk assessments is demonstrated. A generalized formulation of the risk assessment notion is proposed in applying risk-oriented approaches to the problem of enhancing reliability and safety in nuclear power engineering. The solution of problems using the developed risk-oriented approaches aimed at achieving more reliable and safe operation of NPPs is described. The results of studies aimed at determining the need (advisability) to modernize/replace NPP elements and systems are presented together with the results obtained from elaborating the methodical principles of introducing the repair concept based on the equipment technical state. The possibility of reducing the scope of tests and altering the NPP systems maintenance strategy is substantiated using the risk-oriented approach. A probabilistic model for estimating the validity of boric acid concentration measurements is developed.

  18. Recommendations on the use of expert judgment in safety and reliability engineering studies. Two offshore case studies

    International Nuclear Information System (INIS)

    Hokstada, Per; Oien, Knut; Reinertsen, Rune

    1998-01-01

    This paper provides guidance on the process of establishing input data to safety and reliability engineering analyses when no or little field data exist, and expert judgment is required. Some recommendations are directly related to a discussion of basic requirements for scientific work. Further, two case studies are discussed in order to highlight some actual problem areas that are experienced when using expert judgment, and some recommendations for handling these problems are given. The first case describes how expert judgment was used to analyse the safe operation of an umbilical on a semisubmersible drilling rig, and the second case is related to establishing generic failure rates/probabilities for components of offshore safety systems

  19. 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.

  20. Application of system reliability analytical method, GO-FLOW

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi; Fukuto, Junji; Mitomo, Nobuo; Miyazaki, Keiko; Matsukura, Hiroshi; Kobayashi, Michiyuki

    1999-01-01

    The Ship Research Institute proceed a developmental study on GO-FLOW method with various advancing functionalities for the system reliability analysis method occupying main parts of PSA (Probabilistic Safety Assessment). Here was attempted to intend to upgrade functionality of the GO-FLOW method, to develop an analytical function integrated with dynamic behavior analytical function, physical behavior and probable subject transfer, and to prepare a main accident sequence picking-out function. In 1997 fiscal year, in dynamic event-tree analytical system, an analytical function was developed by adding dependency between headings. In simulation analytical function of the accident sequence, main accident sequence of MRX for improved ship propulsion reactor became possible to be covered perfectly. And, input data for analysis was prepared with a function capable easily to set by an analysis operator. (G.K.)