WorldWideScience

Sample records for probabilistic safety verification

  1. Safety Verification for Probabilistic Hybrid Systems

    DEFF Research Database (Denmark)

    Zhang, Lijun; She, Zhikun; Ratschan, Stefan

    2010-01-01

    The interplay of random phenomena and continuous real-time control deserves increased attention for instance in wireless sensing and control applications. Safety verification for such systems thus needs to consider probabilistic variations of systems with hybrid dynamics. In safety verification o...... on a number of case studies, tackled using a prototypical implementation....

  2. Safety Verification for Probabilistic Hybrid Systems

    Czech Academy of Sciences Publication Activity Database

    Zhang, J.; She, Z.; Ratschan, Stefan; Hermanns, H.; Hahn, E.M.

    2012-01-01

    Roč. 18, č. 6 (2012), s. 572-587 ISSN 0947-3580 R&D Projects: GA MŠk OC10048; GA ČR GC201/08/J020 Institutional research plan: CEZ:AV0Z10300504 Keywords : model checking * hybrid system s * formal verification Subject RIV: IN - Informatics, Computer Science Impact factor: 1.250, year: 2012

  3. The design and verification of probabilistic safety analysis platform NFRisk

    International Nuclear Information System (INIS)

    Hu Wenjun; Song Wei; Ren Lixia; Qian Hongtao

    2010-01-01

    To increase the technical ability in Probabilistic Safety Analysis (PSA) field in China,it is necessary and important to study and develop indigenous professional PSA platform. Following such principle as 'from structure simplification to modulization to production of cut sets to minimum of cut sets', the algorithms, including simplification algorithm, modulization algorithm, the algorithm of conversion from fault tree to binary decision diagram (BDD), the solving algorithm of cut sets, the minimum algorithm of cut sets, and so on, were designed and developed independently; the design of data management and operation platform was completed all alone; the verification and validation of NFRisk platform based on 3 typical fault trees was finished on our own. (authors)

  4. Tensit - a novel probabilistic simulation tool for safety assessments. Tests and verifications using biosphere models

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

    2004-06-01

    This report documents the verification of a new simulation tool for dose assessment put together in a package under the name Tensit (Technical Nuclide Simulation Tool). The tool is developed to solve differential equation systems describing transport and decay of radionuclides. It is capable of handling both deterministic and probabilistic simulations. The verifications undertaken shows good results. Exceptions exist only where the reference results are unclear. Tensit utilise and connects two separate commercial softwares. The equation solving capability is derived from the Matlab/Simulink software environment to which Tensit adds a library of interconnectable building blocks. Probabilistic simulations are provided through a statistical software named at{sub R}isk that communicates with Matlab/Simulink. More information about these softwares can be found at www.palisade.com and www.mathworks.com. The underlying intention of developing this new tool has been to make available a cost efficient and easy to use means for advanced dose assessment simulations. The mentioned benefits are gained both through the graphical user interface provided by Simulink and at{sub R}isk, and the use of numerical equation solving routines in Matlab. To verify Tensit's numerical correctness, an implementation was done of the biosphere modules for dose assessments used in the earlier safety assessment project SR 97. Acquired probabilistic results for deterministic as well as probabilistic simulations have been compared with documented values. Additional verification has been made both with another simulation tool named AMBER and also against the international test case from PSACOIN named Level 1B. This report documents the models used for verification with equations and parameter values so that the results can be recreated. For a background and a more detailed description of the underlying processes in the models, the reader is referred to the original references. Finally, in the

  5. Tensit - a novel probabilistic simulation tool for safety assessments. Tests and verifications using biosphere models

    International Nuclear Information System (INIS)

    Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

    2004-06-01

    This report documents the verification of a new simulation tool for dose assessment put together in a package under the name Tensit (Technical Nuclide Simulation Tool). The tool is developed to solve differential equation systems describing transport and decay of radionuclides. It is capable of handling both deterministic and probabilistic simulations. The verifications undertaken shows good results. Exceptions exist only where the reference results are unclear. Tensit utilise and connects two separate commercial softwares. The equation solving capability is derived from the Matlab/Simulink software environment to which Tensit adds a library of interconnectable building blocks. Probabilistic simulations are provided through a statistical software named at R isk that communicates with Matlab/Simulink. More information about these softwares can be found at www.palisade.com and www.mathworks.com. The underlying intention of developing this new tool has been to make available a cost efficient and easy to use means for advanced dose assessment simulations. The mentioned benefits are gained both through the graphical user interface provided by Simulink and at R isk, and the use of numerical equation solving routines in Matlab. To verify Tensit's numerical correctness, an implementation was done of the biosphere modules for dose assessments used in the earlier safety assessment project SR 97. Acquired probabilistic results for deterministic as well as probabilistic simulations have been compared with documented values. Additional verification has been made both with another simulation tool named AMBER and also against the international test case from PSACOIN named Level 1B. This report documents the models used for verification with equations and parameter values so that the results can be recreated. For a background and a more detailed description of the underlying processes in the models, the reader is referred to the original references. Finally, in the perspective of

  6. Formalizing Probabilistic Safety Claims

    Science.gov (United States)

    Herencia-Zapana, Heber; Hagen, George E.; Narkawicz, Anthony J.

    2011-01-01

    A safety claim for a system is a statement that the system, which is subject to hazardous conditions, satisfies a given set of properties. Following work by John Rushby and Bev Littlewood, this paper presents a mathematical framework that can be used to state and formally prove probabilistic safety claims. It also enables hazardous conditions, their uncertainties, and their interactions to be integrated into the safety claim. This framework provides a formal description of the probabilistic composition of an arbitrary number of hazardous conditions and their effects on system behavior. An example is given of a probabilistic safety claim for a conflict detection algorithm for aircraft in a 2D airspace. The motivation for developing this mathematical framework is that it can be used in an automated theorem prover to formally verify safety claims.

  7. Probabilistic safety assessment in radioactive waste disposal

    International Nuclear Information System (INIS)

    Robinson, P.C.

    1987-07-01

    Probabilistic safety assessment codes are now widely used in radioactive waste disposal assessments. This report gives an overview of the current state of the field. The relationship between the codes and the regulations covering radioactive waste disposal is discussed and the characteristics of current codes is described. The problems of verification and validation are considered. (author)

  8. Validation and verification plan for safety and PRA codes

    International Nuclear Information System (INIS)

    Ades, M.J.; Crowe, R.D.; Toffer, H.

    1991-04-01

    This report discusses a verification and validation (V ampersand V) plan for computer codes used for safety analysis and probabilistic risk assessment calculations. The present plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River Office (DOE-SRO) to bring the essential safety analysis and probabilistic risk assessment codes in compliance with verification and validation requirements

  9. Probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hoertner, H.; Schuetz, B.

    1982-09-01

    For the purpose of assessing applicability and informativeness on risk-analysis methods in licencing procedures under atomic law, the choice of instruments for probabilistic analysis, the problems in and experience gained in their application, and the discussion of safety goals with respect to such instruments are of paramount significance. Naturally, such a complex field can only be dealt with step by step, making contribution relative to specific problems. The report on hand shows the essentials of a 'stocktaking' of systems relability studies in the licencing procedure under atomic law and of an American report (NUREG-0739) on 'Quantitative Safety Goals'. (orig.) [de

  10. Probabilistic Decision Graphs - Combining Verification and AI Techniques for Probabilistic Inference

    DEFF Research Database (Denmark)

    Jaeger, Manfred

    2004-01-01

    We adopt probabilistic decision graphs developed in the field of automated verification as a tool for probabilistic model representation and inference. We show that probabilistic inference has linear time complexity in the size of the probabilistic decision graph, that the smallest probabilistic ...

  11. Reload core safety verification

    International Nuclear Information System (INIS)

    Svetlik, M.; Minarcin, M.

    2003-01-01

    This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

  12. Probabilistic safety analysis procedures guide

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Buslik, A.J.

    1984-01-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of tissues affecting reactor safety. This guide addresses the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant and from loss of offsite electric power. The scope includes analyses of problem-solving (cognitive) human errors, a determination of importance of the various core damage accident sequences, and an explicit treatment and display of uncertainties for the key accident sequences. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance) and the risk associated with external accident initiators, as consensus is developed regarding suitable methodologies in these areas. This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are essential for regulatory decision making. Methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study

  13. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    verification' are used differently in different countries. The way that these terms have been used in this Safety Guide is explained in Section 2. The term 'design' as used here includes the specifications for the safe operation and management of the plant. This Safety Guide identifies the key recommendations for carrying out the safety assessment and the independent verification. It provides detailed guidance in support of IAEA, Safety of Nuclear Power Plants: Design, Safety Standards Series No. NS-R-1 (2000), particularly in the area of safety analysis. However, this does not include all the technical details which are available and reference is made to other IAEA publications on specific design issues and safety analysis methods. Specific deterministic or probabilistic safety targets or radiological limits can vary in different countries and are the responsibility of the regulatory body. This Safety Guide provides some references to targets and limits established by international organizations. Operators, and sometimes designers, may also set their own safety targets which may be more stringent than those set by the regulator or may address different aspects of safety. In some countries operators are expected to do this as part of their 'ownership' of the entire safety case. This Safety Guide does not include specific recommendations for the safety assessment of those plant systems for which dedicated Safety Guides exist. Section 2 defines the terms 'safety assessment', 'safety analysis' and 'independent verification' and outlines their relationship. Section 3 gives the key recommendations for the safety assessment of the principal and plant design requirements. Section 4 gives the key recommendations for safety analysis. It describes the identification of postulated initiating events (PIEs), which are used throughout the safety assessment including the safety analysis, the deterministic transient analysis and severe accident analysis, and the probabilistic safety analysis

  14. Living probabilistic safety assessment (LPSA)

    International Nuclear Information System (INIS)

    1999-08-01

    Over the past few years many nuclear power plant organizations have performed probabilistic safety assessments (PSAs) to identify and understand key plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. PSA is an effective tool for this purpose as it assists plant management to target resources where the largest benefit to plant safety can be obtained. However, any PSA which is to be used in this way must have a credible and defensible basis. Thus, it is very important to have a high quality 'living PSA' accepted by the plant and the regulator. With this background in mind, the IAEA has prepared this report on Living Probabilistic Safety Assessment (LPSA) which addresses the updating, documentation, quality assurance, and management and organizational requirements for LPSA. Deficiencies in the areas addressed in this report would seriously reduce the adequacy of the LPSA as a tool to support decision making at NPPs. This report was reviewed by a working group during a Technical Committee Meeting on PSA Applications to Improve NPP Safety held in Madrid, Spain, from 23 to 27 February 1998

  15. Aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jordan Cizelj, R.; Kozuh, M.

    1995-01-01

    Aging is a phenomenon, which is influencing on unavailability of all components of the plant. The influence of aging on Probabilistic Safety Assessment calculations was estimated for Electrical Power Supply System. The average increase of system unavailability due to aging of system components was estimated and components were prioritized regarding their influence on change of system unavailability and relative increase of their unavailability due to aging. After the analysis of some numerical results, the recommendation for a detailed research of aging phenomena and its influence on system availability is given. (author)

  16. Probabilistic safety analysis using microcomputer

    International Nuclear Information System (INIS)

    Futuro Filho, F.L.F.; Mendes, J.E.S.; Santos, M.J.P. dos

    1990-01-01

    The main steps of execution of a Probabilistic Safety Assessment (PSA) are presented in this report, as the study of the system description, construction of event trees and fault trees, and the calculation of overall unavailability of the systems. It is also presented the use of microcomputer in performing some tasks, highlightning the main characteristics of a software to perform adequately the job. A sample case of fault tree construction and calculation is presented, using the PSAPACK software, distributed by the IAEA (International Atomic Energy Agency) for training purpose. (author)

  17. Probabilistic Requirements (Partial) Verification Methods Best Practices Improvement. Variables Acceptance Sampling Calculators: Derivations and Verification of Plans. Volume 1

    Science.gov (United States)

    Johnson, Kenneth L.; White, K, Preston, Jr.

    2012-01-01

    The NASA Engineering and Safety Center was requested to improve on the Best Practices document produced for the NESC assessment, Verification of Probabilistic Requirements for the Constellation Program, by giving a recommended procedure for using acceptance sampling by variables techniques. This recommended procedure would be used as an alternative to the potentially resource-intensive acceptance sampling by attributes method given in the document. This document contains the outcome of the assessment.

  18. Prospects for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.

    1992-01-01

    This article provides some reflections on future developments of Probabilistic Safety Assessment (PSA) in view of the present state of the art and evaluates current trends in the use of PSA for safety management. The main emphasis is on Level 1 PSA, although Level 2 aspects are also highlighted to some extent. As a starting point, the role of PSA is outlined from a historical perspective, demonstrating the rapid expansion of the uses of PSA. In this context the wide spectrum of PSA applications and the associated benefits to the users are in focus. It should be kept in mind, however, that PSA, in spite of its merits, is not a self-standing safety tool. It complements deterministic analysis and thus improves understanding and facilitating prioritization of safety issues. Significant progress in handling PSA limitations - such as reliability data, common-cause failures, human interactions, external events, accident progression, containment performance, and source-term issues - is described. This forms a background for expected future developments of PSA. Among the most important issues on the agenda for the future are PSA scope extensions, methodological improvements and computer code advancements, and full exploitation of the potential benefits of applications to operational safety management. Many PSA uses, if properly exercised, lead to safety improvements as well as major burden reductions. The article provides, in addition, International Atomic Energy Agency (IAEA) perspective on the topics covered, as reflected in the current PSA programs of the agency. 74 refs., 6 figs., 1 tab

  19. Consideration of aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Titina, B.; Cepin, M.

    2007-01-01

    Probabilistic safety assessment is a standardised tool for assessment of safety of nuclear power plants. It is a complement to the safety analyses. Standard probabilistic models of safety equipment assume component failure rate as a constant. Ageing of systems, structures and components can theoretically be included in new age-dependent probabilistic safety assessment, which generally causes the failure rate to be a function of age. New age-dependent probabilistic safety assessment models, which offer explicit calculation of the ageing effects, are developed. Several groups of components are considered which require their unique models: e.g. operating components e.g. stand-by components. The developed models on the component level are inserted into the models of the probabilistic safety assessment in order that the ageing effects are evaluated for complete systems. The preliminary results show that the lack of necessary data for consideration of ageing causes highly uncertain models and consequently the results. (author)

  20. Probabilistic safety assessment - regulatory perspective

    International Nuclear Information System (INIS)

    Solanki, R.B.; Paul, U.K.; Hajra, P.; Agarwal, S.K.

    2002-01-01

    Full text: Nuclear power plants (NPPs) have been designed, constructed and operated mainly based on deterministic safety analysis philosophy. In this approach, a substantial amount of safety margin is incorporated in the design and operational requirements. Additional margin is incorporated by applying the highest quality engineering codes, standards and practices, and the concept of defence-in-depth in design and operating procedures, by including conservative assumptions and acceptance criteria in plant response analysis of postulated initiating events (PIEs). However, as the probabilistic approach has been improved and refined over the years, it is possible for the designer, operator and regulator to get a more detailed and realistic picture of the safety importance of plant design features, operating procedures and operational practices by using probabilistic safety assessment (PSA) along with the deterministic methodology. At present, many countries including USA, UK and France are using PSA insights in their decision making along with deterministic basis. India has also made substantial progress in the development of methods for carrying out PSA. However, consensus on the use of PSA in regulatory decision-making has not been achieved yet. This paper emphasises on the requirements (e.g.,level of details, key modelling assumptions, data, modelling aspects, success criteria, sensitivity and uncertainty analysis) for improving the quality and consistency in performance and use of PSA that can facilitate meaningful use of the PSA insights in the regulatory decision-making in India. This paper also provides relevant information on international scenario and various application areas of PSA along with progress made in India. The PSA perspective presented in this paper may help in achieving consensus on the use of PSA for regulatory / utility decision-making in design and operation of NPPs

  1. Probabilistic Programming : A True Verification Challenge

    NARCIS (Netherlands)

    Katoen, Joost P.; Finkbeiner, Bernd; Pu, Geguang; Zhang, Lijun

    2015-01-01

    Probabilistic programs [6] are sequential programs, written in languages like C, Java, Scala, or ML, with two added constructs: (1) the ability to draw values at random from probability distributions, and (2) the ability to condition values of variables in a program through observations. For a

  2. Probabilistic safety assessment for seismic events

    International Nuclear Information System (INIS)

    1993-10-01

    This Technical Document on Probabilistic Safety Assessment for Seismic Events is mainly associated with the Safety Practice on Treatment of External Hazards in PSA and discusses in detail one specific external hazard, i.e. earthquakes

  3. Probabilistic Requirements (Partial) Verification Methods Best Practices Improvement. Variables Acceptance Sampling Calculators: Empirical Testing. Volume 2

    Science.gov (United States)

    Johnson, Kenneth L.; White, K. Preston, Jr.

    2012-01-01

    The NASA Engineering and Safety Center was requested to improve on the Best Practices document produced for the NESC assessment, Verification of Probabilistic Requirements for the Constellation Program, by giving a recommended procedure for using acceptance sampling by variables techniques as an alternative to the potentially resource-intensive acceptance sampling by attributes method given in the document. In this paper, the results of empirical tests intended to assess the accuracy of acceptance sampling plan calculators implemented for six variable distributions are presented.

  4. Integrated Deterministic-Probabilistic Safety Assessment Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, P.; Vorobyev, Y.; Sanchez-Perea, M.; Queral, C.; Jimenez Varas, G.; Rebollo, M. J.; Mena, L.; Gomez-Magin, J.

    2014-02-01

    IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be based on the coupling of deterministic (consequences) and probabilistic (frequency) considerations to address the mutual interactions between stochastic disturbances (e.g. failures of the equipment, human actions, stochastic physical phenomena) and deterministic response of the plant (i.e. transients). This paper gives a general overview of some IDPSA methods as well as some possible applications to PWR safety analyses. (Author)

  5. The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

  6. Developing Probabilistic Safety Performance Margins for Unknown and Underappreciated Risks

    Science.gov (United States)

    Benjamin, Allan; Dezfuli, Homayoon; Everett, Chris

    2015-01-01

    Probabilistic safety requirements currently formulated or proposed for space systems, nuclear reactor systems, nuclear weapon systems, and other types of systems that have a low-probability potential for high-consequence accidents depend on showing that the probability of such accidents is below a specified safety threshold or goal. Verification of compliance depends heavily upon synthetic modeling techniques such as PRA. To determine whether or not a system meets its probabilistic requirements, it is necessary to consider whether there are significant risks that are not fully considered in the PRA either because they are not known at the time or because their importance is not fully understood. The ultimate objective is to establish a reasonable margin to account for the difference between known risks and actual risks in attempting to validate compliance with a probabilistic safety threshold or goal. In this paper, we examine data accumulated over the past 60 years from the space program, from nuclear reactor experience, from aircraft systems, and from human reliability experience to formulate guidelines for estimating probabilistic margins to account for risks that are initially unknown or underappreciated. The formulation includes a review of the safety literature to identify the principal causes of such risks.

  7. Verification of safety critical software

    International Nuclear Information System (INIS)

    Son, Ki Chang; Chun, Chong Son; Lee, Byeong Joo; Lee, Soon Sung; Lee, Byung Chai

    1996-01-01

    To assure quality of safety critical software, software should be developed in accordance with software development procedures and rigorous software verification and validation should be performed. Software verification is the formal act of reviewing, testing of checking, and documenting whether software components comply with the specified requirements for a particular stage of the development phase[1]. New software verification methodology was developed and was applied to the Shutdown System No. 1 and 2 (SDS1,2) for Wolsung 2,3 and 4 nuclear power plants by Korea Atomic Energy Research Institute(KAERI) and Atomic Energy of Canada Limited(AECL) in order to satisfy new regulation requirements of Atomic Energy Control Boars(AECB). Software verification methodology applied to SDS1 for Wolsung 2,3 and 4 project will be described in this paper. Some errors were found by this methodology during the software development for SDS1 and were corrected by software designer. Outputs from Wolsung 2,3 and 4 project have demonstrated that the use of this methodology results in a high quality, cost-effective product. 15 refs., 6 figs. (author)

  8. New Aspects of Probabilistic Forecast Verification Using Information Theory

    Science.gov (United States)

    Tödter, Julian; Ahrens, Bodo

    2013-04-01

    This work deals with information-theoretical methods in probabilistic forecast verification, particularly concerning ensemble forecasts. Recent findings concerning the "Ignorance Score" are shortly reviewed, then a consistent generalization to continuous forecasts is motivated. For ensemble-generated forecasts, the presented measures can be calculated exactly. The Brier Score (BS) and its generalizations to the multi-categorical Ranked Probability Score (RPS) and to the Continuous Ranked Probability Score (CRPS) are prominent verification measures for probabilistic forecasts. Particularly, their decompositions into measures quantifying the reliability, resolution and uncertainty of the forecasts are attractive. Information theory sets up a natural framework for forecast verification. Recently, it has been shown that the BS is a second-order approximation of the information-based Ignorance Score (IGN), which also contains easily interpretable components and can also be generalized to a ranked version (RIGN). Here, the IGN, its generalizations and decompositions are systematically discussed in analogy to the variants of the BS. Additionally, a Continuous Ranked IGN (CRIGN) is introduced in analogy to the CRPS. The useful properties of the conceptually appealing CRIGN are illustrated, together with an algorithm to evaluate its components reliability, resolution, and uncertainty for ensemble-generated forecasts. This algorithm can also be used to calculate the decomposition of the more traditional CRPS exactly. The applicability of the "new" measures is demonstrated in a small evaluation study of ensemble-based precipitation forecasts.

  9. Probabilistic assessment of nuclear safety and safeguards

    International Nuclear Information System (INIS)

    Higson, D.J.

    1987-01-01

    Nuclear reactor accidents and diversions of materials from the nuclear fuel cycle are perceived by many people as particularly serious threats to society. Probabilistic assessment is a rational approach to the evaluation of both threats, and may provide a basis for decisions on appropriate actions to control them. Probabilistic method have become standard tools used in the analysis of safety, but there are disagreements on the criteria to be applied when assessing the results of analysis. Probabilistic analysis and assessment of the effectiveness of nuclear material safeguards are still at an early stage of development. (author)

  10. Probabilistic studies for a safety assurance program

    International Nuclear Information System (INIS)

    Iyer, S.S.; Davis, J.F.

    1985-01-01

    The adequate supply of energy is always a matter of concern for any country. Nuclear power has played, and will continue to play an important role in supplying this energy. However, safety in nuclear power production is a fundamental prerequisite in fulfilling this role. This paper outlines a program to ensure safe operation of a nuclear power plant utilizing the Probabilistic Safety Studies

  11. Probabilistic safety assessment goals in Canada

    International Nuclear Information System (INIS)

    Snell, V.G.

    1986-01-01

    CANDU safety philosphy, both in design and in licensing, has always had a strong bias towards quantitative probabilistically-based goals derived from comparative safety. Formal probabilistic safety assessment began in Canada as a design tool. The influence of this carried over later on into the definition of the deterministic safety guidelines used in CANDU licensing. Design goals were further developed which extended the consequence/frequency spectrum of 'acceptable' events, from the two points defined by the deterministic single/dual failure analysis, to a line passing through lower and higher frequencies. Since these were design tools, a complete risk summation was not necessary, allowing a cutoff at low event frequencies while preserving the identification of the most significant safety-related events. These goals gave a logical framework for making decisions on implementing design changes proposed as a result of the Probabilistic Safety Analysis. Performing this analysis became a regulatory requirement, and the design goals remained the framework under which this was submitted. Recently, there have been initiatives to incorporate more detailed probabilistic safety goals into the regulatory process in Canada. These range from far-reaching safety optimization across society, to initiatives aimed at the nuclear industry only. The effectiveness of the latter is minor at very low and very high event frequencies; at medium frequencies, a justification against expenditures per life saved in other industries should be part of the goal setting

  12. Probabilistic safety goals. Phase 3 - Status report

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J.-E. (VTT (Finland)); Knochenhauer, M. (Relcon Scandpower AB, Sundbyberg (Sweden))

    2009-07-15

    The first phase of the project (2006) described the status, concepts and history of probabilistic safety goals for nuclear power plants. The second and third phases (2007-2008) have provided guidance related to the resolution of some of the problems identified, and resulted in a common understanding regarding the definition of safety goals. The basic aim of phase 3 (2009) has been to increase the scope and level of detail of the project, and to start preparations of a guidance document. Based on the conclusions from the previous project phases, the following issues have been covered: 1) Extension of international overview. Analysis of results from the questionnaire performed within the ongoing OECD/NEA WGRISK activity on probabilistic safety criteria, including participation in the preparation of the working report for OECD/NEA/WGRISK (to be finalised in phase 4). 2) Use of subsidiary criteria and relations between these (to be finalised in phase 4). 3) Numerical criteria when using probabilistic analyses in support of deterministic safety analysis (to be finalised in phase 4). 4) Guidance for the formulation, application and interpretation of probabilistic safety criteria (to be finalised in phase 4). (LN)

  13. Probabilistic safety goals. Phase 3 - Status report

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Knochenhauer, M.

    2009-07-01

    The first phase of the project (2006) described the status, concepts and history of probabilistic safety goals for nuclear power plants. The second and third phases (2007-2008) have provided guidance related to the resolution of some of the problems identified, and resulted in a common understanding regarding the definition of safety goals. The basic aim of phase 3 (2009) has been to increase the scope and level of detail of the project, and to start preparations of a guidance document. Based on the conclusions from the previous project phases, the following issues have been covered: 1) Extension of international overview. Analysis of results from the questionnaire performed within the ongoing OECD/NEA WGRISK activity on probabilistic safety criteria, including participation in the preparation of the working report for OECD/NEA/WGRISK (to be finalised in phase 4). 2) Use of subsidiary criteria and relations between these (to be finalised in phase 4). 3) Numerical criteria when using probabilistic analyses in support of deterministic safety analysis (to be finalised in phase 4). 4) Guidance for the formulation, application and interpretation of probabilistic safety criteria (to be finalised in phase 4). (LN)

  14. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

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

  15. Probabilistic optimization of safety coefficients

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  16. Probabilistic safety analysis and interpretation thereof

    International Nuclear Information System (INIS)

    Steininger, U.; Sacher, H.

    1999-01-01

    Increasing use of the instrumentation of PSA is being made in Germany for quantitative technical safety assessment, for example with regard to incidents which must be reported and forwarding of information, especially in the case of modification of nuclear plants. The Commission for Nuclear Reactor Safety recommends regular execution of PSA on a cycle period of ten years. According to the PSA guidance instructions, probabilistic analyses serve for assessing the degree of safety of the entire plant, expressed as the expectation value for the frequency of endangering conditions. The authors describe the method, action sequence and evaluation of the probabilistic safety analyses. The limits of probabilistic safety analyses arise in the practical implementation. Normally the guidance instructions for PSA are confined to the safety systems, so that in practice they are at best suitable for operational optimisation only to a limited extent. The present restriction of the analyses has a similar effect on power output operation of the plant. This seriously degrades the utilitarian value of these analyses for the plant operators. In order to further develop PSA as a supervisory and operational optimisation instrument, both authors consider it to be appropriate to bring together the specific know-how of analysts, manufacturers, plant operators and experts. (orig.) [de

  17. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

    Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

  18. A probabilistic bridge safety evaluation against floods.

    Science.gov (United States)

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

    2016-01-01

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

  19. Probabilistic safety assessment as a standpoint for decision making

    International Nuclear Information System (INIS)

    Cepin, M.

    2001-01-01

    This paper focuses on the role of probabilistic safety assessment in decision-making. The prerequisites for use of the results of probabilistic safety assessment and the criteria for the decision-making based on probabilistic safety assessment are discussed. The decision-making process is described. It provides a risk evaluation of impact of the issue under investigation. Selected examples are discussed, which highlight the described process. (authors)

  20. Probabilistic safety assessment for food irradiation facility

    International Nuclear Information System (INIS)

    Solanki, R.B.; Prasad, M.; Sonawane, A.U.; Gupta, S.K.

    2012-01-01

    Highlights: ► Different considerations are required in PSA for Non-Reactor Nuclear Facilities. ► We carried out PSA for food irradiation facility as a part of safety evaluation. ► The results indicate that the fatal exposure risk is below the ‘acceptable risk’. ► Adequate operator training and observing good safety culture would reduce the risk. - Abstract: Probabilistic safety assessment (PSA) is widely used for safety evaluation of Nuclear Power Plants (NPPs) worldwide. The approaches and methodologies are matured and general consensus exists on using these approaches in PSA applications. However, PSA applications for safety evaluation for non-reactor facilities are limited. Due to differences in the processes in nuclear reactor facilities and non-reactor facilities, the considerations are different in application of PSA to these facilities. The food irradiation facilities utilize gamma irradiation sources, X-ray machines and electron accelerators for the purpose of radiation processing of variety of food items. This is categorized as Non-Reactor Nuclear Facility. In this paper, the application of PSA to safety evaluation of food irradiation facility is presented considering the ‘fatality due to radiation overexposure’ as a risk measure. The results indicate that the frequency of the fatal exposure is below the numerical acceptance guidance for the risk to the individual. Further, it is found that the overall risk to the over exposure can be reduced by providing the adequate operator training and observing good safety culture.

  1. The LaSalle probabilistic safety analysis

    International Nuclear Information System (INIS)

    Frederick, L.G.; Massin, H.L.; Crane, G.R.

    1987-01-01

    A probabilistic safety analysis has been performed for LaSalle County Station, a twin-unit General Electric BWR5 Mark II nuclear power plant. A primary objective of this PSA is to provide engineers with a useful and useable tool for making design decisions, performing technical specification optimization, evaluating proposed regulatory changes to equipment and procedures, and as an aid in operator training. Other objectives are to identify the hypothetical accident sequences that would contribute to core damage frequency, and to provide assurance that the total expected frequency of core-damaging accidents is below 10 -4 per reactor-year in response to suggested goals. (orig./HSCH)

  2. Deterministic and probabilistic approach to safety analysis

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1980-01-01

    The examples discussed in this paper show that reliability analysis methods fairly well can be applied in order to interpret deterministic safety criteria in quantitative terms. For further improved extension of applied reliability analysis it has turned out that the influence of operational and control systems and of component protection devices should be considered with the aid of reliability analysis methods in detail. Of course, an extension of probabilistic analysis must be accompanied by further development of the methods and a broadening of the data base. (orig.)

  3. Probabilistic safety assessment activities at Ignalina NPP

    International Nuclear Information System (INIS)

    Bagdonas, A.

    1999-01-01

    The Barselina Project was initiated in the summer 1991. The project was a multilateral co-operation between Lithuania, Russia and Sweden up until phase 3, and phase 4 has been performed as a bilateral between Lithuania and Sweden. The long-range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. During phase 3, from 1993 to 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. During phase 4, from 1994 to 1996, the PSA was further developed, taking into account plant changes, improved modelling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The model reflected the plant status before the outage 1996. During phase 4+, 1998 to 1999 the PSA model was upgraded taking into account the newest plant modifications. The new PSA model of CPS/AZRT was developed. Modelling was based on the Single Failure Analysis

  4. Probabilistic safety assessment of the Fugen NPS

    International Nuclear Information System (INIS)

    Sotsu, Masutake; Iguchi, Yukihiro; Mizuno, Kouichi; Sato, Shinichirou; Shimizu, Miwako

    1999-01-01

    We performed a probabilistic safety assessment (PSA) on the Fugen NPS. The main topic of assessment was internal factors. We assessment core damage frequency (level 1 PSA) and containment damage frequency (level 2 PSA) during rated operation, and core damage frequency during shutdown (PSA during shutdowns). Our assessment showed that the core damage frequency of Fugen is well below the IAEA criteria for existing plants, that the conditional containment damage during shutdown is almost the target value of 0.1, and that the core damage frequency during shutdown is almost the same as that assessed during operation. These results confirm that the Fugen plant maintains a sufficient safety margin during shutdowns for regular inspections and for refueling. We developed and verified the effectiveness of an accident management plan incorporating the results of the assessment. (author)

  5. 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)

  6. PROBABILISTIC MODEL FOR AIRPORT RUNWAY SAFETY AREAS

    Directory of Open Access Journals (Sweden)

    Stanislav SZABO

    2017-06-01

    Full Text Available The Laboratory of Aviation Safety and Security at CTU in Prague has recently started a project aimed at runway protection zones. The probability of exceeding by a certain distance from the runway in common incident/accident scenarios (take-off/landing overrun/veer-off, landing undershoot is being identified relative to the runway for any airport. As a result, the size and position of safety areas around runways are defined for the chosen probability. The basis for probability calculation is a probabilistic model using statistics from more than 1400 real-world cases where jet airplanes have been involved over the last few decades. Other scientific studies have contributed to understanding the issue and supported the model’s application to different conditions.

  7. 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

  8. Bounding probabilistic safety assessment probabilities by reality

    International Nuclear Information System (INIS)

    Fragola, J.R.; Shooman, M.L.

    1991-01-01

    The investigation of the failure in systems where failure is a rare event makes the continual comparisons between the developed probabilities and empirical evidence difficult. The comparison of the predictions of rare event risk assessments with historical reality is essential to prevent probabilistic safety assessment (PSA) predictions from drifting into fantasy. One approach to performing such comparisons is to search out and assign probabilities to natural events which, while extremely rare, have a basis in the history of natural phenomena or human activities. For example the Segovian aqueduct and some of the Roman fortresses in Spain have existed for several millennia and in many cases show no physical signs of earthquake damage. This evidence could be used to bound the probability of earthquakes above a certain magnitude to less than 10 -3 per year. On the other hand, there is evidence that some repetitive actions can be performed with extremely low historical probabilities when operators are properly trained and motivated, and sufficient warning indicators are provided. The point is not that low probability estimates are impossible, but continual reassessment of the analysis assumptions, and a bounding of the analysis predictions by historical reality. This paper reviews the probabilistic predictions of PSA in this light, attempts to develop, in a general way, the limits which can be historically established and the consequent bounds that these limits place upon the predictions, and illustrates the methodology used in computing such limits. Further, the paper discusses the use of empirical evidence and the requirement for disciplined systematic approaches within the bounds of reality and the associated impact on PSA probabilistic estimates

  9. Accident simulator development for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Amendola, A.; Mancini, G.

    1985-01-01

    This paper describes the basic features of a new concept of incident simulator, Response System Analyzed (RSA) which is being developed within the CEC JRC Research Program on Reactor Safety. Focusing on somewhat different aims than actual simulators, RSA development extends the field of application of simulators to the area of risk and reliability analysis and in particular to the identification of relevant sequences, to the modeling of human behavior and to the validation of operating procedures. The fundamental components of the project, i.e. the deterministic transient model of the plant, the automatic probabilistic driver and the human possible intervention modeling, are discussed in connection with the problem of their dynamic interaction. The analyses so far performed by separately testing RSA on significant study cases have shown encouraging results and have proven the feasibility of the overall program

  10. Verification of reactor safety codes

    International Nuclear Information System (INIS)

    Murley, T.E.

    1978-01-01

    The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

  11. Probabilistic safety analysis : a new nuclear power plants licensing method

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de.

    1982-04-01

    After a brief retrospect of the application of Probabilistic Safety Analysis in the nuclear field, the basic differences between the deterministic licensing method, currently in use, and the probabilistic method are explained. Next, the two main proposals (by the AIF and the ACRS) concerning the establishment of the so-called quantitative safety goals (or simply 'safety goals') are separately presented and afterwards compared in their most fundamental aspects. Finally, some recent applications and future possibilities are discussed. (Author) [pt

  12. Probabilistic safety assessment in nuclear power plant management

    International Nuclear Information System (INIS)

    Holloway, N.J.

    1989-06-01

    Probabilistic Safety Assessment (PSA) techniques have been widely used over the past few years to assist in understanding how engineered systems respond to abnormal conditions, particularly during a severe accident. The use of PSAs in the design and operation of such systems thus contributes to the safety of nuclear power plants. Probabilistic safety assessments can be maintained to provide a continuous up-to-date assessment (Living PSA), supporting the management of plant operations and modifications

  13. SCALE criticality safety verification and validation package

    International Nuclear Information System (INIS)

    Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

    1998-01-01

    Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

  14. Probabilistic safety analysis applied to RBMK reactors

    International Nuclear Information System (INIS)

    Gerez Martin, L.; Fernandez Ramos, P.

    1995-01-01

    The project financed by the European Union ''Revision of RBMK Reactor Safety was divided into nine Topic Groups dealing with different aspects of safety. The area covered by Topic Group 9 was Probabilistic Safety Analysis. TG9 will have touched on some of the problems discussed by other groups, although in terms of the systematic quantification of the impact of design characteristics and RBMK reactor operating practices on the risk of core damage. On account of the reduced time scale and the resources available for the project, the analysis was made using a simplified method based on the results of PSAs conducted in Western countries and on the judgement of the group members. The simplifies method is based on the concepts of Qualification, Redundancy and Automatic Actuation of the systems considered. PSA experience shows that systems complying with the above-mentioned concepts have a failure probability of 1.0E-3 when redundancy is simple, ie two similar equipment items capable of carrying out the same function. In general terms, this value can be considered to be dominated by potential common cause failures. The value considered above changes according to factors that have a positive effect upon it, such as an additional redundancy with a different equipment item (eg a turbo pumps and a motor pump), individual trains with good separations, etc, or a negative effect, such as the absence of suitable periodical tests, the need for operators to perform manual operations, etc. Similarly, possible actions required by the operator during accident sequences are assigned failure probability values between 1 and 1.0E-4, according to the complexity of the action (including local actions to be performed outside the control room) and the time available

  15. The role of probabilistic safety assessment in the design

    International Nuclear Information System (INIS)

    Green, A.; Ingham, E.L.

    1989-01-01

    The use of probabilistic safety assessment (PSA) for Heysham 2 and Torness marked a major change in the design approach to nuclear safety within the U.K. Design Safety Guidelines incorporating probabilistic safety targets required that design justification would necessitate explicit consideration of the consequence of accidents in relation to their frequency. The paper discusses these safety targets and their implications, the integration of PSA into the design process and an outline of the methodology. The influence of PSA on the design is discussed together with its role in the overall demonstration of reactor safety. (author)

  16. Guidance for the definition and application of probabilistic safety criteria

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Knochenhauer, M.

    2011-05-01

    The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

  17. Guidance for the definition and application of probabilistic safety criteria

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J.-E. (VTT Technical Research Centre of Finland (Finland)); Knochenhauer, M. (Scandpower AB (Sweden))

    2011-05-15

    The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

  18. A dynamic probabilistic safety margin characterization approach in support of Integrated Deterministic and Probabilistic Safety Analysis

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Rai, Ajit; Zio, Enrico

    2016-01-01

    The challenge of Risk-Informed Safety Margin Characterization (RISMC) is to develop a methodology for estimating system safety margins in the presence of stochastic and epistemic uncertainties affecting the system dynamic behavior. This is useful to support decision-making for licensing purposes. In the present work, safety margin uncertainties are handled by Order Statistics (OS) (with both Bracketing and Coverage approaches) to jointly estimate percentiles of the distributions of the safety parameter and of the time required for it to reach these percentiles values during its dynamic evolution. The novelty of the proposed approach consists in the integration of dynamic aspects (i.e., timing of events) into the definition of a dynamic safety margin for a probabilistic Quantification of Margin and Uncertainties (QMU). The system here considered for demonstration purposes is the Lead–Bismuth Eutectic- eXperimental Accelerator Driven System (LBE-XADS). - Highlights: • We integrate dynamic aspects into the definition of a safety margins. • We consider stochastic and epistemic uncertainties affecting the system dynamics. • Uncertainties are handled by Order Statistics (OS). • We estimate the system grace time during accidental scenarios. • We apply the approach to an LBE-XADS accidental scenario.

  19. Automatic Probabilistic Program Verification through Random Variable Abstraction

    Directory of Open Access Journals (Sweden)

    Damián Barsotti

    2010-06-01

    Full Text Available The weakest pre-expectation calculus has been proved to be a mature theory to analyze quantitative properties of probabilistic and nondeterministic programs. We present an automatic method for proving quantitative linear properties on any denumerable state space using iterative backwards fixed point calculation in the general framework of abstract interpretation. In order to accomplish this task we present the technique of random variable abstraction (RVA and we also postulate a sufficient condition to achieve exact fixed point computation in the abstract domain. The feasibility of our approach is shown with two examples, one obtaining the expected running time of a probabilistic program, and the other the expected gain of a gambling strategy. Our method works on general guarded probabilistic and nondeterministic transition systems instead of plain pGCL programs, allowing us to easily model a wide range of systems including distributed ones and unstructured programs. We present the operational and weakest precondition semantics for this programs and prove its equivalence.

  20. Technical safety requirements control level verification

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  1. Technical safety requirements control level verification; TOPICAL

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  2. Integrated program of using of Probabilistic Safety Analysis in Spain

    International Nuclear Information System (INIS)

    1998-01-01

    Since 25 June 1986, when the CSN (Nuclear Safety Conseil) approve the Integrated Program of Probabilistic Safety Analysis, this program has articulated the main activities of CSN. This document summarize the activities developed during these years and reviews the Integrated programme

  3. Analysis of truncation limit in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Cepin, Marko

    2005-01-01

    A truncation limit defines the boundaries of what is considered in the probabilistic safety assessment and what is neglected. The truncation limit that is the focus here is the truncation limit on the size of the minimal cut set contribution at which to cut off. A new method was developed, which defines truncation limit in probabilistic safety assessment. The method specifies truncation limits with more stringency than presenting existing documents dealing with truncation criteria in probabilistic safety assessment do. The results of this paper indicate that the truncation limits for more complex probabilistic safety assessments, which consist of larger number of basic events, should be more severe than presently recommended in existing documents if more accuracy is desired. The truncation limits defined by the new method reduce the relative errors of importance measures and produce more accurate results for probabilistic safety assessment applications. The reduced relative errors of importance measures can prevent situations, where the acceptability of change of equipment under investigation according to RG 1.174 would be shifted from region, where changes can be accepted, to region, where changes cannot be accepted, if the results would be calculated with smaller truncation limit

  4. Probabilistic studies for safety at optimum cost

    International Nuclear Information System (INIS)

    Pitner, P.

    1999-01-01

    By definition, the risk of failure of very reliable components is difficult to evaluate. How can the best strategies for in service inspection and maintenance be defined to limit this risk to an acceptable level at optimum cost? It is not sufficient to design structures with margins, it is also essential to understand how they age. The probabilistic approach has made it possible to develop well proven concepts. (author)

  5. A utility theoretic view on probabilistic safety criteria

    International Nuclear Information System (INIS)

    Holmberg, J.E.

    1997-03-01

    A probabilistic safety criterion specifies the maximum acceptable hazard rates of various accidental consequences. Assuming that the criterion depends also on the benefit of the process to society and on the licensing time applied, we can regard such statements as preference relations. In this paper, a probabilistic safety criterion is interpreted to mean that if the accident hazard rate is higher than the accident hazard rate criterion, then the optimal stopping time of a hazardous process is shorter than the licensing time. This interpretation yields a condition for a feasible utility function. In particular, we derive such a condition for the parameters of a linear plus exponential utility function. (orig.) (12 refs.)

  6. Probabilistic safety analysis vs probabilistic fracture mechanics -relation and necessary merging

    International Nuclear Information System (INIS)

    Nilsson, Fred

    1997-01-01

    A comparison is made between some general features of probabilistic fracture mechanics (PFM) and probabilistic safety assessment (PSA) in its standard form. We conclude that: Result from PSA is a numerically expressed level of confidence in the system based on the state of current knowledge. It is thus not any objective measure of risk. It is important to carefully define the precise nature of the probabilistic statement and relate it to a well defined situation. Standardisation of PFM methods is necessary. PFM seems to be the only way to obtain estimates of the pipe break probability. Service statistics are of doubtful value because of scarcity of data and statistical inhomogeneity. Collection of service data should be directed towards the occurrence of growing cracks

  7. Representation of human behaviour in probabilistic safety analysis

    International Nuclear Information System (INIS)

    Whittingham, R.B.

    1991-01-01

    This paper provides an overview of the representation of human behaviour in probabilistic safety assessment. Human performance problems which may result in errors leading to accidents are considered in terms of methods of identification using task analysis, screening analysis of critical errors, representation and quantification of human errors in fault trees and event trees and error reduction measures. (author) figs., tabs., 43 refs

  8. Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

    International Nuclear Information System (INIS)

    Ergun, E; Kadiroglu, O.S.

    1999-01-01

    The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

  9. Risk measures in living probabilistic safety assessment

    International Nuclear Information System (INIS)

    Holmberg, J.; Niemelae, I.

    1993-05-01

    The main objectives of the study are: to define risk measures and suggested uses of them in various living PSA applications for the operational safety management and to describe specific model features required for living PSA applications. The report is based on three case studies performed within the Nordic research project Safety Evaluation by Use of Living PSA and Safety Indicators. (48 refs., 11 figs., 17 tabs.)

  10. Results of the CANDU 3 probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    1995-01-01

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10 -6 /year. 4 refs., 1 fig

  11. To dimension safety valves. Probabilist study

    International Nuclear Information System (INIS)

    Noel, Robert; Couvreur, Denis

    1982-01-01

    The gauge of safety valves of a steam pressure apparatus is usually determined according to an operating situation envelope which it is admitted covers all that can happen in reality. For the safety of the dryer-superheaters of turbines in nuclear power stations, Electricite de France and Alsthom-Atlantique made a reliability study; its method is exposed and the results are discussed. Such a study is heavy going and complex, but in return it permits a better quantitative understanding of the various dimension and operating parameters of an installation which condition its safety. It is therefore a source of progress [fr

  12. The dialectical thinking about deterministic and probabilistic safety analysis

    International Nuclear Information System (INIS)

    Qian Yongbai; Tong Jiejuan; Zhang Zuoyi; He Xuhong

    2005-01-01

    There are two methods in designing and analysing the safety performance of a nuclear power plant, the traditional deterministic method and the probabilistic method. To date, the design of nuclear power plant is based on the deterministic method. It has been proved in practice that the deterministic method is effective on current nuclear power plant. However, the probabilistic method (Probabilistic Safety Assessment - PSA) considers a much wider range of faults, takes an integrated look at the plant as a whole, and uses realistic criteria for the performance of the systems and constructions of the plant. PSA can be seen, in principle, to provide a broader and realistic perspective on safety issues than the deterministic approaches. In this paper, the historical origins and development trend of above two methods are reviewed and summarized in brief. Based on the discussion of two application cases - one is the changes to specific design provisions of the general design criteria (GDC) and the other is the risk-informed categorization of structure, system and component, it can be concluded that the deterministic method and probabilistic method are dialectical and unified, and that they are being merged into each other gradually, and being used in coordination. (authors)

  13. Applications of probabilistic risk analysis in nuclear criticality safety design

    International Nuclear Information System (INIS)

    Chang, J.K.

    1992-01-01

    Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

  14. CANDU 6 probabilistic safety study summary

    International Nuclear Information System (INIS)

    1988-07-01

    This report summarizes the methodology, phenomenology and results relevent to the assessment of severe events in a CANDU 6 (formerly designated CANDU 600) station. The station design being analysed is based on a CANDU 6 Mark I currently operating in Canada. This evaluation includes event frequency and fission product release assessments but does not include assessment of radiation dose to the public, so that the information is equivalent to a level 2 Probabilistic Risk Assessment (PRA). The study has shown that the predicted overall average frequency for core melt in a CANDU 6 Mark I is 4.4 x 10 -6 events/year. This low frequency is, in large part due to the heavy water moderator which acts as a heat sink, prevents UO 2 melting and maintains core geometry for many events which could otherwise result in a core melt. The consequences for most core melts will be limited to the release of a fraction of noble gases and organic iodides. Other isotopes will be condensed or dissolved in the containment atmosphere and are ultimately retained in the pool of water in the basement where they are unavailable for release. Most core melts (∼ 90%) can be mitigated by operator action so that there is no danger of consequential damage to the containment structure and leak tightness. The frequency and consequences of less likely, more severe core melt sequences are also discussed in this report and shown to be small contributors to public risk

  15. Savannah River Site K-Reactor Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  16. Contribution of operating feedback to probabilistic safety studies

    International Nuclear Information System (INIS)

    Guio, J.M. de; Lannoy, A.

    1992-03-01

    This paper presents the method used for PWR unit operation feedback analysis and its contribution to probabilistic safety studies. The targets were as follows: - use of failure data banks to assess reliability parameters, - use of event data banks to identify and quantify main system initiating events, - determination of a standard operating profile. These studies, performed in the context of nuclear power plant safety programs, prove useful not only to safety engineers but also to equipment experts, designers, operators and maintenance specialists. They constitute basic data for studies in all these areas or the departure point for new investigations. (authors). 3 figs., 3 tabs., 3 refs

  17. Probabilistic safety goals. Phase 2 - Status report

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Bjoerkman, K.; Rossi, J.; Knochenhauer, M.; Xuhong He; Persson, A.; Gustavsson, H.

    2008-07-01

    The second phase of the project, the outcome of which is described in this project report has mainly dealt with four issues: 1) Consistency in the usage of safety goals 2) Criteria for assessment of results from PSA level 2 3) Overview of international safety goals and experiences from their use 4) Safety goals related to other man-made risks in society. Consistency in judgement over time has been perceived to be one of the main problems in the usage of safety goals. Safety goals defined in the 80ies were met in the beginning with PSA:s performed to the standards of that time, i.e., by PSA:s that were quite limited in scope and level of detail compared to today's state of the art. This issue was investigated by performing a comparative review was performed of three generations of the same PSA, focusing on the impact from changes over time in component failure data, IE frequency, and modelling of the plant, including plant changes and changes in success criteria. It proved to be very time-consuming and in some cases next to impossible to correctly identify the basic causes for changes in PSA results. A multitude of different sub-causes turned out to combined and difficult to differentiate. Thus, rigorous book-keeping is needed in order to keep track of how and why PSA results change. This is especially important in order to differentiate 'real' differences due to plant changes and updated component and IE data from differences that are due to general PSA development (scope, level of detail, modelling issues). (au)

  18. Probabilistic safety goals. Phase 2 - Status report

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J.-E.; Bjoerkman, K. Rossi, J. (VTT (Finland)); Knochenhauer, M.; Xuhong He; Persson, A.; Gustavsson, H. (Relcon Scandpower AB, Sundbyberg (Sweden))

    2008-07-15

    The second phase of the project, the outcome of which is described in this project report has mainly dealt with four issues: 1) Consistency in the usage of safety goals 2) Criteria for assessment of results from PSA level 2 3) Overview of international safety goals and experiences from their use 4) Safety goals related to other man-made risks in society. Consistency in judgement over time has been perceived to be one of the main problems in the usage of safety goals. Safety goals defined in the 80ies were met in the beginning with PSA:s performed to the standards of that time, i.e., by PSA:s that were quite limited in scope and level of detail compared to today's state of the art. This issue was investigated by performing a comparative review was performed of three generations of the same PSA, focusing on the impact from changes over time in component failure data, IE frequency, and modelling of the plant, including plant changes and changes in success criteria. It proved to be very time-consuming and in some cases next to impossible to correctly identify the basic causes for changes in PSA results. A multitude of different sub-causes turned out to combined and difficult to differentiate. Thus, rigorous book-keeping is needed in order to keep track of how and why PSA results change. This is especially important in order to differentiate 'real' differences due to plant changes and updated component and IE data from differences that are due to general PSA development (scope, level of detail, modelling issues). (au)

  19. Probabilistic Elastic Part Model: A Pose-Invariant Representation for Real-World Face Verification.

    Science.gov (United States)

    Li, Haoxiang; Hua, Gang

    2018-04-01

    Pose variation remains to be a major challenge for real-world face recognition. We approach this problem through a probabilistic elastic part model. We extract local descriptors (e.g., LBP or SIFT) from densely sampled multi-scale image patches. By augmenting each descriptor with its location, a Gaussian mixture model (GMM) is trained to capture the spatial-appearance distribution of the face parts of all face images in the training corpus, namely the probabilistic elastic part (PEP) model. Each mixture component of the GMM is confined to be a spherical Gaussian to balance the influence of the appearance and the location terms, which naturally defines a part. Given one or multiple face images of the same subject, the PEP-model builds its PEP representation by sequentially concatenating descriptors identified by each Gaussian component in a maximum likelihood sense. We further propose a joint Bayesian adaptation algorithm to adapt the universally trained GMM to better model the pose variations between the target pair of faces/face tracks, which consistently improves face verification accuracy. Our experiments show that we achieve state-of-the-art face verification accuracy with the proposed representations on the Labeled Face in the Wild (LFW) dataset, the YouTube video face database, and the CMU MultiPIE dataset.

  20. A probabilistic safety assessment PEER review: Case study on the use of probabilistic safety assessment for safety decisions

    International Nuclear Information System (INIS)

    1989-10-01

    The purpose of this case study is to illustrate, using an actual example, the organizing and carrying out of an independent peer review of a draft full-scope (level 3) probabilistic safety assessment. The specific findings of the peer review are of less importance than the approach taken, the interaction between sponsor and study team, and the technical and administrative issues that can arise during a peer review. This case study will examine the following issues: how the scope of the peer review was established, based on how it was to be used by the review sponsoring body; how the level of effort was determined, and what this determination meant for the technical quality of the review; how the team of peer reviewers was selected; how the review itself was carried out; what findings were made; what was done with these findings by both the review sponsoring body and the PSA analysis team. 9 refs, 2 figs, 1 tab

  1. Probabilistic safety analysis and radiological protection

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.; Goes, A.G.A.

    1990-05-01

    The author presents a brief description of NUREG-1150 and NUREG-0956, both documents of great importance in the risk area. Based on document's recommendations and following NUREG-1150 similar methodology, a calculation model is proposed in this publication, with the purpose of analyzing the consequences of a severe accident in Angra-I Power Station. The suggested model can be divided in two stages: the first one called front-end considers the power station system safety during the accident, and the second called back-end cares for accident consequences. 9 refs. (B.C.A.)

  2. Probabilistic safety assessment of the nuclear facilities in Cuba

    International Nuclear Information System (INIS)

    Rivero O, J.J.; Salomon L, J.

    1991-01-01

    During 1986-1990 basis were established for further developing probabilistic safety assessment (PSA) of Juragua NPP. A team work was consolidated and carried out the preliminary studies of the small break LOCA initiating event. A significant achievement was the creation of the ANCON code, which allows the evaluation of complex fault trees in personal computers, and has been applied in PSA modelling, and specialist qualification. The paper describes the main results and future activities in this field. (author)

  3. Use of probabilistic safety analyses in severe accident management

    International Nuclear Information System (INIS)

    Neogy, P.; Lehner, J.

    1991-01-01

    An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs

  4. Results of the CANDU 3 probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaitly, R K [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10{sup -6}/year. 4 refs., 1 fig.

  5. Probabilistic assessment of NPP safety under aircraft impact

    International Nuclear Information System (INIS)

    Birbraer, A.N.; Roleder, A.J.; Arhipov, S.B.

    1999-01-01

    Methodology of probabilistic assessment of NPP safety under aircraft impact is described below. The assessment is made taking into account not only the fact of aircraft fall onto the NPP building, but another casual parameters too, namely an aircraft class, velocity and mass, as well as point and angle of its impact with the building structure. This analysis can permit to justify the decrease of the required structure strength and dynamic loads on the NPP equipment. It can also be especially useful when assessing the safety of existing NPP. (author)

  6. An overview-probabilistic safety analysis for research reactors

    International Nuclear Information System (INIS)

    Liu Jinlin; Peng Changhong

    2015-01-01

    For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

  7. Influence of probabilistic safety analysis on design and operation of PWR plants

    International Nuclear Information System (INIS)

    Bastl, W.; Hoertner, H.; Kafka, P.

    1978-01-01

    This paper gives a comprehensive presentation of the connections and influences of probabilistic safety analysis on design and operation of PWR plants. In this context a short historical retrospective view concerning probabilistic reliability analysis is given. In the main part of this paper some examples are presented in detail, showing special outcomes of such probabilistic investigations. Additional paragraphs illustrate some activities and issues in the field of probabilistic safety analysis

  8. Angra-1 probabilistic safety study-phase B

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.; Gibelli, S.M.O.

    1988-05-01

    This study represents the Phase B of the Angra-1 Probabilistic Safety Study and is the the final report prepared for the IAEA under Research Contract No. 3423/R2/RB. The three main items covered in this report are the establishment of interim safety goals, analysis of Angra-1 operational experience and development of emergency procedures to address severe accidents. For establishment of interim safety goals a methodology for calculating consequences and risks associated to the Angra-1 operation was developed based on the available data and codes. The proposed safety goals refer to the individual risk of early fatality for people living in the vicinity of the plant, colective risk of cancer fatalities for people living near the plant, the propobability of core melt occurrence and the probability of dominant accident sequences. (author) [pt

  9. Probabilistic safety criteria on high burnup HWR fuels

    International Nuclear Information System (INIS)

    Marino, A.C.

    2002-01-01

    BACO is a code for the simulation of the thermo-mechanical and fission gas behaviour of a cylindrical fuel rod under operation conditions. Their input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA-CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. The techniques for sensitivity analysis defined in BACO are: the 'extreme case analysis', the 'parametric analysis' and the 'probabilistic (or statistics) analysis'. We analyse the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs taking account probabilistic safety criteria and using the BACO's sensitivity analysis. (author)

  10. Computer codes for level 1 probabilistic safety assessment

    International Nuclear Information System (INIS)

    1990-06-01

    Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs

  11. 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.)

  12. 77 FR 26822 - Pipeline Safety: Verification of Records

    Science.gov (United States)

    2012-05-07

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No. PHMSA-2012-0068] Pipeline Safety: Verification of Records AGENCY: Pipeline and Hazardous Materials... issuing an Advisory Bulletin to remind operators of gas and hazardous liquid pipeline facilities to verify...

  13. Probabilistic safety analysis second level of WWER-TOI

    International Nuclear Information System (INIS)

    Chekin, A.A.; Bajkova, E.V.; Levin, V.N.; Shishina, E.S.

    2015-01-01

    Probabilistic safety assessment (PSA) of Level-1 and Level-2 gives a comprehensive qualitative and quantitative evaluation of the safety of the project. The operation of the unit at rated power is considered. As sources of radioactivity in the development of the second-level PSA, nuclear fuel in the core of the reactor is considered. As initiating events, internal initiating events (including de-energizing) are considered, which may arise due to failures of NPP systems, equipment or components, or due to erroneous actions of personnel. In general, an assessment of the level of project safety shows that the WWER-TOI project complies with the requirements of the TOR, as well as all the requirements of modern Russian and foreign regulatory documents in the field of security [ru

  14. Prioritization of R and D programs on probabilistic reactor safety

    International Nuclear Information System (INIS)

    Husseiny, A.A.

    1982-01-01

    An interactive computer code based on the multiattribute utility theory has been developed with graphic capabilities to use in selection of probabilistic reactor safety RandD programs. Utility values and proper graphic representation are made through lottery games on the computer terminal. The code is applied to prioritize a set of RandD programs on LWR safety based on attributes including regulatory issues, institutional issues and operation problems. The methodology is described here in detail with its applications. Some of the input includes statistical distributions and subjective judgments on institutional issues. The flexibility of the approach provides a tool for decision makers whether on individual or group level to assess LWR safety priorities and continuously update their strategies

  15. Probabilist methods applied to electric source problems in nuclear safety

    International Nuclear Information System (INIS)

    Carnino, A.; Llory, M.

    1979-01-01

    Nuclear Safety has frequently been asked to quantify safety margins and evaluate the hazard. In order to do so, the probabilist methods have proved to be the most promising. Without completely replacing determinist safety, they are now commonly used at the reliability or availability stages of systems as well as for determining the likely accidental sequences. In this paper an application linked to the problem of electric sources is described, whilst at the same time indicating the methods used. This is the calculation of the probable loss of all the electric sources of a pressurized water nuclear power station, the evaluation of the reliability of diesels by event trees of failures and the determination of accidental sequences which could be brought about by the 'total electric source loss' initiator and affect the installation or the environment [fr

  16. Uncertainty estimation in nuclear power plant probabilistic safety assessment

    International Nuclear Information System (INIS)

    Guarro, S.B.; Cummings, G.E.

    1989-01-01

    Probabilistic Risk Assessment (PRA) was introduced in the nuclear industry and the nuclear regulatory process in 1975 with the publication of the Reactor Safety Study by the U.S. Nuclear Regulatory Commission. Almost fifteen years later, the state-of-the-art in this field has been expanded and sharpened in many areas, and about thirty-five plant-specific PRAs (Probabilistic Risk Assessments) have been performed by the nuclear utility companies or by the U.S. Nuclear Regulatory commission. Among the areas where the most evident progress has been made in PRA and PSA (Probabilistic Safety Assessment, as these studies are more commonly referred to in the international community outside the U.S.) is the development of a consistent framework for the identification of sources of uncertainty and the estimation of their magnitude as it impacts various risk measures. Techniques to propagate uncertainty in reliability data through the risk models and display its effect on the top level risk estimates were developed in the early PRAs. The Seismic Safety Margin Research Program (SSMRP) study was the first major risk study to develop an approach to deal explicitly with uncertainty in risk estimates introduced not only by uncertainty in component reliability data, but by the incomplete state of knowledge of the assessor(s) with regard to basic phenomena that may trigger and drive a severe accident. More recently NUREG-1150, another major study of reactor risk sponsored by the NRC, has expanded risk uncertainty estimation and analysis into the realm of model uncertainty related to the relatively poorly known post-core-melt phenomena which determine the behavior of the molten core and of the rector containment structures

  17. Probabilistic safety assessment for Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Stack, D.S.; Kindinger, J.P.; Deremer, R.K.

    1995-01-01

    This paper gives results from the first comprehensive level-3 probabilistic safety assessment (PSA), including consideration of external events, for the Hanford tank farm (HTF). This work was sponsored by the U.S. Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM). At the HTF, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/saltcake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is ∼60 million gal, containing ∼200 million Ci of radioactivity

  18. Dependencies, human interactions and uncertainties in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.

    1990-01-01

    In the context of Probabilistic Safety Assessment (PSA), three areas were investigated in a 4-year Nordic programme: dependencies with special emphasis on common cause failures, human interactions and uncertainty aspects. The approach was centered around comparative analyses in form of Benchmark/Reference Studies and retrospective reviews. Weak points in available PSAs were identified and recommendations were made aiming at improving consistency of the PSAs. The sensitivity of PSA-results to basic assumptions was demonstrated and the sensitivity to data assignment and to choices of methods for analysis of selected topics was investigated. (author)

  19. Defining initiating events for purposes of probabilistic safety assessment

    International Nuclear Information System (INIS)

    1993-09-01

    This document is primarily directed towards technical staff involved in the performance or review of plant specific Probabilistic Safety Assessment (PSA). It highlights different approaches and provides typical examples useful for defining the Initiating Events (IE). The document also includes the generic initiating event database, containing about 300 records taken from about 30 plant specific PSAs. In addition to its usefulness during the actual performance of a PSA, the generic IE database is of the utmost importance for peer reviews of PSAs, such as the IAEA's International Peer Review Service (IPERS) where reference to studies on similar NPPs is needed. 60 refs, figs and tabs

  20. Probabilistic safety criteria at the safety function/system level

    International Nuclear Information System (INIS)

    1989-09-01

    A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  1. Selection and verification of safety parameters in safety parameter display system for nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yuangfang

    1992-02-01

    The method and results for safety parameter selection and its verification in safety parameter display system of nuclear power plants are introduced. According to safety analysis, the overall safety is divided into six critical safety functions, and a certain amount of safety parameters which can represent the integrity degree of each function and the causes of change are strictly selected. The verification of safety parameter selection is carried out from the view of applying the plant emergency procedures and in the accident man oeuvres on a full scale nuclear power plant simulator

  2. Comment on 'The meaning of probability in probabilistic safety analysis'

    International Nuclear Information System (INIS)

    Yellman, Ted W.; Murray, Thomas M.

    1995-01-01

    A recent article in Reliability Engineering and System Safety argues that there is 'fundamental confusion over how to interpret the numbers which emerge from a Probabilistic Safety Analysis [PSA]', [Watson, S. R., The meaning of probability in probabilistic safety analysis. Reliab. Engng and System Safety, 45 (1994) 261-269.] As a standard for comparison, the author employs the 'realist' interpretation that a PSA output probability should be a 'physical property' of the installation being analyzed, 'objectively measurable' without controversy. The author finds all the other theories and philosophies discussed wanting by this standard. Ultimately, he argues that the outputs of a PSA should be considered to be no more than constructs of the computational procedure chosen - just an 'argument' or a 'framework for the debate about safety' rather than a 'representation of truth'. He even suggests that 'competing' PSA's be done - each trying to 'argue' for a different message. The commentors suggest that the position the author arrives at is an overreaction to the subjectivity which is part of any complex PSA, and that that overreaction could in fact easily lead to the belief that PSA's are meaningless. They suggest a broader interpretation, one based strictly on relative frequency--a concept which the commentors believe the author abandoned too quickly. Their interpretation does not require any 'tests' to determine whether a statement of likelihood is qualified to be a 'true' probability and it applies equally well in pure analytical models. It allows anyone's proper numerical statement of the likelihood of an event to be considered a probability. It recognizes that the quality of PSA's and their results will vary. But, unlike the author, the commentors contend that a PSA should always be a search for truth--not a vehicle for adversarial pleadings

  3. Implementation of probabilistic safety concepts in international codes

    International Nuclear Information System (INIS)

    Borges, J.F.

    1977-01-01

    Recent progress in the implementation of safety concepts in international structure codes is briefly presented. Special attention is paid to the work of the Joint-Committee on Structural Safety. The discussion is centered on some problems such as: safety differentiation, definition and combination of actions, spaces for checking safety and non-linear structural behaviour. When discussing safety differentiation it should be considered that the total probability of failure derives from a theoretical probability of failure and a probability of failure due to error and gross negligence. Optimization of design criteria should take into account both causes of failure. The quantification of reliability implies a probabilistic idealization of all basic variables. Steps taken to obtain an improved definition of different types of actions and rules for their combination are described. Safety checking can be carried out in terms of basic variables, action-effects, or any other suitable variable. However, the advantages and disadvantages of the different types of formulation should be discussed, particularly in the case of non-linear structural behaviour. (orig.) [de

  4. Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Ewing, D.J.F.; Campbell, J.F.

    1994-01-01

    This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

  5. Integrated deterministic and probabilistic safety assessment: Concepts, challenges, research directions

    International Nuclear Information System (INIS)

    Zio, Enrico

    2014-01-01

    Highlights: • IDPSA contributes to robust risk-informed decision making in nuclear safety. • IDPSA considers time-dependent interactions among component failures and system process. • Also, IDPSA considers time-dependent interactions among control and operator actions. • Computational efficiency by advanced Monte Carlo and meta-modelling simulations. • Efficient post-processing of IDPSA output by clustering and data mining. - Abstract: Integrated deterministic and probabilistic safety assessment (IDPSA) is conceived as a way to analyze the evolution of accident scenarios in complex dynamic systems, like nuclear, aerospace and process ones, accounting for the mutual interactions between the failure and recovery of system components, the evolving physical processes, the control and operator actions, the software and firmware. In spite of the potential offered by IDPSA, several challenges need to be effectively addressed for its development and practical deployment. In this paper, we give an overview of these and discuss the related implications in terms of research perspectives

  6. Probabilistic safety considerations for the final disposal of radioactive waste

    International Nuclear Information System (INIS)

    Berg, H.P.; Gruendler, D.; Wurtinger, W.

    1992-01-01

    In order to demonstrate the safety-related balanced concept of the plant design with respect to the operational phase, probabilistic safety considerations were made for the planned German repository for radioactive wastes, the Konrad repository. These considerations are described with respect to the handling and transfer system in the above-ground and underground facility. The operational sequences and the features of a repository are similar to those of conventional transportation and loading facilities and mining techniques. Hence, failure sequences and probability data were derived from these conventional areas. Incidents taken into consideration are e. g. collision of vehicles, fires, drop of waste packages due to failures of lifting equipment. The statistical data used were made available by authorities, insurance companies, and expert organizations. These data have been converted into probability data which were used for the determination of the frequencies for all radiologically relevant incidents. (author)

  7. 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

  8. 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.

  9. Integrated deterministic and probabilistic safety assessment: Concepts, challenges, research directions

    Energy Technology Data Exchange (ETDEWEB)

    Zio, Enrico, E-mail: enrico.zio@ecp.fr [Ecole Centrale Paris and Supelec, Chair on System Science and the Energetic Challenge, European Foundation for New Energy – Electricite de France (EDF), Grande Voie des Vignes, 92295 Chatenay-Malabry Cedex (France); Dipartimento di Energia, Politecnico di Milano, Via Ponzio 34/3, 20133 Milano (Italy)

    2014-12-15

    Highlights: • IDPSA contributes to robust risk-informed decision making in nuclear safety. • IDPSA considers time-dependent interactions among component failures and system process. • Also, IDPSA considers time-dependent interactions among control and operator actions. • Computational efficiency by advanced Monte Carlo and meta-modelling simulations. • Efficient post-processing of IDPSA output by clustering and data mining. - Abstract: Integrated deterministic and probabilistic safety assessment (IDPSA) is conceived as a way to analyze the evolution of accident scenarios in complex dynamic systems, like nuclear, aerospace and process ones, accounting for the mutual interactions between the failure and recovery of system components, the evolving physical processes, the control and operator actions, the software and firmware. In spite of the potential offered by IDPSA, several challenges need to be effectively addressed for its development and practical deployment. In this paper, we give an overview of these and discuss the related implications in terms of research perspectives.

  10. Probabilistic safety assessment in the chemical and nuclear industries

    CERN Document Server

    Fullwood, Ralph R

    2000-01-01

    Probabilistic Safety Analysis (PSA) determines the probability and consequences of accidents, hence, the risk. This subject concerns policy makers, regulators, designers, educators and engineers working to achieve maximum safety with operational efficiency. Risk is analyzed using methods for achieving reliability in the space program. The first major application was to the nuclear power industry, followed by applications to the chemical industry. It has also been applied to space, aviation, defense, ground, and water transportation. This book is unique in its treatment of chemical and nuclear risk. Problems are included at the end of many chapters, and answers are in the back of the book. Computer files are provided (via the internet), containing reliability data, a calculator that determines failure rate and uncertainty based on field experience, pipe break calculator, event tree calculator, FTAP and associated programs for fault tree analysis, and a units conversion code. It contains 540 references and many...

  11. Verification and validation process for the safety software in KNICS

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Lee, Jang-Soo; Kim, Jang-Yeol

    2004-01-01

    This paper describes the Verification and Validation (V and V ) process for safety software of Programmable Logic Controller (PLC), Digital Reactor Protection System (DRPS), and Engineered Safety Feature-Component Control System (ESF-CCS) that are being developed in Korea Nuclear Instrumentation and Control System (KNICS) projects. Specifically, it presents DRPS V and V experience according to the software development life cycle. The main activities of DRPS V and V process are preparation of software planning documentation, verification of Software Requirement Specification (SRS), Software Design Specification (SDS) and codes, and testing of the integrated software and the integrated system. In addition, they include software safety analysis and software configuration management. SRS V and V of DRPS are technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, preparing integrated system test plan, software safety analysis, and software configuration management. Also, SDS V and V of RPS are technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, preparing integrated software test plan, software safety analysis, and software configuration management. The code V and V of DRPS are traceability analysis, source code inspection, test case and test procedure generation, software safety analysis, and software configuration management. Testing is the major V and V activity of software integration and system integration phase. Software safety analysis at SRS phase uses Hazard Operability (HAZOP) method, at SDS phase it uses HAZOP and Fault Tree Analysis (FTA), and at implementation phase it uses FTA. Finally, software configuration management is performed using Nu-SCM (Nuclear Software Configuration Management) tool developed by KNICS project. Through these activities, we believe we can achieve the functionality, performance, reliability and safety that are V

  12. The KNICS approach for verification and validation of safety software

    International Nuclear Information System (INIS)

    Cha, Kyung Ho; Sohn, Han Seong; Lee, Jang Soo; Kim, Jang Yeol; Cheon, Se Woo; Lee, Young Joon; Hwang, In Koo; Kwon, Kee Choon

    2003-01-01

    This paper presents verification and validation (VV) to be approached for safety software of POSAFE-Q Programmable Logic Controller (PLC) prototype and Plant Protection System (PPS) prototype, which consists of Reactor Protection System (RPS) and Engineered Safety Features-Component Control System (ESF-CCS) in development of Korea Nuclear Instrumentation and Control System (KNICS). The SVV criteria and requirements are selected from IEEE Std. 7-4.3.2, IEEE Std. 1012, IEEE Std. 1028 and BTP-14, and they have been considered for acceptance framework to be provided within SVV procedures. SVV techniques, including Review and Inspection (R and I), Formal Verification and Theorem Proving, and Automated Testing, are applied for safety software and automated SVV tools supports SVV tasks. Software Inspection Support and Requirement Traceability (SIS-RT) supports R and I and traceability analysis, a New Symbolic Model Verifier (NuSMV), Statemate MAGNUM (STM) ModelCertifier, and Prototype Verification System (PVS) are used for formal verification, and McCabe and Cantata++ are utilized for static and dynamic software testing. In addition, dedication of Commercial-Off-The-Shelf (COTS) software and firmware, Software Safety Analysis (SSA) and evaluation of Software Configuration Management (SCM) are being performed for the PPS prototype in the software requirements phase

  13. Probabilistic Safety Assessment of Waste from PyroGreen Processes

    International Nuclear Information System (INIS)

    Ju, Hee Jae; Ham, In hye; Hwang, Il Soon

    2016-01-01

    The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

  14. Probabilistic Safety Assessment of Waste from PyroGreen Processes

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Hee Jae; Ham, In hye; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2016-05-15

    The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

  15. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  16. A probabilistic safety analysis of incidents in nuclear research reactors

    International Nuclear Information System (INIS)

    Lopes, V. M.; Sordi, G. M. A. A.; Moralles, M.; Filho, T. M.

    2012-01-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64. (authors)

  17. Uncertainty propagation in probabilistic safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Fleming, P.V.

    1981-09-01

    The uncertainty propagation in probabilistic safety analysis of nuclear power plants, is done. The methodology of the minimal cut is implemented in the computer code SVALON and the results for several cases are compared with corresponding results obtained with the SAMPLE code, which employs the Monte Carlo method to propagate the uncertanties. The results have show that, for a relatively small number of dominant minimal cut sets (n approximately 25) and error factors (r approximately 5) the SVALON code yields results which are comparable to those obtained with SAMPLE. An analysis of the unavailability of the low pressure recirculation system of Angra 1 for both the short and long term recirculation phases, are presented. The results for the short term phase are in good agreement with the corresponding one given in WASH-1400. (E.G.) [pt

  18. Probabilistic safety analysis forecast for Trillo 1 NPP

    International Nuclear Information System (INIS)

    Carretero Fernandino, J.A.; Martin Alvarez, L.; gomez, F.; Cuallado, G.

    1995-01-01

    The performance of Probabilistic Safety Analyses (PSA) at Trillo 1 NPP is facing a number of challenges, unprecedented in previous PSAs carried out in Spain, due to the particular design characteristics of the plant. On account of this, it has been necessary to implemented specific approaches and methodological alternatives to perform a PSA which, while maintaining detail level and requirements in line with PSAs carried out previously in Spain, offers a solution technically adapted to the characteristics of the SIEMENS-KWU design as opposed to other Spanish reactors with a basic Westinghouse-General Electric design, which are based on standard US design. The purpose of this paper is to describe the most significant characteristics of the PSA at Trillo 1 NPP and the methodology used to date, taking into account current project progress

  19. Frequently Asked Questions in Fire Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Kang, Dae Il; Kim, Kil Yoo; Park, Gee Yong

    2010-05-01

    The FAQs(Frequently Asked Questions) in the Fire Probabilistic Safety Assessment(FPSA) are the issues occurred during performing the engineering evaluation based on NFPA-805. In this report, the background and resolutions are reviewed and described for 17 FAQs related to FPSA among 57 FAQs. The current FAQs related to FPSA are the issues concerning to NUREG/CR-6850, and are almost resolved but for the some FAQ, the current resolutions would be changed depending on the results of the future or on-going research. Among FAQs related to FPSA, best estimate approaches are suggested concerning to the conservative method of NUREG/CR-6850. If these best estimate solutions are used in the FPSA of nuclear power plants, realistic evaluation results of fire risk would be obtained

  20. Effects of relay chatter in seismic probabilistic safety analysis

    International Nuclear Information System (INIS)

    Reed, J.W.; Shiu, K.K.

    1985-01-01

    In the Zion and Indian Point Probabilistic Safety Studies, relay chatter was dismissed as a credible event and hence was not formally included in the analyses. Although little discussion is given in the Zion and Indian Point PSA documentation concerning the basis for this decision, it has been expressed informally that it was assumed that the operators will be able to reset all relays in a timely manner. Currently, it is the opinion of many professionals that this may be an oversimplification. The three basic areas which must be considered in addressing relay chatter include the fragility of the relays per se, the reliability of the operators to reset the relays and finally the systems response aspects. Each of these areas is reviewed and the implications for seismic PSA are discussed. Finally, recommendations for future research are given

  1. Application of probabilistic safety assessment for Macedonian electric power system

    International Nuclear Information System (INIS)

    Kancev, D.; Causevski, A.; Cepin, M.; Volkanovski, A.

    2007-01-01

    Due to the complex and integrated nature of a power system, failures in any part of the system can cause interruptions, which range from inconveniencing a small number of local residents to a major and widespread catastrophic disruption of supply known as blackout. The objective of the paper is to show that the methods and tools of probabilistic safety assessment are applicable for assessment and improvement of real power systems. The method used in this paper is developed based on the fault tree analysis and is adapted for the power system reliability analysis. A particular power system i.e. the Macedonian power system is the object of the analysis. The results show that the method is suitable for application of real systems. The reliability of Macedonian power system assumed as the static system is assessed. The components, which can significantly impact the power system are identified and analysed in more details. (author)

  2. Hybrid probabilistic and possibilistic safety assessment. Methodology and application

    International Nuclear Information System (INIS)

    Kato, Kazuyuki; Amano, Osamu; Ueda, Hiroyoshi; Ikeda, Takao; Yoshida, Hideji; Takase, Hiroyasu

    2002-01-01

    This paper presents a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to the safety assessment of geological disposal of high-level radioactive waste. Uncertainties associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts, while variability was formulated by means of probability density functions (pdfs) based on available data sets. The exercise demonstrated the applicability of the new methodology and, in particular, its advantage in quantifying uncertainties based on expert opinion and in providing information on the dependence of assessment results on the level of conservatism. In addition, it was shown that sensitivity analysis can identify key parameters contributing to uncertainties associated with results of the overall assessment. The information mentioned above can be utilized to support decision-making and to guide the process of disposal system development and optimization of protection against potential exposure. (author)

  3. 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

  4. Probabilistic safety assessment past, present and future. An IAEA perspective

    International Nuclear Information System (INIS)

    Lederman, L.; Niehaus, F.; Tomic, B.

    1996-01-01

    Despite the high level of development that probabilistic safety assessment (PSA) methods have reached, a number of issues place constraints on its use in supporting decision making on safety matters. A recent publication of the International Nuclear Safety Advisory Group (INSAG) represents an important step in reaching international consensus on the use of PSA. PSA is ''strongly encouraged'' by INSAG; however, it is noted that ''PSA methodology is not sufficiently mature for its present status to be frozen''. The main aspects of the report are discussed in this paper. The paper next discusses three main categories of PSA application, namely the adequacy of design and procedures, optimization of operational activities and regulatory applications. For each of the applications, the objectives, specific modelling requirements and the prospects for implementation are presented. Consistent with its statutory functions, an important aspect of the work of the IAEA is to reach international consensus on the possibilities of and limitations on the use of PSA methods. Whereas past efforts have been concentrated on promotion and assistance to perform Level 1 PSAs, work is now extending with emphasis on PSA applications, Level 2 and Level 3 analysis, external events and shutdown risks. The main elements of IAEA's PSA Programme are discussed. Finally some challenges related to the use of PSA in the backfitting of nuclear power plants in Eastern Europe and countries of the former USSR are addressed. (orig.)

  5. Probabilistic safety assessment of Narora Atomic Power Project

    International Nuclear Information System (INIS)

    Babar, A.K.; Saraf, R.K.; Kakodkar, A.; Sanyasi Rao, V.V.S.

    1989-01-01

    Various safety studies on Pressurised Water and Boiling Water reactors have been conducted. However, a detailed report on probabilistic safety assessment (PSA) of PHWRs is not available. PSA level I results of the standardised 235 MWe PHWR under construction at Narora are presented herein. Fault Tree analysis of various initiating events (IEs), safety systems has been completed. Event Tree analysis has been performed for all the dominating IEs to identify the accident sequences and a list of the dominating accident sequences is included. Analysis has been carried out using Monte Carlo simulation to propagate the uncertanities in failure rate data. Further uncertainty analysis is extended to obtain distributions for the accident sequences and core damage frequency. Some noteworthy results of the study apart from the various design modifications incorporated during the design phase are: (i) The accident sequences resulting from station blackout are dominant contributors to the core damage frequency. (ii) Class-IV transients, small break LOCA are significant IEs. Main steam line break is likely to induce steam generator tube ruptures. (iii) Moderator circulation, fire fighting system, secondary steam relief are relatively important in core damage frequency reductions. (iv) Under accidental situations human errors are likely to be asociated with valving in shutdown cooling and fire fighting systems. (author). 14 tabs., 14 figs., 15 refs

  6. 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

  7. Regulatory review of probabilistic safety assessment (PSA) Level 2

    International Nuclear Information System (INIS)

    2001-07-01

    Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from deterministic analysis. Many regulatory authorities consider the current state of the art in PSA to be sufficiently well developed for results to be used centrally in the regulatory decision making process-referred to as risk informed regulation. For these applications to be successful, it will be necessary for the regulatory authority to have a high degree of confidence in the PSA. However, at the 1994 IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997, the IAEA and OECD Nuclear Energy Agency agreed to produce, in cooperation, guidance on Regulatory Review of PSA. This led to the publication of IAEA-TECDOC-1135 on the Regulatory Review of Probabilistic Safety Assessment (PSA) Level 1, which gives advice for the review of Level 1 PSA for initiating events occurring at power plants. This TECDOC extends the coverage to address the regulatory review of Level 2 PSA.These publications are intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable level of quality so that it can be used as the

  8. A comparison of integrated safety analysis and probabilistic risk assessment

    International Nuclear Information System (INIS)

    Damon, Dennis R.; Mattern, Kevin S.

    2013-01-01

    The U.S. Nuclear Regulatory Commission conducted a comparison of two standard tools for risk informing the regulatory process, namely, the Probabilistic Risk Assessment (PRA) and the Integrated Safety Analysis (ISA). PRA is a calculation of risk metrics, such as Large Early Release Frequency (LERF), and has been used to assess the safety of all commercial power reactors. ISA is an analysis required for fuel cycle facilities (FCFs) licensed to possess potentially critical quantities of special nuclear material. A PRA is usually more detailed and uses more refined models and data than an ISA, in order to obtain reasonable quantitative estimates of risk. PRA is considered fully quantitative, while most ISAs are typically only partially quantitative. The extension of PRA methodology to augment or supplant ISAs in FCFs has long been considered. However, fuel cycle facilities have a wide variety of possible accident consequences, rather than a few surrogates like LERF or core damage as used for reactors. It has been noted that a fuel cycle PRA could be used to better focus attention on the most risk-significant structures, systems, components, and operator actions. ISA and PRA both identify accident sequences; however, their treatment is quite different. ISA's identify accidents that lead to high or intermediate consequences, as defined in 10 Code of Federal Regulations (CFR) 70, and develop a set of Items Relied on For Safety (IROFS) to assure adherence to performance criteria. PRAs identify potential accident scenarios and estimate their frequency and consequences to obtain risk metrics. It is acceptable for ISAs to provide bounding evaluations of accident consequences and likelihoods in order to establish acceptable safety; but PRA applications usually require a reasonable quantitative estimate, and often obtain metrics of uncertainty. This paper provides the background, features, and methodology associated with the PRA and ISA. The differences between the

  9. Probabilistic safety assessment of nuclear power plants: a monograph

    International Nuclear Information System (INIS)

    Solanki, R.B.; Prasad, Mahendra

    2007-11-01

    This monograph on probabilistic safety assessment (PSA) is addressed to the wide community of professionals engaged in the nuclear industry and concerned with the safety issues of nuclear power plants (NPPs). While the monograph describes PSA of NPPs, the principles described in this monograph can be extended to other facilities like spent fuel storage, fuel reprocessing plants and non-nuclear facilities like chemical plants, refineries etc. as applicable. The methodology for risk assessment in chemical plants or refineries is generally known as quantitative risk analysis (QRA). The fundamental difference between NPP and chemical plant is that in NPPs the hazardous material (fuel and fission products) are contained at a single location (i.e. inside containment), whereas in a chemical plant and reprocessing plants, the hazardous material is present simultaneously at many places, like pipelines, reaction towers, storage tanks, etc. Also unlike PSA, QRA does not deal with levels; it uses an integrated approach combining all the levels. The monograph covers the areas of broad interest in the field of PSA such as historical perspective, fundamentals of PSA, strengths and weaknesses of PSA, applications of PSA, role of PSA in the regulatory decision making and issues for advancement of PSA

  10. Qualitative uncertainty analysis in probabilistic safety assessment context

    International Nuclear Information System (INIS)

    Apostol, M.; Constantin, M; Turcu, I.

    2007-01-01

    In Probabilistic Safety Assessment (PSA) context, an uncertainty analysis is performed either to estimate the uncertainty in the final results (the risk to public health and safety) or to estimate the uncertainty in some intermediate quantities (the core damage frequency, the radionuclide release frequency or fatality frequency). The identification and evaluation of uncertainty are important tasks because they afford credit to the results and help in the decision-making process. Uncertainty analysis can be performed qualitatively or quantitatively. This paper performs a preliminary qualitative uncertainty analysis, by identification of major uncertainty in PSA level 1- level 2 interface and in the other two major procedural steps of a level 2 PSA i.e. the analysis of accident progression and of the containment and analysis of source term for severe accidents. One should mention that a level 2 PSA for a Nuclear Power Plant (NPP) involves the evaluation and quantification of the mechanisms, amount and probabilities of subsequent radioactive material releases from the containment. According to NUREG 1150, an important task in source term analysis is fission products transport analysis. The uncertainties related to the isotopes distribution in CANDU NPP primary circuit and isotopes' masses transferred in the containment, using SOPHAEROS module from ASTEC computer code will be also presented. (authors)

  11. Co-simulation for real time safety verification of nuclear power plants

    International Nuclear Information System (INIS)

    Boafo, E.K.; Zhang, L.; Nasimi, E.; Gabbar, H.A.

    2015-01-01

    Small and major accidents and near misses are still occurring in nuclear power plants (NPPs). Risk level has increased with the degradation of NPP equipment and instrumentations. In order to achieve NPP safety, it is important to continuously evaluate risk for all potential hazard and fault propagation scenarios and map protection layers to fault / failure / hazard propagation scenarios to be able to evaluate and verify safety level during NPP operation. There are major limitations in current real time safety verification tools, as it is mainly offline and with no integration to NPP simulation tools. The main goal of this research is to develop real time safety verification with co-simulation tool to be integrated with plant operation support systems. This includes the development of static and dynamic fault semantic network (FSN) to model all possible fault propagation scenarios and the interrelationships among associated process variables. Safety and protection layers along with their reliability are mapped to FSN so that safety levels can be verified during plant operation. Errors between multiphysics models and real time data are modeled to accurately and dynamically tune FSN for each fault propagation scenario. The detailed methodology will show how to integrate process models, construction of static FSN with fault propagation scenarios, and evaluation and tuning of dynamic FSN with probabilistic and process variable interaction values. Principle Component Analysis method is used reduce dimensionality and reduce process variables associated with each fault scenario. Then map independent protection layers (IPL) to FSN with estimated reliability measures of each protection layer to accurately verify safety for different operational scenarios. Intelligent algorithms is used with multivariate techniques to accurate define the interrelation among process variables, in terms of signal strength and time delay, using Genetic Programming (GP), which will provide basis

  12. Incorporating organizational factors into probabilistic safety assessment of nuclear power plants through canonical probabilistic models

    Energy Technology Data Exchange (ETDEWEB)

    Galan, S.F. [Dpto. de Inteligencia Artificial, E.T.S.I. Informatica (UNED), Juan del Rosal, 16, 28040 Madrid (Spain)]. E-mail: seve@dia.uned.es; Mosleh, A. [2100A Marie Mount Hall, Materials and Nuclear Engineering Department, University of Maryland, College Park, MD 20742 (United States)]. E-mail: mosleh@umd.edu; Izquierdo, J.M. [Area de Modelado y Simulacion, Consejo de Seguridad Nuclear, Justo Dorado, 11, 28040 Madrid (Spain)]. E-mail: jmir@csn.es

    2007-08-15

    The {omega}-factor approach is a method that explicitly incorporates organizational factors into Probabilistic safety assessment of nuclear power plants. Bayesian networks (BNs) are the underlying formalism used in this approach. They have a structural part formed by a graph whose nodes represent organizational variables, and a parametric part that consists of conditional probabilities, each of them quantifying organizational influences between one variable and its parents in the graph. The aim of this paper is twofold. First, we discuss some important limitations of current procedures in the {omega}-factor approach for either assessing conditional probabilities from experts or estimating them from data. We illustrate the discussion with an example that uses data from Licensee Events Reports of nuclear power plants for the estimation task. Second, we introduce significant improvements in the way BNs for the {omega}-factor approach can be constructed, so that parameter acquisition becomes easier and more intuitive. The improvements are based on the use of noisy-OR gates as model of multicausal interaction between each BN node and its parents.

  13. Incorporating organizational factors into probabilistic safety assessment of nuclear power plants through canonical probabilistic models

    International Nuclear Information System (INIS)

    Galan, S.F.; Mosleh, A.; Izquierdo, J.M.

    2007-01-01

    The ω-factor approach is a method that explicitly incorporates organizational factors into Probabilistic safety assessment of nuclear power plants. Bayesian networks (BNs) are the underlying formalism used in this approach. They have a structural part formed by a graph whose nodes represent organizational variables, and a parametric part that consists of conditional probabilities, each of them quantifying organizational influences between one variable and its parents in the graph. The aim of this paper is twofold. First, we discuss some important limitations of current procedures in the ω-factor approach for either assessing conditional probabilities from experts or estimating them from data. We illustrate the discussion with an example that uses data from Licensee Events Reports of nuclear power plants for the estimation task. Second, we introduce significant improvements in the way BNs for the ω-factor approach can be constructed, so that parameter acquisition becomes easier and more intuitive. The improvements are based on the use of noisy-OR gates as model of multicausal interaction between each BN node and its parents

  14. Safety-specific benefit of the probabilistic evaluation of older nuclear power plants

    International Nuclear Information System (INIS)

    Hoertner, H.; Koeberlein, K.

    1991-01-01

    The report summarizes the experience of the GRS obtained within the framework of a probabilistic evaluation of older nuclear power plants and the German risk study. The applied methodology and the problems involved are explained first. After a brief summary of probabilistic analyses carried out for German nuclear power plants, reliability analyses for older systems are discussed in detail. The findings from the probabilistic safety analyses and the conclusions drawn are presented. (orig.) [de

  15. Probabilistic analysis of safety in industrial irradiation plants

    International Nuclear Information System (INIS)

    Alderete, F.; Elechosa, C.

    2006-01-01

    The Argentinean Nuclear Regulatory Authority is carrying out the Probabilistic Safety Analysis (PSA) of the two industrial irradiation plants existent in the country. The objective of this presentation is to show from the regulatory point of view, the advantages of applying this tool, as well as the appeared difficulties; for it will be made a brief description of the facilities, of the method and of the normative one. Both plants are multipurpose facilities classified as 'industrial irradiator category IV' (panoramic irradiator with source deposited in pool). Basically, the execution of an APS consists of the following stages: 1. Identification of initiating events. 2. Modeling of Accidental Sequences (Event Trees). 3. Analysis of Systems (Fault trees). 4. Quantification of Accidental Sequences. The argentine normative doesn't demand to these facilities the realization of an APS, however the basic standard of Radiological Safety establishes that in the design of this type of facilities in the cases that is justified, should make sure that the annual probability of occurrence of an accidental sequence and the resulting dose in a person gives as result an radiological risk inferior to the risk limit adopted as acceptance criteria. On the other hand the design standard specifies for these irradiators it demands a maximum fault rate of 10 -2 for the related components with the systems of radiological safety. In our case, the possible initiating events have been identified that carried out to not wanted situations (about people exposure, radioactive contamination). Then, for each one of the significant initiating events, the corresponding accidental sequences were modeled and the safety systems that intervene in this sequences by means of fault trees were analyzed, for then to determine the fault probabilities of the same ones. At the moment they are completing these fault trees, but the difficulty resides in the impossibility of obtaining real data of the reliability

  16. The use of probabilistic safety assessments for improving nuclear safety in Europe

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1992-01-01

    The political changes in Europe broadened the scope of international nuclear safety matters considerably. The Western world started to receive reliable and increasingly detailed information on Eastern European nuclear technology and took note of a broad range of technical and administrative problems relevant for nuclear safety in these countries. Reunification made Germany a focus of information exchange on these matters. Here, cooperation with the former German Democratic Republic and with other Eastern European countries as well as safety analyses of Soviet-built nuclear power plants started rather early. Meanwhile, these activities are progressing toward all-European cooperation in the nuclear safety sector. This cooperation includes the use of probabilistic safety assessments (PSAs) addressing applications in both Western and Eastern Europe as well as the further development of this methodology in a converging Europe

  17. A Probabilistic Mass Estimation Algorithm for a Novel 7- Channel Capacitive Sample Verification Sensor

    Science.gov (United States)

    Wolf, Michael

    2012-01-01

    A document describes an algorithm created to estimate the mass placed on a sample verification sensor (SVS) designed for lunar or planetary robotic sample return missions. A novel SVS measures the capacitance between a rigid bottom plate and an elastic top membrane in seven locations. As additional sample material (soil and/or small rocks) is placed on the top membrane, the deformation of the membrane increases the capacitance. The mass estimation algorithm addresses both the calibration of each SVS channel, and also addresses how to combine the capacitances read from each of the seven channels into a single mass estimate. The probabilistic approach combines the channels according to the variance observed during the training phase, and provides not only the mass estimate, but also a value for the certainty of the estimate. SVS capacitance data is collected for known masses under a wide variety of possible loading scenarios, though in all cases, the distribution of sample within the canister is expected to be approximately uniform. A capacitance-vs-mass curve is fitted to this data, and is subsequently used to determine the mass estimate for the single channel s capacitance reading during the measurement phase. This results in seven different mass estimates, one for each SVS channel. Moreover, the variance of the calibration data is used to place a Gaussian probability distribution function (pdf) around this mass estimate. To blend these seven estimates, the seven pdfs are combined into a single Gaussian distribution function, providing the final mean and variance of the estimate. This blending technique essentially takes the final estimate as an average of the estimates of the seven channels, weighted by the inverse of the channel s variance.

  18. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  19. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  20. 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

  1. 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

  2. Bridging probabilistic safety assessment studies with information Management System

    International Nuclear Information System (INIS)

    Luanco, E. M.

    2010-01-01

    Probabilistic Safety Assessment (PSA) is a critical business often known in conjunction with either new build or life extension of nuclear power plant. However, it is not so often referred to the operation phase of the plant, although it could bring a lot of long term benefits to the operator. The purpose of this paper is to discuss the potential contribution of PSA with day to day operation in bridging the deficiencies and specific failures characteristics of critical Structure System and Component (SSC) with the results of PSA studies. From and Information System prospective, the use of Information Management system (IMS) -also known as EAM solution -widely used by the majority of nuclear operators- is the potential vehicle to bridge the 2 worlds of PSA and daily operation. Most EAM solution get reliability management functionalities which are not really integrated with PSA tools and data and thus cannot provide the anticipated benefits of addressing typical aging phenomena beyond the only predictive models used by the PSA studies. The paper will also discuss potential integration scenario between PSA tools and EAM solutions. (authors)

  3. Application of probabilistic safety assessment to Rokkasho reprocessing plant, (2)

    International Nuclear Information System (INIS)

    Miyata, Takashi; Takebe, Kazumi; Tamauchi, Yoshikazu

    2008-01-01

    A probabilistic safety assessment (PSA) is made on the boiling accident of a highly active liquid waste tank, which may result in significant consequences, in accordance with the procedure for PSA developed for nuclear power plants. Obtained as results are the frequency of boiling accident of a certain tank of 2.0x10 -8 /y (frequency of boiling accident of any tank of 4.1x10 0-8 /y), its error factor of approx. 6, and information on the relative risk importance based on the FV index and RAW for various components, systems and activities of personnel and on the sensitivity of key parameters. Furthermore, the effect of the time required for repairing failed instruments on the frequency of accident, how to deal with the common cause of failure of the duplicated dynamic components, one of which is at least in operation, and conservative exposure dose in the event of an accident are examined. The database for the Rokkasho reprocessing plant has not been established yet, but the PSA results utilizing available failure rate databases of existing nuclear power plants and reprocessing plants in Japan and abroad can be used effectively to optimize operations and maintenance, if they are interpreted properly and some uncertainties are taken into account. (author)

  4. BURD, Bayesian estimation in data analysis of Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Jang, Seung-cheol; Park, Jin-Kyun

    2008-01-01

    1 - Description of program or function: BURD (Bayesian Update for Reliability Data) is a simple code that can be used to obtain a Bayesian estimate easily in the data analysis of PSA (Probabilistic Safety Assessment). According to the Bayes' theorem, basically, the code facilitates calculations of posterior distribution given the prior and the likelihood (evidence) distributions. The distinctive features of the program, BURD, are the following: - The input consists of the prior and likelihood functions that can be chosen from the built-in statistical distributions. - The available prior distributions are uniform, Jeffrey's non informative, beta, gamma, and log-normal that are most-frequently used in performing PSA. - For likelihood function, the user can choose from four statistical distributions, e.g., beta, gamma, binomial and poisson. - A simultaneous graphic display of the prior and posterior distributions facilitate an intuitive interpretation of the results. - Export facilities for the graphic display screen and text-type outputs are available. - Three options for treating zero-evidence data are provided. - Automatic setup of an integral calculus section for a Bayesian updating. 2 - Methods: The posterior distribution is estimated in accordance with the Bayes' theorem, given the prior and the likelihood (evidence) distributions. 3 - Restrictions on the complexity of the problem: The accuracy of the results depends on the calculational error of the statistical function library in MS Excel

  5. Probabilistic safety assessment for Balakovo 1000 MW NPP

    International Nuclear Information System (INIS)

    Foden, R.W.

    1995-01-01

    In July 1993 the Commission of the European Communities (CEC) placed a contract with NNC Ltd (National Nuclear Corporation) for performing a Probabilistic Safety Assessment (PSA) for a 1000 MW NPP in the Russian Federation. The contract is part (Project 3.1) of the 1991 TACIS (Technical Assistance to the CIS) programme. This paper describes the objectives and scope of the Project and provides a description of the progress that has been made. For this Project, NNC is the leader of a Consortium of Western European companies that has been formed to undertake this Project and other Projects in the TACIS 91 programme. NNC therefore has overall responsibility for the coordination and management of the complete PSA Project. Other members of the Consortium involved in this Project are Empresarios Agrupados from Spain, Belgatom from Belgium and AEA-Technology from the UK. The analytical work for the Project is performed by the Russian Company Atomenergoproekt in Moscow, under contract to NNC. The official recipient institution for the results of the Project is the Russian Utility, Rosenergatom. The NPP chosen to be the subject of the Project is the Balakovo Unit 4 VVER 1000. (author)

  6. Application of probabilistic safety assessment to research reactors

    International Nuclear Information System (INIS)

    1989-07-01

    This document has been prepared to assist in the performance of a research reactor probabilistic safety assessment (PSA). It offers examples of experience gained by a number of Member States in carrying out PSA for research reactors. These examples are illustrative of the types of approach adopted, the problems that arise and the judgements entered into when conducting a PSA. The illustrative examples of experiences gained are discussed in a series of thirteen chapters which address some of the issues that arise in a PSA. The examples are not exhaustive and offer evidence of how other analyses have approached the task of preparing a PSA, for their particular plant. The principles should be capable of being utilised and the various issues which are discussed should be translated into the needs of the analyst. Each PSA will make its own demands on the analyst depending on the reactor and so the illustrations must only be used as guidance and not adopted as published, without critical appreciation. Refs, figs and tabs

  7. Probabilistic safety analysis of transportation of spent fuel

    International Nuclear Information System (INIS)

    Subramaniam, Chitra

    1999-11-01

    The report presents the results of the study carried out to estimate the accident risk involved in the transport of spent fuel from Rajasthan Atomic Power Station near Kota to the fuel reprocessing plant at Tarapur. The technique of probabilistic safety analysis is used. The fuel considered is the Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally-cooled shipping cask over a distance of 822 km by rail. The Indian rail accident statistics are used to estimate the basic rail accident frequency. The possible ways in which a release of radioactive material can occur from the spent fuel cask are identified by the fault tree analysis technique. The release sequences identified are classified into eight accident severity categories, and release fractions are assigned to each. The consequences resulting from the release are estimated by the computer code RADTRAN 4. Results of the risk analysis indicate that the accident risk values are very low and hence acceptable. Parametric studies show that the risk would continue to be small even if the controlling parameters were to simultaneously take extreme adverse values. (author)

  8. Probabilistic safety assessment of the PLUTO Research Reactor

    International Nuclear Information System (INIS)

    Preston, J.F.; Coates, D.A.

    1990-01-01

    The preliminary finding of a probabilistic safety assessment (PSA) carried out in support of a licensing submission are presented. The research reactor, a 25 MW highly enriched thermal reactor moderated and cooled by D 2 O, is housed in a steel containment building equipped with an active extract system to mitigate any possible release. A full PSA (to level 3) was performed based on the current operational plant making as much use of the plant operational records as possible. A medium sized event tree-fault tree approach was used to allow realistic modelling of operator actions. For reasons of practicality only plant damage states of core melt, fuel damage, and tritium release were defined, all release accident sequences being assigned to one of these states. Prior to discharge to the environment the releases were further sub-divided dependent upon the success of the active extract system. The individual and societal risks were calculated taking account of meterological and demographic conditions. The provisional results indicate that the core melt frequency is in the region of 1 x 10 -4 /yr, the dominant contributor being an unisolatable gross leakage beyond the capabilities of the recovery systems. The core melt frequency is comparable with those of power reactors of a similar age; however, the core inventory and hence release is much smaller; therefore the consequences are much reduced. The risk to an individual at any fixed location 100 m from the plant is assessed as 1 x 10 -6 ; the societal risk is estimated as 6 x 10 -4 . The main contributor to the dose received is from the released iodine. Additional benefit is being obtained from the PSA in several ways: the insights obtained into the function and operation are being incorporated into the operational safety document, whilst the source term results are being used to assist in the refurbishment/improvement of the active extract system

  9. Use and development of probabilistic safety assessment - CSNI WGRISK

    International Nuclear Information System (INIS)

    Siu, Nathan; Monninger, John; Gomez-Cobo, Ana; Kao, Tsu-Mu; Schoen, Gerhard; Gunsell, Lars; Nyman, Ralph; Jelinek, Tomas; Hultquist, Goeran; Rapp, Anders; Eriksson, Stefan; Lantaron, Alfredo; Vojnovic, Djordje; Husarcek, Jan; Kovacs, Zoltan; Versteeg, M.F.; Lopez Morones, Ramon; Lee, Chang-Ju; Fukuda, Mamoru; Burgazzi, Luciano; Caporali, Rino; RoeWEKAMP, Marina; MACSUGA, Geza; Bareith, Attila; Lanore, J.M.; Sorel, Vincent; Virolainen, Reino; Patrik, Milan; Mlady, Ondrej; Raducu, Gheorghe; De Gelder, Pieter; Hendrickx, Isabelle; Lanore, Jeanne-Marie; Murphy, Joseph A.; Shepherd, Charles; Pyy, Pekka T.; Mauny, Elisabeth

    2007-01-01

    The CSNI WGRISK produced a report in July 2002 on 'The Use and Development of Probabilistic Safety Assessment in NEA Member Countries'. This provides a description of the PSA programmes in the member countries at the time that the report was produced. However, there have been significant developments in PSA since 2002. Consequently, a decision was made at the WGRISK meeting in October 2005 to produce an updated version of the report. The aim was to produce an updated, stand alone version of the report that presents an analysis of the position on the use and development of PSA in the WGRISK member countries as of spring 2006. A detailed questionnaire was circulated to WGRISK members and to the IAEA to ascertain the state of the art in PSA use and development at the end of 2006. Detailed responses were prepared by 20 countries totalling several hundred pages of information. After first compilation of information, an updating round was organized by showing to the countries all the answers and the summary made of them by a small group of experts. The process led to some clarifications and more consistency in the report. The collected information was finally analyzed and summarized to reach the conclusions presented in this report. The set of section headings in the report is as follows: Executive summary. 1. Introduction. 2. PSA Framework and Environment. 3. Numerical Safety Criteria. 4. PSA Standards and Guidance. 5. Status and Scope of PSA Programmes. 6. PSA Methodology and Data. 7. PSA Applications. 8. Results and Insights from the PSAs. 9. Future Developments. Appendix A: Overview of the Status of PSA Programmes. Appendix B: Contact information. Appendix C: Questionnaire and Guidance to authors

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  11. Use of a probabilistic safety study in the design of the Italian reference PWR

    International Nuclear Information System (INIS)

    Richardson, D.C.; Russino, G.; Valentini, V.

    1985-01-01

    The intent of this paper is to provide a description of the experience gained in having performed a Probabilistic Safety Study (PSS) on the proposed Italian reference pressurized water reactor. The experience revealed that through careful application of probabilistic techniques, Probabilistic Risk Assessment (PRA) can be used as a tool to develop an optimum plant design in terms of safety and cost. Furthermore, the PSS can also be maintained as a living document and a tool to assess additional regulatory requirements that may be imposed during the construction and operational life of the plant. Through the use of flexible probabilistic techniques, the probabilistic safety model can provide a living safety assessment starting from the conceptual design and continuing through the construction, testing and operational phases. Moreover, the probabilistic safety model can be used during the operational phase of the plant as a method to evaluate the operational experience and identify potential problems before they occur. The experience, overall, provided additional insights into the various aspects of the plants design and operation that would not have been identified through the use of traditional safety evaluation techniques

  12. Technique for unit testing of safety software verification and validation

    International Nuclear Information System (INIS)

    Li Duo; Zhang Liangju; Feng Junting

    2008-01-01

    The key issue arising from digitalization of the reactor protection system for nuclear power plant is how to carry out verification and validation (V and V), to demonstrate and confirm the software that performs reactor safety functions is safe and reliable. One of the most important processes for software V and V is unit testing, which verifies and validates the software coding based on concept design for consistency, correctness and completeness during software development. The paper shows a preliminary study on the technique for unit testing of safety software V and V, focusing on such aspects as how to confirm test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed here was successfully used in the work of unit testing on safety software of a digital reactor protection system. (authors)

  13. A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sunghyun; Kim, Namyeong; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant.

  14. A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Park, Sunghyun; Kim, Namyeong; Jae, Moosung

    2015-01-01

    In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant

  15. Regulatory review of probabilistic safety assessment (PSA) level 1

    International Nuclear Information System (INIS)

    2000-02-01

    Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from the deterministic analysis. Many regulatory authorities consider that the current state of the art in PSA (especially Level 1 PSA) is sufficiently well developed that it can be used centrally in the regulatory decision making process - referred to as 'risk informed regulation'. For these applications to be successful, it will be necessary for regulatory authorities to have a high degree of confidence in PSA. However, at the IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process in 1994 and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' Meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997 the IAEA and OECD Nuclear Energy Agency agreed to produce in co-operation a technical document on the regulatory review of PSA. This publication is intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable standard so that it can be used as the basis for taking risk informed decisions within a regulatory decision making process. The document gives guidance on how to set about reviewing a PSA and on the technical issues that need to be addressed. This publication gives guidance for the review of Level 1 PSA for

  16. Response to "Improving Patient Safety With Error Identification in Chemotherapy Orders by Verification Nurses"
.

    Science.gov (United States)

    Zhu, Ling-Ling; Lv, Na; Zhou, Quan

    2016-12-01

    We read, with great interest, the study by Baldwin and Rodriguez (2016), which described the role of the verification nurse and details the verification process in identifying errors related to chemotherapy orders. We strongly agree with their findings that a verification nurse, collaborating closely with the prescribing physician, pharmacist, and treating nurse, can better identify errors and maintain safety during chemotherapy administration.

  17. Applications of probabilistic safety assessment (PSA) for nuclear power plants

    International Nuclear Information System (INIS)

    2001-02-01

    This report, which compiles information on a comprehensive set of PSA applications in the areas of NPP design, operation, and accident mitigation and management, is the culmination of an IAEA project on PSA Applications and Tools to Improve NPP Safety. In this regard, the Technical Committee Meeting (TCM) held in Madrid in February 1998 allowed participants to review and provide very valuable comments for this report. Several important facts related to PSA and its applications were highlighted during this TCM: living PSAs are the basis for the risk informed approach to decision making; development and use of safety/risk monitors as tools for configuration management is spreading fast; the different uses of PSA to support NPP testing and maintenance planning and optimization are amongst the most widespread PSA applications; plant specific PSAs are being used to support the safety upgrading programmes of plants built to earlier standards; not all countries have a regulatory framework for the use of the probabilistic approach in decision making. Some countries are still far from 'risk-informed' regulation, and this means that there is still considerable work ahead, both for regulators and utilities, to clarify approaches, to establish a framework and to reach a common understanding in relation to the use of PSA in decision making. This report is based on the premise that the use of PSA can provide useful information for the decision maker. This report is intended to provide an overview of current PSA applications. Section 2 addresses the PSA application process, outlines the general requirements for PSA tools and provides a discussion on PSA aspects such as PSA level, scope and level of detail, which have to be considered when planning/performing PSA applications. Section 3 discusses the technical aspects of individual applications and is divided into three parts. Section 3.1 is dedicated to the design related PSA applications. The second part of Section 3 considers

  18. Technical Issues and Proposes on the Legislation of Probabilistic Safety Assessment in Periodic Safety Review

    International Nuclear Information System (INIS)

    Hwang, Seok-Won; Jeon, Ho-Jun; Na, Jang-Hwan

    2015-01-01

    Korean Nuclear Power Plants have performed a comprehensive safety assessment reflecting design and procedure changes and using the latest technology every 10 years. In Korea, safety factors of PSR are revised to 14 by revision of IAEA Safety Guidelines in 2003. In the revised safety guidelines, safety analysis field was subdivided into deterministic safety analysis, PSA (Probabilistic safety analysis), and hazard analysis. The purpose to examine PSA as a safety factor on PSR is to make sure that PSA results and assumptions reflect the latest state of NPPs, validate the level of computer codes and analytical models, and evaluate the adequacy of PSA instructions. In addition, its purpose is to derive the plant design change, operating experience of other plants and safety enhancement items as well. In Korea, PSA is introduced as a new factor. Thus, the overall guideline development and long-term implementation strategy are needed. Today in Korea, full-power PSA model revision and low-power and shutdown (LPSD) PSA model development is being performed as a part of the post Fukushima action items for operating plants. The scope of the full-power PSA is internal/external level 1, 2 PSA. But in case of fire PSA, the scope is level 1 PSA using new method, NUREG/CR-6850. In case of LPSD PSA, level 1 PSA for all operating plants, and level 2 PSA for 2 demonstration plants are under development. The result of the LPSD PSA will be used as major input data for plant specific SAMG (Severe Accident Management Guideline). The scope of PSA currently being developed in Korea cannot fulfill 'All Mode, All Scope' requirements recommended in the IAEA Safety Guidelines. Besides the legislation of PSA, step-by-step development strategy for non-performed scopes such as level 3 PSA and new fire PSA is one of the urgent issues in Korea. This paper suggests technical issues and development strategies for each PSA technical elements.

  19. Project for the completion of a probabilistic safety analysis of an industrial irradiation

    International Nuclear Information System (INIS)

    Ferro, R.; Troncoso, M.

    1995-01-01

    The probabilistic safety analysis is a very valuable instrument in safety studies of facilities with potential risk for the personnel, population and environment. One of the possible field of use of PSA techniques in the safety studies for industrial irradiation where serious accidents have occurred. For this reason a project has been undertaken to carry out the PSA in the Irradiation Plant of Research Institute of the Food Industry, which complements the safety studies of this facility

  20. Binary Decision Tree Development for Probabilistic Safety Assessment Applications

    International Nuclear Information System (INIS)

    Simic, Z.; Banov, R.; Mikulicic, V.

    2008-01-01

    The aim of this article is to describe state of the development for the relatively new approach in the probabilistic safety analysis (PSA). This approach is based on the application of binary decision diagrams (BDD) representation for the logical function on the quantitative and qualitative analysis of complex systems that are presented by fault trees and event trees in the PSA applied for the nuclear power plants risk determination. Even BDD approach offers full solution comparing to the partial one from the conventional quantification approach there are still problems to be solved before new approach could be fully implemented. Major problem with full application of BDD is difficulty of getting any solution for the PSA models of certain complexity. This paper is comparing two approaches in PSA quantification. Major focus of the paper is description of in-house developed BDD application with implementation of the original algorithms. Resulting number of nodes required to represent the BDD is extremely sensitive to the chosen order of variables (i.e., basic events in PSA). The problem of finding an optimal order of variables that form the BDD falls under the class of NP-complete complexity. This paper presents an original approach to the problem of finding the initial order of variables utilized for the BDD construction by various dynamical reordering schemes. Main advantage of this approach compared to the known methods of finding the initial order is with better results in respect to the required working memory and time needed to finish the BDD construction. Developed method is compared against results from well known methods such as depth-first, breadth-first search procedures. Described method may be applied in finding of an initial order for fault trees/event trees being created from basic events by means of logical operations (e.g. negation, and, or, exclusive or). With some testing models a significant reduction of used memory has been achieved, sometimes

  1. Probabilistic safety assessment model in consideration of human factors based on object-oriented bayesian networks

    International Nuclear Information System (INIS)

    Zhou Zhongbao; Zhou Jinglun; Sun Quan

    2007-01-01

    Effect of Human factors on system safety is increasingly serious, which is often ignored in traditional probabilistic safety assessment methods however. A new probabilistic safety assessment model based on object-oriented Bayesian networks is proposed in this paper. Human factors are integrated into the existed event sequence diagrams. Then the classes of the object-oriented Bayesian networks are constructed which are converted to latent Bayesian networks for inference. Finally, the inference results are integrated into event sequence diagrams for probabilistic safety assessment. The new method is applied to the accident of loss of coolant in a nuclear power plant. the results show that the model is not only applicable to real-time situation assessment, but also applicable to situation assessment based certain amount of information. The modeling complexity is kept down and the new method is appropriate to large complex systems due to the thoughts of object-oriented. (authors)

  2. Applications of nuclear safety probabilistic risk assessment to nuclear security for optimized risk mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Donnelly, S.K.; Harvey, S.B. [Amec Foster Wheeler, Toronto, Ontario (Canada)

    2016-06-15

    Critical infrastructure assets such as nuclear power generating stations are potential targets for malevolent acts. Probabilistic methodologies can be applied to evaluate the real-time security risk based upon intelligence and threat levels. By employing this approach, the application of security forces and other protective measures can be optimized. Existing probabilistic safety analysis (PSA) methodologies and tools employed. in the nuclear industry can be adapted to security applications for this purpose. Existing PSA models can also be adapted and enhanced to consider total plant risk, due to nuclear safety risks as well as security risks. By creating a Probabilistic Security Model (PSM), safety and security practitioners can maximize the safety and security of the plant while minimizing the significant costs associated with security upgrades and security forces. (author)

  3. Probabilistic safety analysis for control rod drive system of ET-RR-1

    International Nuclear Information System (INIS)

    Nasr, M.; Nasser, O.

    1988-01-01

    The International Atomic Energy Agency (IAEA) co-ordinated a Research programme on Probabilistic Safety Analysis (PSA) for research reactors; with the participation of several countries. In the framework of this project (Project Int. 9/063) the Egyptian Atomic Energy Authority decided to perform a PSA study on the ET-RR-1 (Egypt Thermal Research Reactor). The study is conducted in collaboration between the nuclear regulatory and safety centre (NRSC) and the reactor department of the nuclear research centre at Inchass. The present work is a part of the PSA study on ET-RR- it is concerning a probabilistic safety analysis of the control rod drive mechanism

  4. Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this Safety Guide is to provide recommendations for meeting the IAEA safety requirements in performing or managing a level 2 probabilistic safety assessment (PSA) project for a nuclear power plant; thus it complements the Safety Guide on level 1 PSA. One of the aims of this Safety Guide is to promote a standard framework, standard terms and a standard set of documents for level 2 PSAs to facilitate regulatory and external peer review of their results. It describes all elements of the level 2 PSA that need to be carried out if the starting point is a fully comprehensive level 1 PSA. Contents: 1. Introduction; 2. PSA project management and organization; 3. Identification of design aspects important to severe accidents and acquisition of information; 4. Interface with level 1 PSA: Grouping of sequences; 5. Accident progression and containment analysis; 6. Source terms for severe accidents; 7. Documentation of the analysis: Presentation and interpretation of results; 8. Use and applications of the PSA; Annex I: Example of a typical schedule for a level 2 PSA; Annex II: Computer codes for simulation of severe accidents; Annex III: Sample outline of documentation for a level 2 PSA study.

  5. Probabilistic assessment methods as a tool for developing nations to make safety decisions

    International Nuclear Information System (INIS)

    Gumley, P.; Inamdar, S.V.

    1985-01-01

    This paper advocates the use of probabilistic safety assessment methods in making safety decisions. It discusses the question of adequate safety - what it means to a country buying a nuclear power plant, and how probabilistic safety assessment studies of the reference plant can be used for ensuring this adequate safety. It is proposed that adequate safety means ensuring that the plant would behave, in accident conditions, in a manner similar to the way it is expected to behave were it in the country of origin. For this one needs to know how the plant responds under somewhat altered conditions. These altered conditions can arise from such factors as varying reliability of electrical grids, different manufacturing technology, local systems design and operator capability. In the design of nuclear power plants, the traditional approach to safety has led to the belief that availability and effectiveness of safety systems alone are all that is required to ensure plant safety. This belief can result in design oversights leading to potential problems arising from the power production systems and the service systems. Participation by the buying country in the design of such systems, and understanding the safety implications thereof, can be facilitated by probabilistic safety assessment methods. This philosophy is illustrated in this paper by examples. (author)

  6. Application of verification and validation on safety parameter display systems

    International Nuclear Information System (INIS)

    Thomas, N.C.

    1983-01-01

    Offers some explanation of how verification and validation (VandV) can support development and licensing of the Safety Parameter Display Systems (SPDS). Advocates that VandV can be more readily accepted within the nuclear industry if a better understanding exists of what the objectives of VandV are and should be. Includes a discussion regarding a reasonable balance of costs and benefits of VandV as applied to the SPDS and to other digital systems. Represents the author's perception of the regulator's perspective based on background information and experience, and discussions with regulators about their current concerns and objectives. Suggests that the introduction of the SPDS into the Control Room is a first step towards growing dependency on use of computers

  7. Comparison between Canadian probabilistic safety assessment methods formulated by Atomic Energy of Canada limited and probabilistic risk assessment methods

    International Nuclear Information System (INIS)

    Shapiro, H.S.; Smith, J.E.

    1989-01-01

    The procedures used by Atomic Energy of Canada Limited (AECL) to perform probabilistic safety assessments (PRAs) differ somewhat from conventionally accepted probabilistic risk assessment (PRA) procedures used elsewhere. In Canada, PSA is used by AECL as an audit tool for an evolving design. The purpose is to assess the safety of the plant in engineering terms. Thus, the PSA procedures are geared toward providing engineering feedback so that necessary changes can be made to the design at an early stage, input can be made to operating procedures, and test and maintenance programs can be optimized in terms of costs. Most PRAs, by contrast, are performed in plants that are already built. Their main purpose is to establish the core melt frequency and the risk to the public due to core melt. Also, any design modification is very expensive. The differences in purpose and timing between PSA and PRA have resulted in differences in methodology and scope. The PSA procedures are used on all plants being designed by AECL

  8. Probabilistic safety assessment of the dual-cooled waste transmutation blanket for the FDS-I

    International Nuclear Information System (INIS)

    Hu, L.; Wu, Y.

    2006-01-01

    The subcritical dual-cooled waste transmutation (DWT) blanket is one of the key components of fusion-driven subcritical system (FDS-I). The probabilistic safety assessment (PSA) can provide valuable information on safety characteristics of FDS-I to give recommendations for the optimization of the blanket concepts and the improvement of the design. Event tree method has been adopted to probabilistically analyze the safety of the DWT blanket for FDS-I using the home-developed PSA code RiskA. The blanket melting frequency has been calculated and compared with the core melting frequencies of PWRs and a fast reactor. Sensitivity analysis of the safety systems has been performed. The results show that the current preliminary design of the FDS-I is very attractive in safety

  9. Probabilistic methods applied to the safety of nuclear power plant: annual report - 1980. Part. 1: theoretical fundaments

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Hesles, J.B.S.; Milidiu, R.L.; Maciel, C.C.; Gibelli, S.M.O.; Oliveira, L.C.; Fleming, P.V.; Rivera, R.R.J.

    1981-02-01

    The probabilistic Safety Analysis Group from COPPE was founded in 1980. This first part of the report shows the theoretical fundaments used for reliability analysis of some safety systems for Angra-1 [pt

  10. VERIFICATION OF THE FOOD SAFETY MANAGEMENT SYSTEM IN DEEP FROZEN FOOD PRODUCTION PLANT

    Directory of Open Access Journals (Sweden)

    Peter Zajác

    2010-07-01

    Full Text Available In work is presented verification of food safety management system of deep frozen food. Main emphasis is on creating set of verification questions within articles of standard STN EN ISO 22000:2006 and on searching of effectiveness in food safety management system. Information were acquired from scientific literature sources and they pointed out importance of implementation and upkeep of effective food safety management system. doi:10.5219/28

  11. Survey of probabilistic methods in safety and risk assessment for nuclear power plant licensing

    International Nuclear Information System (INIS)

    1984-04-01

    After an overview about the goals and general methods of probabilistic approaches in nuclear safety the main features of probabilistic safety or risk assessment (PRA) methods are discussed. Mostly in practical applications not a full-fledged PRA is applied but rather various levels of analysis leading from unavailability assessment of systems over the more complex analysis of the probable core damage stages up to the assessment of the overall health effects on the total population from a certain practice. The various types of application are discussed in relation to their limitation and benefits for different stages of design or operation of nuclear power plants. This gives guidance for licensing staff to judge the usefulness of the various methods for their licensing decisions. Examples of the application of probabilistic methods in several countries are given. Two appendices on reliability analysis and on containment and consequence analysis provide some more details on these subjects. (author)

  12. Use of the t-distribution to construct seismic hazard curves for seismic probabilistic safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Yee, Eric [KEPCO International Nuclear Graduate School, Dept. of Nuclear Power Plant Engineering, Ulsan (Korea, Republic of)

    2017-03-15

    Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered.

  13. Use of the t-distribution to construct seismic hazard curves for seismic probabilistic safety assessments

    International Nuclear Information System (INIS)

    Yee, Eric

    2017-01-01

    Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered

  14. Probabilistic safety assessment for high-level waste tanks at Hanford

    International Nuclear Information System (INIS)

    Sullivan, L.H.; MacFarlane, D.R.; Stack, D.W.

    1996-01-01

    Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA), including consideration of external events, for the 18 tank farms at the Hanford Tank Farm (HTF). This work was sponsored by the Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM)

  15. Present and future of probabilistic safety analysis of Juragua Nuclear Power Plant

    International Nuclear Information System (INIS)

    Salomon, J.; Rivero, J.J.

    1993-01-01

    This work present the main conditions of probabilistic safety analysis of Juragua Nuclear Power Plant, which includes the following aspects: Staff preparedness; Creation of ANCON code; Analysis activity; IAEA technical assistance project. The present situation of PSA National Project and its perspectives development are reported

  16. A review of the report ''IAEA safety targets and probabilistic risk assessment'' prepared for Greenpeace International

    International Nuclear Information System (INIS)

    1991-01-01

    At the request of the Director General, INSAG reviewed the report ''IAEA Safety Targets and Probabilistic Risk Assessment'' prepared for Greenpeace International by the Gesellschaft fuer Oekologische Forschung und Beratung mbH, Hannover, Germany. The conclusions of the report as well as the review results of INSAG experts are reproduced in this document

  17. Comparison of plant-specific probabilistic safety assessments and lessons learned

    International Nuclear Information System (INIS)

    Balfanz, H.P.; Berg, H.P.; Steininger, U.

    2001-01-01

    Probabilistic safety assessments (PSA) have been performed for all German nuclear power plants in operation. These assessments are mainly based on the recent German PSA guide and an earlier draft, respectively. However, comparison of these PSA show differences in the results which are discussed in this paper. Lessons learned from this comparison and further development of the PSA methodology are described. (orig.) [de

  18. Initiating events in the safety probabilistic analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Stasiulevicius, R.

    1989-01-01

    The importance of the initiating event in the probabilistic safety analysis of nuclear power plants are discussed and the basic procedures necessary for preparing reports, quantification and grouping of the events are described. The examples of initiating events with its occurence medium frequency, included those calculated for OCONEE reactor and Angra-1 reactor are presented. (E.G.)

  19. Insights from the Probabilistic Safety Assessment Application to Subsurface Operations at the Preclosure Facilities

    International Nuclear Information System (INIS)

    Hwang, Mee Jeong; Jung, Jong Tae

    2009-01-01

    In this paper, we present the insights obtained through the PSA (Probabilistic Safety Assessment) application to subsurface operation at the preclosure facilities of the repository. At present, medium-low level waste repository has been constructed in Korea, and studies for disposal of high level wastes are under way. Also, safety analysis for repository operation has been performed. Thus, we performed a probabilistic safety analysis for surface operation at the preclosure facilities with PSA methodology for a nuclear power plant. Since we don't have a code to analyze the waste repository safety analysis, we used the codes, AIMS (Advanced Information Management System for PSA) and FTREX (Fault Tree Reliability Evaluation eXpert) which are developed for a nuclear power plant's PSA to develop ET (Event Tree) and FT (Fault Tree), and to quantify for an example analysis

  20. Determination of the number of software tests using probabilistic safety assessment

    International Nuclear Information System (INIS)

    Kang, H. K.; Seong, T. Y.; Lee, K. Y.

    2000-01-01

    The broader usage of digital equipment in nuclear power plants gives rise to the safety problems of software. The field test should be performed before the software is used in critical applications because it is well known that software shows non-linear response when it is applied to different target systems in different environment. In the case of safety-critical applications, the result of tests contains usually zero failure case and the satisfiable number of tests is hard to be determined. In this paper, we suggests the method to determine the number of software tests without failure using the probabilistic safety assessment. From the result of the probabilistic safety assessment on total system, the desirable unavailability of software is calculated and the number of tests is determined

  1. Probabilistic safety analysis of radiation treatments with linear accelerator (Spanish Ed.)

    International Nuclear Information System (INIS)

    2012-02-01

    This publication addresses the issue of accidental exposures of radiotherapy patients and how to avoid them. More proactive approaches are required to anticipate and thus avoid situations that could lead to accidental exposures. In this context, the International Atomic Energy Agency (IAEA) and the Ibero American Forum of Radiation and Nuclear and Safety Regulatory Agencies (the FORO) have applied proactive methods, such as probabilistic safety assessment to radiotherapy treatments with accelerators. The methodology and results of this exercise are described in this publication.

  2. Automatic Verification of Timing Constraints for Safety Critical Space Systems

    Science.gov (United States)

    Fernandez, Javier; Parra, Pablo; Sanchez Prieto, Sebastian; Polo, Oscar; Bernat, Guillem

    2015-09-01

    In this paper is presented an automatic process of verification. We focus in the verification of scheduling analysis parameter. This proposal is part of process based on Model Driven Engineering to automate a Verification and Validation process of the software on board of satellites. This process is implemented in a software control unit of the energy particle detector which is payload of Solar Orbiter mission. From the design model is generated a scheduling analysis model and its verification model. The verification as defined as constraints in way of Finite Timed Automatas. When the system is deployed on target the verification evidence is extracted as instrumented points. The constraints are fed with the evidence, if any of the constraints is not satisfied for the on target evidence the scheduling analysis is not valid.

  3. Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

    International Nuclear Information System (INIS)

    Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

    1996-01-01

    The rules that are currently under application to verify the acceptance of flaws in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation to reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R6 procedure for assessing the integrity of the structure. (authors). 5 refs., 5 figs

  4. Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

    International Nuclear Information System (INIS)

    Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

    1995-01-01

    The current rules applied to verify the flaws acceptance in nuclear components rely on deterministic criteria supposed to ensure the plant safe operation. The interest in have a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation do reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R 6 procedure for assessing the structure integrity. (author). 5 refs., 5 figs., 1 tab

  5. Projected Impact of Compositional Verification on Current and Future Aviation Safety Risk

    Science.gov (United States)

    Reveley, Mary S.; Withrow, Colleen A.; Leone, Karen M.; Jones, Sharon M.

    2014-01-01

    The projected impact of compositional verification research conducted by the National Aeronautic and Space Administration System-Wide Safety and Assurance Technologies on aviation safety risk was assessed. Software and compositional verification was described. Traditional verification techniques have two major problems: testing at the prototype stage where error discovery can be quite costly and the inability to test for all potential interactions leaving some errors undetected until used by the end user. Increasingly complex and nondeterministic aviation systems are becoming too large for these tools to check and verify. Compositional verification is a "divide and conquer" solution to addressing increasingly larger and more complex systems. A review of compositional verification research being conducted by academia, industry, and Government agencies is provided. Forty-four aviation safety risks in the Biennial NextGen Safety Issues Survey were identified that could be impacted by compositional verification and grouped into five categories: automation design; system complexity; software, flight control, or equipment failure or malfunction; new technology or operations; and verification and validation. One capability, 1 research action, 5 operational improvements, and 13 enablers within the Federal Aviation Administration Joint Planning and Development Office Integrated Work Plan that could be addressed by compositional verification were identified.

  6. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

  7. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    1990-04-01

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  8. Report on probabilistic safety assessment (PSA) quality assurance in utilization of risk information

    International Nuclear Information System (INIS)

    2006-12-01

    Recently in Japan, introduction of nuclear safety regulations using risk information such as probabilistic safety assessment (PSA) has been considered and utilization of risk information in the rational and practical measures on safety assurance has made a progress to start with the operation or inspection area. The report compiled results of investigation and studies of PSA quality assurance in risk-informed activities in the USA. Relevant regulatory guide and standard review plan as well as issues and recommendations were reviewed for technical adequacy and advancement of probabilistic risk assessment technology in risk-informed decision making. Useful and important information to be referred as issues in PSA quality assurance was identified. (T. Tanaka)

  9. Current regulatory developments concerning the implementation of probabilistic safety analyses for external hazards in Germany

    International Nuclear Information System (INIS)

    Krauss, Matias; Berg, Heinz-Peter

    2014-01-01

    The Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) initiated in September 2003 a comprehensive program for the revision of the national nuclear safety regulations which has been successfully completed in November 2012. These nuclear regulations take into account the current recommendations of the International Atomic Energy Agency (IAEA) and Western European Nuclear Regulators Association (WENRA). In this context, the recommendations and guidelines of the Nuclear Safety Standards Commission (KTA) and the technical documents elaborated by the respective expert group on Probabilistic Safety Analysis for Nuclear Power Plants (FAK PSA) are being updated or in the final process of completion. A main topic of the revision was the issue external hazards. As part of this process and in the light of the accident at Fukushima and the findings of the related actions resulting in safety reviews of nuclear power plants at national level in Germany and on European level, a revision of all relevant standards and documents has been made, especially the recommendations of KTA and FAK PSA. In that context, not only design issues with respect to events such as earthquakes and floods have been discussed, but also methodological issues regarding the implementation of improved probabilistic safety analyses on this topic. As a result of the revision of the KTA 2201 series 'Design of Nuclear Power Plants against Seismic Events' with their parts 1 to 6, part 1 'Principles' was published as the first standard in November 2011, followed by the revised versions of KTA 2201.2 (soil) and 2201.4 (systems and components) in 2012. The modified the standard KTA 2201.3 (structures) is expected to be issued before the end of 2013. In case of part 5 (seismic instrumentation) and part 6 (post>seismic actions) draft amendments are expected in 2013. The expert group 'Probabilistic Safety Assessments for Nuclear Power Plants' (FAK PSA) is an advisory body of the Federal

  10. Probabilistic safety analysis of earth retaining structures during earthquakes

    Science.gov (United States)

    Grivas, D. A.; Souflis, C.

    1982-07-01

    A procedure is presented for determining the probability of failure of Earth retaining structures under static or seismic conditions. Four possible modes of failure (overturning, base sliding, bearing capacity, and overall sliding) are examined and their combined effect is evaluated with the aid of combinatorial analysis. The probability of failure is shown to be a more adequate measure of safety than the customary factor of safety. As Earth retaining structures may fail in four distinct modes, a system analysis can provide a single estimate for the possibility of failure. A Bayesian formulation of the safety retaining walls is found to provide an improved measure for the predicted probability of failure under seismic loading. The presented Bayesian analysis can account for the damage incurred to a retaining wall during an earthquake to provide an improved estimate for its probability of failure during future seismic events.

  11. Probabilistic safety evaluation: Development of procedures with applications on components used in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Dillstroem, P. [Det Norske Veritas AB, Stockholm (Sweden)

    2000-12-01

    A probabilistic procedure has been developed by SAQ Kontroll AB to calculate two different failure probabilities, P{sub F}: Probability of failure, defect size given by NDT/NDE. Probability of failure, defect not detected by NDT/NDE. Based on the procedure, SAQ Kontroll AB has developed a computer program PROPSE (PRObabilistic Program for Safety Evaluation). Within PROPSE, the following features are implemented: Two different algorithms to calculate the probability of failure are included: Simple Monte Carlo Simulation (MCS), with an error estimate on P{sub F}. First-Order Reliability Method (FORM), with sensitivity factors using the most probable point of failure in a standard normal space. Using these factors, it is possible to rank the parameters within an analysis. Estimation of partial safety factors, given an input target failure probability and characteristic values for fracture toughness, yield strength, tensile strength and defect depth. Extensive validation has been carried out, using the probabilistic computer program STAR6 from Nuclear Electric and the deterministic program SACC from SAQ Kontroll AB. The validation showed that the results from PROPSE were correct, and that the algorithms used in STAR6 were not intended to work for a general problem, when the standard deviation is either 'small' or 'large'. Distributions, to be used in a probabilistic analysis, are discussed. Examples on data to be used are also given.

  12. Probabilistic safety evaluation: Development of procedures with applications on components used in nuclear power plants

    International Nuclear Information System (INIS)

    Dillstroem, P.

    2000-12-01

    A probabilistic procedure has been developed by SAQ Kontroll AB to calculate two different failure probabilities, P F : Probability of failure, defect size given by NDT/NDE. Probability of failure, defect not detected by NDT/NDE. Based on the procedure, SAQ Kontroll AB has developed a computer program PROPSE (PRObabilistic Program for Safety Evaluation). Within PROPSE, the following features are implemented: Two different algorithms to calculate the probability of failure are included: Simple Monte Carlo Simulation (MCS), with an error estimate on P F . First-Order Reliability Method (FORM), with sensitivity factors using the most probable point of failure in a standard normal space. Using these factors, it is possible to rank the parameters within an analysis. Estimation of partial safety factors, given an input target failure probability and characteristic values for fracture toughness, yield strength, tensile strength and defect depth. Extensive validation has been carried out, using the probabilistic computer program STAR6 from Nuclear Electric and the deterministic program SACC from SAQ Kontroll AB. The validation showed that the results from PROPSE were correct, and that the algorithms used in STAR6 were not intended to work for a general problem, when the standard deviation is either 'small' or 'large'. Distributions, to be used in a probabilistic analysis, are discussed. Examples on data to be used are also given

  13. Utilization of probabilistic methods for evaluating the safety of PWRs built in France

    International Nuclear Information System (INIS)

    Queniart, D.; Brisbois, J.; Lanore, J.M.

    1985-01-01

    Firstly, it is recalled that, in France, PWRs are designed on a deterministic basis by studying the consequences of a limited number of conventional incidents whose estimated frequency is specified in order-of-magnitude terms and for which it is shown that the consequences, for each category of frequency, predominate over those of the other situations in the same category. These situations are called dimensioning situations. The paper then describes the use made of probabilistic methods. External attacks and loss of redundant systems are examined in particular. A probabilistic approach is in fact well suited to the evaluation of risks due, among other things, to aircraft crashes and the industrial environment. Analysis of the reliability of redundant systems has shown that, in the light of the overall risk assessment objective, their loss should be examined with a view to instituting counteraction to reduce the risks associated with such loss (particularly the introduction of special control procedures). Probabilistic methods are used to evaluate the effectiveness of the counteraction proposed and such a study has been carried out for total loss of electric power supply. Finally, the probabilistic study of hazard initiated post factum by the French safety authorities for the standardized 900 MW(e) power units is described. The study, which is not yet complete, will serve as the basis for a permanent safety analysis tool taking into account control procedures and the total operating experience acquired using these power units. (author)

  14. Review of probabilistic safety assessments by regulatory bodies

    International Nuclear Information System (INIS)

    2002-01-01

    This report provides guidance to assist regulatory bodies in carrying out reviews of the PSAs produced by utilities. In following this guidance, it is important that the regulatory body is able to satisfy itself that the PSA has been carried out to an acceptable standard and that it can be used for its intended applications. The review process becomes an important phase in determining the acceptability of the PSA since this provides a degree of assurance of the PSA scope, validity and limitations, as well as a better understanding of plants themselves. This report is also intended to assist technical experts managing or performing PSA reviews. A particular aim is to promote a standardized framework, terminology and form of documentation for the results of PSA reviews. The information presented in this report supports IAEA Safety Guide No. GS-G-1.2. Recommendations on the scope and methods to be used by the utility in the preparation of a PSA study is provided in IAEA Safety Guide No. NSG- 1.2. Information on these Safety Guides and other IAEA safety standards for nuclear power plants can be found on the following Internet site: http://www.iaea.org/ns/coordinet. The scope of this report covers the review of Level 1, 2 and 3 PSAs for event sequences occurring in all modes of plant operation (including full power, low power and shutdown). Where the scope of the analysis is narrower than this, a subset of the guidance can be identified and used. Information is provided on carrying out the review of a PSA throughout the PSA production process, i.e. from the initial decision to carry out the PSA through to the completion of the study and the production of the final PSA report. However, the same procedure can be applied to a completed PSA or to one already in progress. As a result of the performance of a PSA, changes to the design or operation of the plant are often identified that would increase the level of safety. This might include the addition of further safety

  15. 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)

  16. Probabilistic methods in the field of reactor safety in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Birkhofer, A [Technische Univ. Muenchen (Germany, F.R.). Lehrstuhl fuer Reaktordynamik und Reaktorsicherheit

    1979-01-01

    The present status and future prospects in Germany of reliability, as well as risk analysis, in the field of reactor safety are examined. The development of analytical methods with respect to the available data base is reviewed with consideration of the roles of reliability codes, component data, common mode failures, human influence, structural analysis and process computers. Some examples of the application of probability assessments are discussed and the extension of reliability analysis beyond the loss-of-coolant accident is considered. In the case of risk analysis, the object is to determine not only the probability of failure of systems but also the probability and extent of possible consequences. Some risk studies under investigation in Germany and the methodology of risk analysis are discussed. Reliability and risk analysis are involved to an increasing extent in safety research and licensing procedures and their influence in other fields such as the public perception of risk is also discussed.

  17. Verification of the safety communication protocol in train control system using colored Petri net

    International Nuclear Information System (INIS)

    Chen Lijie; Tang Tao; Zhao Xianqiong; Schnieder, Eckehard

    2012-01-01

    This paper deals with formal and simulation-based verification of the safety communication protocol in ETCS (European Train Control System). The safety communication protocol controls the establishment of safety connection between train and trackside. Because of its graphical user interface and modeling flexibility upon the changes in the system conditions, this paper proposes a composition Colored Petri Net (CPN) representation for both the logic and the timed model. The logic of the protocol is proved to be safe by means of state space analysis: the dead markings are correct; there are no dead transitions; being fair. Further analysis results have been obtained using formal and simulation-based verification approach. The timed models for the open transmit system and the application process are created for the purpose of performance analysis of the safety communication protocol. The models describe the procedure of data transmission and processing, and also provide relevant timed and stochastic factors, as well as time delay and lost packet, which may influence the time for establishment of safety connection of the protocol. Time for establishment of safety connection of the protocol in normal state is verified by formal verification, and then time for establishment of safety connection with different probability of lost packet is simulated. After verification it is found that the time for establishment of safety connection of the safety communication protocol satisfies the safety requirements.

  18. Ageing management by probabilistic safety assessment (PSA) methods

    International Nuclear Information System (INIS)

    Das, M.; Bhawal, R.N.; Maiti, S.C.

    1994-01-01

    The process and safety system of a nuclear power plant must achieve the reliability/availability target throughout the plant life or for extended plant life. It is therefore necessary to assess the trend of component or system ageing and to take preventive measures so that ageing effect can be counter balanced. In this paper a mathematical model has been established to predict ageing effect and to find out time dependent inspection or test interval to upgrade the system availability. (author). 5 figs

  19. Probabilistic Safety Goals for Nuclear Power Plants; Phases 2-4 / Final Report

    International Nuclear Information System (INIS)

    Bengtsson, Lisa; Knochenhauer, Michael; Holmberg, Jan-Erik; Rossi, Jukka

    2011-05-01

    The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Safety goals are defined in different ways in different countries and also used differently. Many countries are presently developing them in connection to the transfer to risk-informed regulation of both operating nuclear power plants (NPP) and new designs. However, it is far from self-evident how probabilistic safety criteria should be defined and used. On one hand, experience indicates that safety goals are valuable tools for the interpretation of results from a probabilistic safety assessment (PSA), and they tend to enhance the realism of a risk assessment. On the other hand, strict use of probabilistic criteria is usually avoided. A major problem is the large number of different uncertainties in a PSA model, which makes it difficult to demonstrate the compliance with a probabilistic criterion. Further, it has been seen that PSA results can change a lot over time due to scope extensions, revised operating experience data, method development, changes in system requirements, or increases of level of detail, mostly leading to an increase of the frequency of the calculated risk. This can cause a problem of consistency in the judgments. The first phase of the project (2006) provided a general description of the issue of probabilistic safety goals for nuclear power plants, of important concepts related to the definition and application of safety goals, and of experiences in Finland and Sweden. The second, third and fourth phases (2007-2009) have been concerned with providing guidance related to the resolution of some of the problems

  20. A probabilistic safety analysis of UF{sub 6} handling at the Portsmouth Gaseous Diffusion Plant

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, G.J.; Lewis, S.R.; Summitt, R.L. [Safety and Reliability Optimization Services (SAROS), Inc., Knoxville, TN (United States)

    1991-12-31

    A probabilistic safety study of UF{sub 6} handling activities at the Portsmouth Gaseous Diffusion Plant has recently been completed. The analysis provides a unique perspective on the safety of UF{sub 6} handling activities. The estimated release frequencies provide an understanding of current risks, and the examination of individual contributors yields a ranking of important plant features and operations. Aside from the probabilistic results, however, there is an even more important benefit derived from a systematic modeling of all operations. The integrated approach employed in the analysis allows the interrelationships among the equipment and the required operations to be explored in depth. This paper summarizes the methods used in the study and provides an overview of some of the technical insights that were obtained. Specific areas of possible improvement in operations are described.

  1. Simplified application of probabilistic safety analysis in nuclear power plants by means of artificial neural networks

    International Nuclear Information System (INIS)

    Oehmgen, T.; Knorr, J.

    2004-01-01

    Probabilistic safety analyses (PSA) are conducted to assess the balanced nature of plant design in terms of technical safety and the administrative management of plant operation in nuclear power plants. In the evaluation shown in this article of the operating experience accumulated in two nuclear power plants, all failures are traced back consistently to the plant media and component levels, respectively, for the calculation of reliability coefficients. Moreover, the use of neural networks for probabilistic calculations is examined. The results are verified on the basis of test examples. Calculations with neural networks are very easy to carry out in a kind of 'black box'. There is a possibility, for instance, to use the system in plant maintenance. (orig.) [de

  2. External flood probabilistic safety analysis of a coastal NPP

    International Nuclear Information System (INIS)

    Pisharady, Ajai S.; Chakraborty, M.K.; Acharya, Sourav; Roshan, A.D.; Bishnoi, L.R.

    2015-01-01

    External events pose a definitive challenge to safety of NPP, solely due to their ability to induce common cause failures. Flooding incidents at Le Blayais NPP, France, Fort Calhoun NPP, USA and Fukushima Daiichi have pointed to the importance of external flooding as an important contributor to NPP risk. A methodology developed for external flood PSA of a coastal NPP vulnerable to flooding due to tsunami, cyclonic storm and intense local precipitation is presented in this paper. Different tasks for EFPSA has been identified along with general approach for completing each task

  3. Uncertainty and sensitivity analysis methodology in a level-I PSA (Probabilistic Safety Assessment)

    International Nuclear Information System (INIS)

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

    1997-01-01

    This work presents a methodology for sensitivity and uncertainty analysis, applicable to a probabilistic safety assessment level I. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and system response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well different graphical visualization for the control of the study. (author) [es

  4. Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

    2009-01-01

    After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

  5. Uncertainty and sensitivity analysis in a Probabilistic Safety Analysis level-1

    International Nuclear Information System (INIS)

    Nunez Mc Leod, Jorge E.; Rivera, Selva S.

    1996-01-01

    A methodology for sensitivity and uncertainty analysis, applicable to a Probabilistic Safety Assessment Level I has been presented. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and systems response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well as different graphical visualization for the control of the study. (author)

  6. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 3: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The third volume of the Probabilistic Safety Assessment contains supporting information for the PSA as follows: Appendix C (continued) with details of the system analysis and reports for the system/top event models; Appendix D with results of the specific engineering analyses of internal initiating events; Appendix E, containing supporting data for the human performance assessment,; Appendix F with details of the estimation of the frequency of leaks at HIFAR and Appendix G, containing event sequence model and quantification results

  7. Probabilistic safety analysis procedures guide, Sections 8-12. Volume 2, Rev. 1

    International Nuclear Information System (INIS)

    McCann, M.; Reed, J.; Ruger, C.; Shiu, K.; Teichmann, T.; Unione, A.; Youngblood, R.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. The first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. This second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  8. Wind Power in Mexico: Simulation of a Wind Farm and Application of Probabilistic Safety Analysis

    OpenAIRE

    C. Martín del Campo–Márquez; P.F. Nelson–Edelstein; M.Á. García–Vázquez

    2009-01-01

    The most important aspects of wind energy in Mexico, including the potential for generating electricity and the major projects planned are presented here. Inparticular, the generation costs are compared to those of other energy sources. The results from the simulation of a 100 MWwind farm in the Tehuantepec Isthmus are also presented. In addition, the environmental impacts related to the wind farm in the mentioned zone are analyzed. Finally, some benefits of using Probabilistic Safety Analysi...

  9. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  10. Probabilistic safety analysis procedures guide. Sections 1-7 and appendices. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    Bari, R.A.; Buslik, A.J.; Cho, N.Z.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. This first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. The second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  11. Suggestions for an improved HRA method for use in Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Parry, Gareth W.

    1995-01-01

    This paper discusses why an improved Human Reliability Analysis (HRA) approach for use in Probabilistic Safety Assessments (PSAs) is needed, and proposes a set of requirements on the improved HRA method. The constraints imposed by the need to embed the approach into the PSA methodology are discussed. One approach to laying the foundation for an improved method, using models from the cognitive psychology and behavioral science disciplines, is outlined

  12. Framework for applying probabilistic safety analysis in nuclear regulation

    International Nuclear Information System (INIS)

    Dimitrijevic, V.B.

    1997-01-01

    The traditional regulatory framework has served well to assure the protection of public health and safety. It has been recognized, however, that in a few circumstances, this deterministic framework has lead to an extensive expenditure on matters hat have little to do with the safe and reliable operation of the plant. Developments of plant-specific PSA have offered a new and powerful analytical tool in the evaluation of the safety of the plant. Using PSA insights as an aid to decision making in the regulatory process is now known as 'risk-based' or 'risk-informed' regulation. Numerous activities in the U.S. nuclear industry are focusing on applying this new approach to modify regulatory requirements. In addition, other approaches to regulations are in the developmental phase and are being evaluated. One is based on the performance monitoring and results and it is known as performance-based regulation. The other, called the blended approach, combines traditional deterministic principles with PSA insights and performance results. (author)

  13. PROBABILISTIC SAFETY ASSESSMENT OF OPERATIONAL ACCIDENTS AT THE WASTE ISOLATION PILOT PLANT

    Energy Technology Data Exchange (ETDEWEB)

    Rucker, D.F.

    2000-09-01

    This report presents a probabilistic safety assessment of radioactive doses as consequences from accident scenarios to complement the deterministic assessment presented in the Waste Isolation Pilot Plant (WIPP) Safety Analysis Report (SAR). The International Council of Radiation Protection (ICRP) recommends both assessments be conducted to ensure that ''an adequate level of safety has been achieved and that no major contributors to risk are overlooked'' (ICRP 1993). To that end, the probabilistic assessment for the WIPP accident scenarios addresses the wide range of assumptions, e.g. the range of values representing the radioactive source of an accident, that could possibly have been overlooked by the SAR. Routine releases of radionuclides from the WIPP repository to the environment during the waste emplacement operations are expected to be essentially zero. In contrast, potential accidental releases from postulated accident scenarios during waste handling and emplacement could be substantial, which necessitates the need for radiological air monitoring and confinement barriers (DOE 1999). The WIPP Safety Analysis Report (SAR) calculated doses from accidental releases to the on-site (at 100 m from the source) and off-site (at the Exclusive Use Boundary and Site Boundary) public by a deterministic approach. This approach, as demonstrated in the SAR, uses single-point values of key parameters to assess the 50-year, whole-body committed effective dose equivalent (CEDE). The basic assumptions used in the SAR to formulate the CEDE are retained for this report's probabilistic assessment. However, for the probabilistic assessment, single-point parameter values were replaced with probability density functions (PDF) and were sampled over an expected range. Monte Carlo simulations were run, in which 10,000 iterations were performed by randomly selecting one value for each parameter and calculating the dose. Statistical information was then derived

  14. PROBABILISTIC SAFETY ASSESSMENT OF OPERATIONAL ACCIDENTS AT THE WASTE ISOLATION PILOT PLANT

    International Nuclear Information System (INIS)

    Rucker, D.F.

    2000-01-01

    This report presents a probabilistic safety assessment of radioactive doses as consequences from accident scenarios to complement the deterministic assessment presented in the Waste Isolation Pilot Plant (WIPP) Safety Analysis Report (SAR). The International Council of Radiation Protection (ICRP) recommends both assessments be conducted to ensure that ''an adequate level of safety has been achieved and that no major contributors to risk are overlooked'' (ICRP 1993). To that end, the probabilistic assessment for the WIPP accident scenarios addresses the wide range of assumptions, e.g. the range of values representing the radioactive source of an accident, that could possibly have been overlooked by the SAR. Routine releases of radionuclides from the WIPP repository to the environment during the waste emplacement operations are expected to be essentially zero. In contrast, potential accidental releases from postulated accident scenarios during waste handling and emplacement could be substantial, which necessitates the need for radiological air monitoring and confinement barriers (DOE 1999). The WIPP Safety Analysis Report (SAR) calculated doses from accidental releases to the on-site (at 100 m from the source) and off-site (at the Exclusive Use Boundary and Site Boundary) public by a deterministic approach. This approach, as demonstrated in the SAR, uses single-point values of key parameters to assess the 50-year, whole-body committed effective dose equivalent (CEDE). The basic assumptions used in the SAR to formulate the CEDE are retained for this report's probabilistic assessment. However, for the probabilistic assessment, single-point parameter values were replaced with probability density functions (PDF) and were sampled over an expected range. Monte Carlo simulations were run, in which 10,000 iterations were performed by randomly selecting one value for each parameter and calculating the dose. Statistical information was then derived from the 10,000 iteration

  15. Development of Nuclear Safety Culture evaluation method for an operation team based on the probabilistic approach

    International Nuclear Information System (INIS)

    Han, Sang Min; Lee, Seung Min; Yim, Ho Bin; Seong, Poong Hyun

    2018-01-01

    Highlights: •We proposed a Probabilistic Safety Culture Healthiness Evaluation Method. •Positive relationship between the ‘success’ states of NSC and performance was shown. •The state probability profile showed a unique ratio regardless of the scenarios. •Cutset analysis provided not only root causes but also the latent causes of failures. •Pro-SCHEMe was found to be applicable to Korea NPPs. -- Abstract: The aim of this study is to propose a new quantitative evaluation method for Nuclear Safety Culture (NSC) in Nuclear Power Plant (NPP) operation teams based on the probabilistic approach. Various NSC evaluation methods have been developed, and the Korea NPP utility company has conducted the NSC assessment according to international practice. However, most of methods are conducted by interviews, observations, and the self-assessment. Consequently, the results are often qualitative, subjective, and mainly dependent on evaluator’s judgement, so the assessment results can be interpreted from different perspectives. To resolve limitations of present evaluation methods, the concept of Safety Culture Healthiness was suggested to produce quantitative results and provide faster evaluation process. This paper presents Probabilistic Safety Culture Healthiness Evaluation Method (Pro-SCHEMe) to generate quantitative inputs for Human Reliability Assessment (HRA) in Probabilistic Safety Assessment (PSA). Evaluation items which correspond to a basic event in PSA are derived in the first part of the paper through the literature survey; mostly from nuclear-related organizations such as the International Atomic Energy Agency (IAEA), the United States Nuclear Regulatory Commission (U.S.NRC), and the Institute of Nuclear Power Operations (INPO). Event trees (ETs) and fault trees (FTs) are devised to apply evaluation items to PSA based on the relationships among such items. The Modeling Guidelines are also suggested to classify and calculate NSC characteristics of

  16. Probabilistic safety assessment; actions and priorities in the EC-frame

    International Nuclear Information System (INIS)

    Amendola, A.; Mancini, G.; Volta, G.

    1987-01-01

    An overview is given of PSA research activities at the JRC and through shared cost actions with national laboratories under the nuclear reactor safety and major hazards of industrial installations programmes. These activities are directed towards the development of methods for PSA, the validation methods and the setting up of appropriate data bases. PSA is also directly or indirectly an emerging theme for the coordination activities in the area of nuclear safety criteria and safety objectives. Finally probabilistic techniques being increasing by being used for safety and reliability in various industrial sectors the CEC supported the preparation and setting up of a European Safety and Reliability Association that carries different types of actions. (orig.)

  17. Evaluation of fire probabilistic safety assessment for a PWR plant

    International Nuclear Information System (INIS)

    Wu, C.H.; Lin, T.J.; Kao, T.M.

    2001-01-01

    The internal fire analysis of the level 1 power operation probability safety assessment (PSA) for Maanshan (PWR) Nuclear Power Plant (MNPP) was updated. The fire analysis adopted a scenario-based PSA approach to systematically evaluate fire and smoke hazards and their associated risk impact to MNPP. The result shows that the core damage frequency (CDF) due to fire is about six times lower than the previous one analyzed by the Atomic Energy Council (AEC), Republic of China in 1987. The plant model was modified to reflect the impact of human events and recovery actions during fire. Many tabulated EXCEL spread-sheets were used for evaluation of the fire risk. The fire-induced CDF for MNPP is found to be 2.1 E-6 per year in this study. The relative results of the fire analysis will provide the bases for further risk-informed fire protection evaluation in the near future. (author)

  18. Development of several data bases related to reactor safety research including probabilistic safety assessment and incident analysis at JAERI

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Oikawa, Tetsukuni; Watanabe, Norio; Izumi, Fumio; Higuchi, Suminori

    1986-01-01

    Presented are several databases developed at JAERI for reactor safety research including probabilistic safety assessment and incident analysis. First described are the recent developments of the databases such as 1) the component failure rate database, 2) the OECD/NEA/IRS information retrieval system, 3) the nuclear power plant database and so on. Then several issues are discussed referring mostly to the operation of the database (data input and transcoding) and to the retrieval and utilization of the information. Finally, emphasis is given to the increasing role which artifitial intelligence techniques such as natural language treatment and expert systems may play in improving the future capabilities of the databases. (author)

  19. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    2002-01-01

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  20. The dynamic flowgraph methodology as a safety analysis tool : programmable electronic system design and verification

    NARCIS (Netherlands)

    Houtermans, M.J.M.; Apostolakis, G.E.; Brombacher, A.C.; Karydas, D.M.

    2002-01-01

    The objective of this paper is to demonstrate the use of the Dynamic Flowgraph Methodology (DFM) during the design and verification of programmable electronic safety-related systems. The safety system consists of hardware as well as software. This paper explains and demonstrates the use of DFM, and

  1. Training to Support Standardization and Improvement of Safety I and C Related Verification and Validation Activities

    Energy Technology Data Exchange (ETDEWEB)

    Ammon, G.; Schoenfelder, C.

    2014-07-01

    In recent years AREVA has conducted several measures to enhance the effectiveness of safety I and C related verification and validation activities within nuclear power plant (NPP) new build as well as modernization projects, thereby further strengthening its commitment to achieving the highest level of safety in nuclear facilities. (Author)

  2. 78 FR 32010 - Pipeline Safety: Public Workshop on Integrity Verification Process

    Science.gov (United States)

    2013-05-28

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No... Hazardous Materials Safety Administration, DOT. ACTION: Notice of public meeting. SUMMARY: This notice is announcing a public workshop to be held on the concept of ``Integrity Verification Process.'' The Integrity...

  3. 77 FR 50723 - Verification, Validation, Reviews, and Audits for Digital Computer Software Used in Safety...

    Science.gov (United States)

    2012-08-22

    ... regulations with respect to software verification and auditing of digital computer software used in the safety... Standards and Records,'' which requires, in part, that a quality assurance program be established and implemented to provide adequate assurance that systems and components important to safety will satisfactorily...

  4. A combined deterministic and probabilistic procedure for safety assessment of components with cracks - Handbook.

    Energy Technology Data Exchange (ETDEWEB)

    Dillstroem, Peter; Bergman, Mats; Brickstad, Bjoern; Weilin Zang; Sattari-Far, Iradj; Andersson, Peder; Sund, Goeran; Dahlberg, Lars; Nilsson, Fred (Inspecta Technology AB, Stockholm (Sweden))

    2008-07-01

    SSM has supported research work for the further development of a previously developed procedure/handbook (SKI Report 99:49) for assessment of detected cracks and tolerance for defect analysis. During the operative use of the handbook it was identified needs to update the deterministic part of the procedure and to introduce a new probabilistic flaw evaluation procedure. Another identified need was a better description of the theoretical basis to the computer program. The principal aim of the project has been to update the deterministic part of the recently developed procedure and to introduce a new probabilistic flaw evaluation procedure. Other objectives of the project have been to validate the conservatism of the procedure, make the procedure well defined and easy to use and make the handbook that documents the procedure as complete as possible. The procedure/handbook and computer program ProSACC, Probabilistic Safety Assessment of Components with Cracks, has been extensively revised within this project. The major differences compared to the last revision are within the following areas: It is now possible to deal with a combination of deterministic and probabilistic data. It is possible to include J-controlled stable crack growth. The appendices on material data to be used for nuclear applications and on residual stresses are revised. A new deterministic safety evaluation system is included. The conservatism in the method for evaluation of the secondary stresses for ductile materials is reduced. A new geometry, a circular bar with a circumferential surface crack has been introduced. The results of this project will be of use to SSM in safety assessments of components with cracks and in assessments of the interval between the inspections of components in nuclear power plants

  5. Applications of Trajectory Solid Angle for Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Wong, Po Kee; Wong, Adam E.; Wong, Anita

    2002-01-01

    In 1974, a well-known research problem in Statistical Mechanics entitled 'To determine and define the probability function P.sub.2 of a particle hitting a predetermined area, given all its parameters of generation and ejection' was openly solicited for its solution from research and development organizations in U.S.A. One of many proposed solutions of the problem, initiated at that time, is by means of the Trajectory Solid Angle (TSA). TSA is defined as the integral of the dot product of the unit tangent of the particle's trajectory to the vector area divided by the square of the position vector connecting between the point of ejection and that of the surface to be hit. The invention provides: (1) The precise and the unique solution of a previously unsolved P.sub.2 problem: (2) Impacts to the governmental NRC safety standards and DOD weapon systems and many activities in the Department of Energy; (3) Impacts to update the contents of text books of physics and mathematics of all levels; (4) Impacts to the scientific instruments with applications in high technologies. The importance of Trajectory Solid Angle can be quoted from a letter by the late Institute Professor P. M. Morse of MIT who reviewed the DOE proposal P7900450 (reference No. 7) in 1979 and addressed to the inventor. 'If the Trajectory Solid Angle is correct it will provide a revolutionary concept in physics'. (authors)

  6. Probabilistic seismic safety study of an existing nuclear power plant

    International Nuclear Information System (INIS)

    Kennedy, R.P.; Cornell, C.A.; Kaplan, S.; Perla, H.F.

    1980-01-01

    This study was conducted as part of an overall safety study of the Oyster Creek nuclear power plant. The earthquake hazard was considered as an initiating event that could result in radioactive release from the site as a result of core melt. The probability of earthquake initiated releases were compared with the probability of releases due to other initiating events. Three steps are necessary to evaluate the probability of earthquake initiated core melt. (1) estimate the ground motion (peak ground acceleration) and uncertainty in this estimate as functions of annual probability of occurrence; (2) estimate the conditional probability of failure and its uncertainty for structures, equipment, piping, controls, etc., as functions of ground acceleration; and (3) combine these estimates to obtain probabilities of earthquake induced failure and uncertainties in such estimates to be used in event trees, system models, and fault trees for evaluating the probability of earthquake induced core melt. This paper concentrates on the first two steps with emphasis on step 2. The major difference between the work presented and previous papers is the development and use of uncertainty estimates for both the ground motion probability estimates and the conditional probability of failure estimates. (orig.)

  7. Probabilistic safety analysis for FRJ-2 motivation, methodology and results

    International Nuclear Information System (INIS)

    Wolters, J.

    1993-01-01

    A PSA of the Research Reactor FRJ-2 was performed to check the twenty-year-old safety system for weak points and to develop accident management as a 'fourth line of defence' against severe accidents according to a German initiative. The total core damage frequency proved to be 1.5·10 -4 /a, which is consistent with figures found for other research reactors. Minor plant modifications will reduce the value by roughly a factor of 4, resulting in a frequency of 3·10 -7 /a for a major release of fission products into the environment caused by an independent failure of the containment. The integrity of the gas-tight steel containment proved not to be endangered by any core damage accident. From the results and insights gained by the PSA, many accident management measures could be identified and defined for the emergency handbook. The most important measure is primary feed and bleed, for which the feed line already exists. (author)

  8. Probabilistic safety analysis for FRJ-2 motivation, methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Wolters, J [Institute for Safety Research and Reactor Technology, Research Center Juelich (Germany)

    1993-07-01

    A PSA of the Research Reactor FRJ-2 was performed to check the twenty-year-old safety system for weak points and to develop accident management as a 'fourth line of defence' against severe accidents according to a German initiative. The total core damage frequency proved to be 1.5{center_dot}10{sup -4}/a, which is consistent with figures found for other research reactors. Minor plant modifications will reduce the value by roughly a factor of 4, resulting in a frequency of 3{center_dot}10{sup -7}/a for a major release of fission products into the environment caused by an independent failure of the containment. The integrity of the gas-tight steel containment proved not to be endangered by any core damage accident. From the results and insights gained by the PSA, many accident management measures could be identified and defined for the emergency handbook. The most important measure is primary feed and bleed, for which the feed line already exists. (author)

  9. The implications of probabilistic risk assessment for safety policy

    International Nuclear Information System (INIS)

    Hayns, M.R.

    1987-01-01

    The use of PRA results in decision making requires a level of understanding on the part of the decision maker which is higher than that obtaining previously. The most important application of PRA lies not in the final results but in the intermediate results which refer to specific systems and operations. Such intermediate results are of great value either at the design stage or later during operation. One of the most 'visible' uses of PRA results is in comparing calculated plant risks with either proposed acceptability criteria, or with other plant, or even natural events. The capability to perform PRA has been established. Only the incorporation of PRA into the licensing process is lacking. The principal conclusions on the implications of PRA for safety policy are as follows: regardless of its state of development, PRA is the only means available for calculating public risk, being able to quantify risk is important in policy related to risk acceptability and to national energy policy. PRAs will be used to establish research and development priorities. Any hazardous plant can be treated using the same methods. More sophisticated methods will be used for solving engineering problems. (author)

  10. Probabilistic safety analysis for FRJ-2 motivation, methodology and results

    International Nuclear Information System (INIS)

    Wolters, J.

    1994-01-01

    A PSA of the Research Reactor FRJ-2 was performed to check the twenty-year-old safety system for weak points and to develop accident management as a 'fourth line of defence' against severe accidents according to a German initiative. The total core damage frequency proved to be 1.5·10 -4 /a, which is consistent with figures found for other research reactors. Minor plant modifications will reduce the value by roughly a factor of 4, resulting in a frequency of 3·10 -7 /a for a major release of fission products into the environment caused by an independent failure of the containment. The integrity of the gas-tight steel containment proved not to be endangered by any core damage accident. From the results and insights gained by the PSA, many accident management measures could be identified and defined for the emergency handbook. The most important measure is primary feed and bleed, for which the feed line already exists. (author)

  11. The Safety Assessment of OPR-1000 for Station Blackout Applying Combined Deterministic and Probabilistic Procedure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu; Ahn, Seung-Hoon; Cho, Dae-Hyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    This is termed station blackout (SBO). However, it does not generally include the loss of available AC power to safety buses fed by station batteries through inverters or by alternate AC sources. Historically, risk analysis results have indicated that SBO was a significant contributor to overall core damage frequency. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident, which is a typical beyond design basis accident and important contributor to overall plant risk, is performed by applying the combined deterministic and probabilistic procedure (CDPP). In addition, discussions are made for reevaluation of SBO risk at OPR-1000 by eliminating excessive conservatism in existing PSA. The safety assessment of OPR-1000 for SBO accident, which is a typical BDBA and significant contributor to overall plant risk, was performed by applying the combined deterministic and probabilistic procedure. However, the reference analysis showed that the CDF and CCDP did not meet the acceptable risk, and it was confirmed that the SBO risk should be reevaluated. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it was demonstrated that the proposed CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

  12. Method to Find Recovery Event Combinations in Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Jung, Woo Sik; Riley, Jeff

    2016-01-01

    These research activities may develop mathematical methods, engineering analyses, and business processes. The research activities of the project covered by this scope are directed toward the specific issues of implementing the methods and strategies on a computational platform, identifying the features and enhancements to EPRI tools that would be necessary to realize significant improvements to the risk assessments performed by the end user. Fault tree analysis is extensively and successfully applied to the risk assessment of safety-critical systems such as nuclear, chemical and aerospace systems. The fault tree analysis is being used together with an event tree analysis in PSA of nuclear power plants. Fault tree solvers for a PSA are mostly based on the cutset-based algorithm. They generate minimal cut sets (MCSs) from a fault tree. The most popular fault tree solver in the PSA industry is FTREX. During the course of this project, certain technical issues (see Sections 2 to 5) have been identified that need to be addressed regarding how minimal cut sets are generated and quantified. The objective of this scope of the work was to develop new methods or techniques to address these technical limitations. By turning on all the cutset initiators (%1, %2, %3, %), all the possible minimal cut sets can be calculated easier than with the original fault tree. It is accomplished by the fact that the number of events in the minimal cut sets are significantly reduced by using cutset initiators instead of random failure events. And byy turning on a few chosen cutset initiators and turning off the other cutset initiators, minimal cut sets of the selected cutset initiator(s) can be easily calculated. As explained in the previous Sections, there is no way to calculate these minimal cut sets by turning off/on the random failure events in the original fault tree

  13. Method to Find Recovery Event Combinations in Probabilistic Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Woo Sik [Sejong University, Seoul (Korea, Republic of); Riley, Jeff [Electric Power Research, Palo Alto (United States)

    2016-05-15

    These research activities may develop mathematical methods, engineering analyses, and business processes. The research activities of the project covered by this scope are directed toward the specific issues of implementing the methods and strategies on a computational platform, identifying the features and enhancements to EPRI tools that would be necessary to realize significant improvements to the risk assessments performed by the end user. Fault tree analysis is extensively and successfully applied to the risk assessment of safety-critical systems such as nuclear, chemical and aerospace systems. The fault tree analysis is being used together with an event tree analysis in PSA of nuclear power plants. Fault tree solvers for a PSA are mostly based on the cutset-based algorithm. They generate minimal cut sets (MCSs) from a fault tree. The most popular fault tree solver in the PSA industry is FTREX. During the course of this project, certain technical issues (see Sections 2 to 5) have been identified that need to be addressed regarding how minimal cut sets are generated and quantified. The objective of this scope of the work was to develop new methods or techniques to address these technical limitations. By turning on all the cutset initiators (%1, %2, %3, %), all the possible minimal cut sets can be calculated easier than with the original fault tree. It is accomplished by the fact that the number of events in the minimal cut sets are significantly reduced by using cutset initiators instead of random failure events. And byy turning on a few chosen cutset initiators and turning off the other cutset initiators, minimal cut sets of the selected cutset initiator(s) can be easily calculated. As explained in the previous Sections, there is no way to calculate these minimal cut sets by turning off/on the random failure events in the original fault tree.

  14. Comparison of a Traditional Probabilistic Risk Assessment Approach with Advanced Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis L; Mandelli, Diego; Zhegang Ma

    2014-11-01

    As part of the Light Water Sustainability Program (LWRS) [1], the purpose of the Risk Informed Safety Margin Characterization (RISMC) [2] Pathway research and development (R&D) is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario. We also describe our approach we are using to represent advanced flooding analysis.

  15. The significance of the probabilistic safety analysis (PSA) in administrative procedures under nuclear law

    International Nuclear Information System (INIS)

    Berg, H.P.

    1994-01-01

    The probabilistic safety analysis (PSA) is a useful tool for safety relevant evaluation of nuclear power plant designed on the basis of deterministic specifications. The PSA yields data identifying reliable or less reliable systems, or frequent or less frequent failure modes to be taken into account for safety engineering. Performance of a PSA in administrative procedures under nuclear law, e.g. licensing, is an obligation laid down in a footnote to criterion 1.1 of the BMI safety criteria catalogue, which has been in force unaltered since 1977. The paper explains the application and achievements of PSA in the phase of reactor development concerned with the conceptual design basis and design features, using as an example the novel PWR. (orig./HP) [de

  16. Application of probabilistic safety goals to regulation of nuclear power plants in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Rzentkowski, G.; Akl, Y.; Yalaoui, S. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2013-07-01

    In the Canadian nuclear regulatory framework, Safety Goals are formulated in addition to the deterministic design requirements and the dose acceptance criteria so that risk to the public that originates from accidents outside the design basis is considered. In principle, application of the Safety Goals ensures that the likelihood of accidents with serious radiological consequences is extremely low, and the potential radiological consequences from severe accidents are limited as far as practicable. Effectively, the Safety Goals extend the plant design envelope to include not only the capabilities of the plant to successfully cope with various plant states, but also practical measures to halt the progression of severe accidents. This paper describes the general approach to the development of the Safety Goals and their application to the existing nuclear power plants in Canada. This general approach is consistent with the currently accepted international practice and Canadian regulatory experience. The results of probabilistic safety assessments indicate that the Safety Goals meet or exceed international safety objectives due to effective implementation of the defence-in-depth principle in the reactor design and plant operation. At the same time, the application of the Safety Goals reveal that practicable measures exist to further enhance the overall level of reactor safety by focusing on severe accident prevention and mitigation. These measures are being currently implemented through refurbishment projects and feedback on operating experience. (author)

  17. Probabilistic safety goals. Phase 1 - Status and experiences in Sweden and Finland

    International Nuclear Information System (INIS)

    Holmberg, J.E.; Knochenhauer, M.

    2007-03-01

    The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: 1) The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. 2) Safety goals as informal or mandatory limits. 3) Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). 4) Relation between safety goals defined on different levels, e.g., for core damage

  18. Probabilistic Safety Goals. Phase 1 Status and Experiences in Sweden and Finland

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, Jan-Erik (VTT, FI-02044 VTT (Finland)); Knochenhauer, Michael (Relcon Scandpower AB, SE-172 25 Sundbyberg (Sweden))

    2007-02-15

    The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. Safety goals as informal or mandatory limits. Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). Relation between safety goals defined on different levels, e.g., for core damage and for

  19. Probabilistic Safety Goals. Phase 1 Status and Experiences in Sweden and Finland

    International Nuclear Information System (INIS)

    Holmberg, Jan-Erik; Knochenhauer, Michael

    2007-02-01

    The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. Safety goals as informal or mandatory limits. Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). Relation between safety goals defined on different levels, e.g., for core damage and for

  20. Probabilistic safety goals. Phase 1 - Status and experiences in Sweden and Finland

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J.E. [VTT (Finland); Knochenhauer, M. [Relcon Scandpower AB (Sweden)

    2007-03-15

    The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: 1) The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. 2) Safety goals as informal or mandatory limits. 3) Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). 4) Relation between safety goals defined on different levels, e.g., for core damage

  1. 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

  2. State of the art on the probabilistic safety assessment (P.S.A.)

    International Nuclear Information System (INIS)

    Devictor, N.; Bassi, A.; Saignes, P.; Bertrand, F.

    2008-01-01

    The use of Probabilistic Safety Assessment (PSA) is internationally increasing as a means of assessing and improving the safety of nuclear and non-nuclear facilities. To support the development of a competence on Probabilistic Safety Assessment, a set of states of the art regarding these tools and their use has been made between 2001 and 2005, in particular on the following topics: - Definition of the PSA of level 1, 2 and 3; - Use of PSA in support to design and operation of nuclear plants (risk-informed applications); - Applications to Non Reactor Nuclear Facilities. The report compiled in a single document these states of the art in order to ensure a broader use; this work has been done in the frame of the Project 'Reliability and Safety of Nuclear Facility' of the Nuclear Development and Innovation Division of the Nuclear Energy Division. As some of these states of the art have been made in support to exchanges with international partners and were written in English, a section of this document is written in English. This work is now applied concretely in support to the design of 4. Generation nuclear systems as Sodium-cooled Fast Reactors and especially Gas-cooled Fast Reactor, that have been the subject of communications during the conferences ANS (Annual Meeting 2007), PSA'08, ICCAP'08 and in the journal Science and Technology of Nuclear Installations. (authors)

  3. Simplified probabilistic approach to determine safety factors in deterministic flaw acceptance criteria

    International Nuclear Information System (INIS)

    Barthelet, B.; Ardillon, E.

    1997-01-01

    The flaw acceptance rules in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a reliable method of evaluating the safety margins and the integrity of components led Electricite de France to launch a study to link safety factors with requested reliability. A simplified analytical probabilistic approach is developed to analyse the failure risk in Fracture Mechanics. Assuming lognormal distributions of the main random variables, it is possible considering a simple Linear Elastic Fracture Mechanics model, to determine the failure probability as a function of mean values and logarithmic standard deviations. The 'design' failure point can be analytically calculated. Partial safety factors on the main variables (stress, crack size, material toughness) are obtained in relation with reliability target values. The approach is generalized to elastic plastic Fracture Mechanics (piping) by fitting J as a power law function of stress, crack size and yield strength. The simplified approach is validated by detailed probabilistic computations with PROBAN computer program. Assuming reasonable coefficients of variations (logarithmic standard deviations), the method helps to calibrate safety factors for different components taking into account reliability target values in normal, emergency and faulted conditions. Statistical data for the mechanical properties of the main basic materials complement the study. The work involves laboratory results and manufacture data. The results of this study are discussed within a working group of the French in service inspection code RSE-M. (authors)

  4. Formal verification and validation of the safety-critical software in a digital reactor protection system

    International Nuclear Information System (INIS)

    Kwon, K. C.; Park, G. Y.

    2006-01-01

    This paper describes the Verification and Validation (V and V) activities for the safety-critical software in a Digital Reactor Protection System (DRPS) that is being developed through the Korea nuclear instrumentation and control system project. The main activities of the DRPS V and V process are a preparation of the software planning documentation, a verification of the software according to the software life cycle, a software safety analysis and a software configuration management. The verification works for the Software Requirement Specification (SRS) of the DRPS consist of a technical evaluation, a licensing suitability evaluation, a inspection and traceability analysis, a formal verification, and preparing a test plan and procedure. Especially, the SRS is specified by the formal specification method in the development phase, and the formal SRS is verified by a formal verification method. Through these activities, we believe we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the nuclear safety-critical software in a DRPS. (authors)

  5. Utilization of a risk matrix based on Probabilistic Safety Analysis to improve nuclear safety in NPP

    International Nuclear Information System (INIS)

    Stubbe, Gerald

    2010-01-01

    The Probabilistic Safety Analysis (PSA) is a systematic and comprehensive methodology to evaluate risks associated with a complex engineered technological entity. Risk in a PSA is defined as a feasible detrimental outcome of an initiator. Those initiators can be 'classical' transient as the loss of main feedwater, loss of the secondary heat sink, etc.. or accident (LOCA - Loss Of Coolant Accident, SGTR - Steam Generator Tube Rupture, LOOP - Loss Of Offsite Power, etc..) In a PSA, risk is characterized by two quantities: the magnitude (severity) of the possible adverse consequence, the likelihood (probability) of occurrence of each consequence. Consequences are expressed numerically (for this purpose: the core damage) and their likelihoods of occurrence are expressed as probabilities or frequencies (i.e., the number of occurrences or the probability of occurrence per unit time). The total risk is the expected loss: the sum of the products of the consequences multiplied by their probabilities. This lead to the parameter CDF: The Core Damage Frequency, which is expressed by unit of time. The main advantage of this risk calculation is to have a global, integrated, overview of the plants and their systems. This allows to have an objective and quantitative point of view on the importance of the equipments, human action, or common cause failures that can challenge the plant's safety. A total PSA model is divided in three levels: Level one, which consider the core damage; Level two, which consider the robustness of the containment; Level three, which consider the impact of the radiological release on the public. For the purpose of the risk matrix, a level one PSA is needed. The scope of a PSA model is important to have a good characterization of the plant's risk. The matrix makes more sense if you have a full scope level one model, containing, furthermore the internal events, the fire and flooding, but also seismic event (if relevant). Asymmetries are also classical in the

  6. Performance Verification for Safety Injection Tank with Fluidic Device

    International Nuclear Information System (INIS)

    Yune, Seok Jeong; Kim, Da Yong

    2014-01-01

    In LBLOCA, the SITs of a conventional nuclear power plant deliver excessive cooling water to the reactor vessel causing the water to flow into the containment atmosphere. In an effort to make it more efficient, Fluidic Device (FD) is installed inside a SIT of Advanced Power Reactor 1400 (APR 1400). FD, a complete passive controller which doesn't require actuating power, controls injection flow rates which are susceptible to a change in the flow resistance inside a vortex chamber of FD. When SIT Emergency Core Cooling (ECC) water level is above the top of the stand pipe, the water enters the vortex chamber through both the top of the stand pipe and the control ports resulting in injection of the water at a large flow rate. When the water level drops below the top of the stand pipe, the water only enters the vortex chamber through the control ports resulting in vortex formation in the vortex chamber and a relatively small flow injection. Performance verification of SIT shall be carried out because SITs play an integral role to mitigate accidents. In this paper, the performance verification method of SIT with FD is presented. In this paper, the equations for calculation of flow resistance coefficient (K) are induced to evaluate on-site performance of APR 1400 SIT with FD. Then, the equations are applied to the performance verification of SIT with FD and good results are obtained

  7. Probabilistic safety analysis of DC power supply requirements for nuclear power plants. Technical report

    International Nuclear Information System (INIS)

    Baranowsky, P.W.; Kolaczkowski, A.M.; Fedele, M.A.

    1981-04-01

    A probabilistic safety assessment was performed as part of the Nuclear Regulatory Commission generic safety task A-30, Adequacy of Safety Related DC Power Supplies. Event and fault tree analysis techniques were used to determine the relative contribution of DC power related accident sequences to the total core damage probability due to shutdown cooling failures. It was found that a potentially large DC power contribution could be substantially reduced by augmenting the minimum design and operational requirements. Recommendations included (1) requiring DC power divisional independence, (2) improved test, maintenance, and surveillance, and (3) requiring core cooling capability be maintained following the loss of one DC power bus and a single failure in another system

  8. Safety probabilistic study of Almirante Alvaro Alberto nuclear power plant-Unit I

    International Nuclear Information System (INIS)

    Lederman, L.; Arrieta, L.A.I.; Fernandes Filho, T.L.; Gibelli, S.M.O.; Berthoud, J.S.; Ambros, P.C.; Soares, H.V.; Camargo, C.T.M.

    1985-04-01

    The phase A of probabilistic safety study of Angra I nuclear power plant is presented, to be used by CNEN and FURNAS Centrais Eletricas S.A. as standard model in operational and safety analysis. The methodology applied is a modernization of WASH 1400/2.11/ study. Angra I safety systems are described. The selection and qualification of initiating sequence accident events which can damage the reactor core are done. The accident scenes are developed using the method of event trees. The reactor in subcritical condition (pressure, fuel temperature within limits and controlled level of reactor vessel) is studied during 24 hours. The uncertainness in failure probabilities of systems and in the determination of sequence frequencies for core danification are evaluated. Total frequency of sequences which cause the fusion of reactor core are presented. (M.C.K.) [pt

  9. A study on the methodology of probabilistic safety assessment for KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kwon, Young Min; Lee, Yong Bum; Jeong, Hae Yong; Yang, Joon Eon; Ha, Kyu Suk; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    Existing Probabilistic Safety Assessment(PSA) is a method for Light Water Reactor or Pressurized Heavy Water Reactor. Because KALIMER is different from these reactor, the new methodology of PSA need to be developed. In this paper, the PSA of Power Reactor Inherently Safety Module(PRISM) is analyzed, and Initiating Event such as Experiential Assessment, Logical Assessment and Failure Mode Effect Analysis(FMEA) is reviewed. Also, Pipe Damage Frequency Method is suggested for KALIMER. And the Reliability Physical method of Passive System, which is a chief safety system of KALIMER, is reviewed and its applicability is investigated. Finally, for the Preliminary PSA of KALIMER, Intermediate Heat Transfer System is analyzed. 23 refs., 10 figs., 13 tabs. (Author)

  10. Probabilistic safety assessment of the radiotherapy treatment with a linear accelerator for medical use

    International Nuclear Information System (INIS)

    Vilaragut Llanes, Juan Jose; Ferro Fernandez, Ruben; Rodriguez MartI, Manuel; Ramirez, Maria Luisa; Perez Mulas, Arturo; Barrientos Montero, Marta; Ortiz Lopez, Pedro; Somoano, Fernando; Delgado RodrIguez, Jose Miguel; Papadopulos, Susana B.; Pereira Jr, Pedro Paulo; Lopez Morones, Ramon; Larrinaga Cortina, Eduardo; Rivero Oliva, Jose de Jesus; Alemanny, Jorge

    2010-01-01

    This paper presents the results of the Probabilistic Safety Assessment to the radiotherapy treatment with an Electron Linear Accelerator for Medical Use, which was conducted in the framework of the Iberian-American Forum of Radiological and Nuclear Regulatory Agencies. Potential accidental exposures during the treatment of patients, workers and members of the public were assessed, although the study was mainly focused on patients. The methodology of failure modes and effects analysis was used to define accident initiating events and methods of event tree and fault tree analysis to determine the accident sequences that may occur. After quantifying the frequency of occurrence of the accident sequences, an important analysis was carried out in order to determine the most significant events from the point of view of safety. The major contributors to risk were identified as well as the most appropriate safety recommendations to reduce it. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. Online probabilistic operational safety assessment of multi-mode engineering systems using Bayesian methods

    International Nuclear Information System (INIS)

    Lin, Yufei; Chen, Maoyin; Zhou, Donghua

    2013-01-01

    In the past decades, engineering systems become more and more complex, and generally work at different operational modes. Since incipient fault can lead to dangerous accidents, it is crucial to develop strategies for online operational safety assessment. However, the existing online assessment methods for multi-mode engineering systems commonly assume that samples are independent, which do not hold for practical cases. This paper proposes a probabilistic framework of online operational safety assessment of multi-mode engineering systems with sample dependency. To begin with, a Gaussian mixture model (GMM) is used to characterize multiple operating modes. Then, based on the definition of safety index (SI), the SI for one single mode is calculated. At last, the Bayesian method is presented to calculate the posterior probabilities belonging to each operating mode with sample dependency. The proposed assessment strategy is applied in two examples: one is the aircraft gas turbine, another is an industrial dryer. Both examples illustrate the efficiency of the proposed method

  13. Overview on the different applications of probabilistic safety assessment for nuclear power plants

    International Nuclear Information System (INIS)

    Berg, Heinz-Peter

    2009-01-01

    Worldwide it can be recognised that the use of probabilistic safety assessment (PSA) in regulatory as well as operational decision-making is state of the art and seen as a successful development. Therefore, in most cases the regulator encourages the performance of PSAs to provide information to complement and support the defence in depth philosophy as well as operational configuration decisions. The main application of the PSA is still as part of integrated safety reviews, in particular in the frame of comprehensive (periodic) safety reviews. Other more specific applications areas of PSA are, among others, design evaluation, event analysis with aid of PSA, evaluation of technical specifications; risk-informed in-service inspection, risk monitoring and accident management. The extent of these applications vary from country to country but has been increasing during the last years. (orig.)

  14. Modifications of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany based upon new version of Emergency Operating Procedures

    International Nuclear Information System (INIS)

    Aldorf, R.

    1997-01-01

    In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to reflect on Probabilistic Safety Assessment-1 basis on impact of Emergency Response Guidelines (as one particular event from the list of other modifications) on Plant Safety. Following highlights help to orient the reader in main general aspects, findings and issues of the work that currently continues on. Older results of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany have revealed that human behaviour during accident progression scenarios represent one of the most important aspects in plant safety. Current effort of Nuclear Power Plants Dukovany (Czech Republic) and Bohunice (Slovak Republic) is focussed on development of qualitatively new symptom-based Emergency Operating Procedures called Emergency Response Guidelines Supplier - Westinghouse Energy Systems Europe, Brussels works in cooperation with teams of specialist from both Nuclear Power Plants. In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to prove on Probabilistic Safety Assessment -1 basis an expected - positive impact of Emergency Response Guidelines on Plant Safety, Since this contract is currently still in progress, it is possible to release only preliminary conclusions and observations. Emergency Response Guidelines compare to original Emergency Operating Procedures substantially reduce uncertainty of general human behaviour during plant response to an accident process. It is possible to conclude that from the current scope Probabilistic Safety Assessment Dukovany point of view (until core damage), Emergency Response Guidelines represent adequately wide basis for mitigating any initiating event

  15. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.M.

    1991-08-01

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990

  16. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.M.

    1991-01-01

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990. (author)

  17. Probabilistic safety assessment for digital instrumentation and control systems in nuclear power plants - a review

    International Nuclear Information System (INIS)

    Lu, L.; Jiang, J.

    2003-01-01

    Deregulation in electricity market has created a great deal of challenges for nuclear power industries [1]. To stay competitive, Nuclear Power Plants (NPPs) will have to find ways to reduce their operational costs and to improve the plant safety. Instrumentation and Control (I and C) systems play an important role in this regard. Thus, new methodologies need to be developed to manage the operation of I and C systems more economically without jeopardizing the overall plant safety. Probabilistic Safety Assessment (PSA) technique is one of the promising methods to deal with such an issue, because PSA analyzes various system operational issues from a probabilistic sense, rather than a worst-case approach. However, there are several limitations when PSA is applied to I and C systems directly. A possible solution to this problem can be found by incorporating PSA with several other approaches. To better understand the issues involved, an attempt has been made in this paper to carry out a literature survey on this and related subject, particularly the effort will be made on: 1) the development of digital I and C systems in NPP, 2) PSA and its potential benefits and limitations, and 3) applications of PSA in various aspects of I and C systems including the resource allocation, the determination of surveillance testing strategies and the design of I and C systems. Finally, some solutions to overcome the aforementioned obstacles when applying PSA in I and C systems are also examined critically. (author)

  18. Light water reactor sequence timing: its significance to probabilistic safety assessment modeling

    International Nuclear Information System (INIS)

    Bley, D.C.; Buttemer, D.R.; Stetkar, J.W.

    1988-01-01

    This paper examines event sequence timing in light water reactor plants from the viewpoint of probabilistic safety assessment (PSA). The analytical basis for the ideas presented here come primarily from the authors' work in support of more than 20 PSA studies over the past several years. Timing effects are important for establishing success criteria for support and safety system response and for identifying the time available for operator recovery actions. The principal results of this paper are as follows: 1. Analysis of event sequence timing is necessary for meaningful probabilistic safety assessment - both the success criteria for systems performance and the probability of recovery are tightly linked to sequence timing. 2. Simple engineering analyses based on first principles are often sufficient to provide adequate resolution of the time available for recovery of PSA scenarios. Only those parameters that influence sequence timing and its variability and uncertainty need be examined. 3. Time available for recovery is the basic criterion for evaluation of human performance, whether time is an explicit parameter of the operator actions analysis or not. (author)

  19. 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.

  20. A Probabilistic Safety Assessment of a Pyro-processed Waste Repository

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Jeong, Jong Tae

    2012-01-01

    A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.

  1. A Survey on Formal Verification Techniques for Safety-Critical Systems-on-Chip

    Directory of Open Access Journals (Sweden)

    Tomás Grimm

    2018-05-01

    Full Text Available The high degree of miniaturization in the electronics industry has been, for several years, a driver to push embedded systems to different fields and applications. One example is safety-critical systems, where the compactness in the form factor helps to reduce the costs and allows for the implementation of new techniques. The automotive industry is a great example of a safety-critical area with a great rise in the adoption of microelectronics. With it came the creation of the ISO 26262 standard with the goal of guaranteeing a high level of dependability in the designs. Other areas in the safety-critical applications domain have similar standards. However, these standards are mostly guidelines to make sure that designs reach the desired dependability level without explicit instructions. In the end, the success of the design to fulfill the standard is the result of a thorough verification process. Naturally, the goal of any verification team dealing with such important designs is complete coverage as well as standards conformity, but as these are complex hardware, complete functional verification is a difficult task. From the several techniques that exist to verify hardware, where each has its pros and cons, we studied six well-established in academia and in industry. We can divide them into two categories: simulation, which needs extremely large amounts of time, and formal verification, which needs unrealistic amounts of resources. Therefore, we conclude that a hybrid approach offers the best balance between simulation (time and formal verification (resources.

  2. Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

    International Nuclear Information System (INIS)

    Chang, Y.H.; Mosleh, A.; Dang, V.N.

    2003-01-01

    The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

  3. The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Vail, J.

    1992-01-01

    Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

  4. Residual Heat Removal System qualitative probabilistic safety analysis before and after auto closure interlock removal

    International Nuclear Information System (INIS)

    Mikulicic, V.; Simic, Z.

    1992-01-01

    The analysis evaluates the consequences of the removal of the auto closure interlock (ACI) on the Residual Heat Removal System (RHRS) suction/isolation valves at the nuclear power plant. The deletion of the RHRS ACI is in part based on a probabilistic safety analysis (PSA) which justifies the removal based on a criterion of increased availability and reliability. Three different areas to be examined in PSA: the likelihood of an interfacing system LOCA; RHRS availability and reliability; and low temperature overpressurization control. The paper emphasizes particularly the RHRS unavailability and reliability evaluation utilizing the current control circuitry configuration and then with the proposed modification to the control circuitry. (author)

  5. An application of probabilistic safety assessment methods to model aircraft systems and accidents

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

    1998-08-01

    A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

  6. Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.H.; Mosleh, A.; Dang, V.N

    2003-03-01

    The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

  7. Use of probabilistic safety assessment for nuclear installations with large inventory of radioactive material

    International Nuclear Information System (INIS)

    1993-06-01

    Experts from several countries, including most of the countries with major nuclear fuel reprocessing programmes, presented their work and related experience in the area of probabilistic safety assessment (PSA) for non-reactor nuclear facilities. The report drafted during the meeting focuses on the following topics: review of experience from PSAs for different types of facilities; development of a structured framework for conducting PSAs for non-reactor nuclear facilities; recommendations regarding the enhancement of information exchange on related matters among Member States; recommendations on areas which need further development and support. 9 papers were presented. A separate abstract was prepared for each of them. Refs, figs and tabs

  8. Probabilistic safety assessment of WWER440 reactors prediction, quantification and management of the risk

    CERN Document Server

    Kovacs, Zoltan

    2014-01-01

    The aim of this book is to summarize probabilistic safety assessment (PSA) of nuclear power plants with WWER440 reactors and  demonstrate that the plants are safe enough for producing energy even in light of the Fukushima accident. The book examines level 1 and 2 full power, low power and shutdown PSA, and summarizes the author's experience gained during the last 35 years in this area. It provides useful examples taken from PSA training courses the author has lectured and organized by the International Atomic Energy Agency. Such training courses were organised in Argonne National Laboratory (

  9. Probabilistic analysis of some safety aspects of a swimming pool reactor

    International Nuclear Information System (INIS)

    Lieber, K.; Nicolescu, T.

    1984-01-01

    A probabilistic risk analysis of some safety aspects without the investigation of radioactivity release has been performed for the 10 MW (thermal) swimming-pool research reactor SAPHIR. Our presentation is focused on the 7 internal initiating events found to be relevant with respect to accident sequences that could result with core melt due to loss of coolant or overcriticality. The results are given by the core melt frequencies for the investigated accident sequences. It could be demonstrated by our investigation that the core melt hazard of the reactor is extremely low. (author)

  10. Procedure proposed for performance of a probabilistic safety analysis for the event of ''Air plane crash''

    International Nuclear Information System (INIS)

    Hoffmann, H.H.

    1998-01-01

    A procedures guide for a probabilistic safety analysis for the external event 'Air plane crash' has been prepared. The method is based on analysis done within the framework of PSA for German NPPs as well as on international documents. Both crashes of military air planes and commercial air planes contribute to the plant risk. For the determination of the plant related crash rate the air traffic will be divided into 3 different categories of air traffic: - The landing and takeoff phase, - the airlane traffic and waiting loop traffic, - the free air traffic, and the air planes into different types and weight classes. (orig./GL) [de

  11. The safety assessment of OPR-1000 nuclear power plant for station blackout accident applying the combined deterministic and probabilistic procedure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2014-08-15

    Highlights: • The combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. • The safety assessment of OPR-1000 nuclear power plant for SBO accident is performed by applying the CDPP. • By estimating the offsite power restoration time appropriately, the SBO risk is reevaluated. • It is concluded that the CDPP is applicable to safety assessment of BDBAs without significant erosion of the safety margin. - Abstract: Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

  12. Probabilistic safety goals for nuclear power plants; Phases 2-4. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bengtsson, L.; Knochenhauer, M. (Scandpower AB (Sweden)); Holmberg, J.-E.; Rossi, J. (VTT Technical Research Centre of Finland (Finland))

    2011-05-15

    Safety goals are defined in different ways in different countries and also used differently. Many countries are presently developing them in connection to the transfer to risk-informed regulation of both operating nuclear power plants (NPP) and new designs. However, it is far from self-evident how probabilistic safety criteria should be defined and used. On one hand, experience indicates that safety goals are valuable tools for the interpretation of results from a probabilistic safety assessment (PSA), and they tend to enhance the realism of a risk assessment. On the other hand, strict use of probabilistic criteria is usually avoided. A major problem is the large number of different uncertainties in a PSA model, which makes it difficult to demonstrate the compliance with a probabilistic criterion. Further, it has been seen that PSA results can change a lot over time due to scope extensions, revised operating experience data, method development, changes in system requirements, or increases of level of detail, mostly leading to an increase of the frequency of the calculated risk. This can cause a problem of consistency in the judgments. This report presents the results from the second, third and fourth phases of the project (2007-2009), which have dealt with providing guidance related to the resolution of some specific problems, such as the problem of consistency in judgement, comparability of safety goals used in different industries, the relationship between criteria on different levels, and relations between criteria for level 2 and 3 PSA. In parallel, additional context information has been provided. This was achieved by extending the international overview by contributing to and benefiting from a survey on PSA safety criteria which was initiated in 2006 within the OECD/NEA Working Group Risk. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by

  13. Probabilistic safety goals for nuclear power plants; Phases 2-4. Final report

    International Nuclear Information System (INIS)

    Bengtsson, L.; Knochenhauer, M.; Holmberg, J.-E.; Rossi, J.

    2011-05-01

    Safety goals are defined in different ways in different countries and also used differently. Many countries are presently developing them in connection to the transfer to risk-informed regulation of both operating nuclear power plants (NPP) and new designs. However, it is far from self-evident how probabilistic safety criteria should be defined and used. On one hand, experience indicates that safety goals are valuable tools for the interpretation of results from a probabilistic safety assessment (PSA), and they tend to enhance the realism of a risk assessment. On the other hand, strict use of probabilistic criteria is usually avoided. A major problem is the large number of different uncertainties in a PSA model, which makes it difficult to demonstrate the compliance with a probabilistic criterion. Further, it has been seen that PSA results can change a lot over time due to scope extensions, revised operating experience data, method development, changes in system requirements, or increases of level of detail, mostly leading to an increase of the frequency of the calculated risk. This can cause a problem of consistency in the judgments. This report presents the results from the second, third and fourth phases of the project (2007-2009), which have dealt with providing guidance related to the resolution of some specific problems, such as the problem of consistency in judgement, comparability of safety goals used in different industries, the relationship between criteria on different levels, and relations between criteria for level 2 and 3 PSA. In parallel, additional context information has been provided. This was achieved by extending the international overview by contributing to and benefiting from a survey on PSA safety criteria which was initiated in 2006 within the OECD/NEA Working Group Risk. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by

  14. Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 & Vol 2

    Energy Technology Data Exchange (ETDEWEB)

    PARSONS, J.E.

    2000-07-15

    The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented.

  15. Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 and Vol 2

    CERN Document Server

    Parsons, J E

    2000-01-01

    The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented.

  16. Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 and Vol 2

    International Nuclear Information System (INIS)

    PARSONS, J.E.

    2000-01-01

    The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented

  17. A probabilistic method for optimization of fire safety in nuclear power plants

    International Nuclear Information System (INIS)

    Hosser, D.; Sprey, W.

    1986-01-01

    As part of a comprehensive fire safety study for German Nuclear Power Plants a probabilistic method for the analysis and optimization of fire safety has been developed. It follows the general line of the American fire hazard analysis, with more or less important modifications in detail. At first, fire event trees in selected critical plant areas are established taking into account active and passive fire protection measures and safety systems endangered by the fire. Failure models for fire protection measures and safety systems are formulated depending on common parameters like time after ignition and fire effects. These dependences are properly taken into account in the analysis of the fire event trees with the help of first-order system reliability theory. In addition to frequencies of fire-induced safety system failures relative weights of event paths, fire protection measures within these paths and parameters of the failure models are calculated as functions of time. Based on these information optimization of fire safety is achieved by modifying primarily event paths, fire protection measures and parameters with the greatest relative weights. This procedure is illustrated using as an example a German 1300 MW PWR reference plant. It is shown that the recommended modifications also reduce the risk to plant personnel and fire damage

  18. Safety verification of non-linear hybrid systems is quasi-decidable

    Czech Academy of Sciences Publication Activity Database

    Ratschan, Stefan

    2014-01-01

    Roč. 44, č. 1 (2014), s. 71-90 ISSN 0925-9856 R&D Projects: GA ČR GCP202/12/J060 Institutional support: RVO:67985807 Keywords : hybrid system s * safety verification * decidability * robustness Subject RIV: IN - Informatics, Computer Science Impact factor: 0.875, year: 2014

  19. A Particle System for Safety Verification of Free Flight in Air Traffic

    NARCIS (Netherlands)

    Blom, H.A.P.; Krystul, J.; Bakker, G.J.

    2006-01-01

    Under free flight, an aircrew has both the freedom to select their trajectory and the responsibility of resolving conflicts with other aircraft. The general belief is that free flight can be made safe under low traffic conditions. Increasing traffic, however, raises safety verification issues. This

  20. Risk-Informed Safety Assurance and Probabilistic Assessment of Mission-Critical Software-Intensive Systems

    Science.gov (United States)

    Guarro, Sergio B.

    2010-01-01

    This report validates and documents the detailed features and practical application of the framework for software intensive digital systems risk assessment and risk-informed safety assurance presented in the NASA PRA Procedures Guide for Managers and Practitioner. This framework, called herein the "Context-based Software Risk Model" (CSRM), enables the assessment of the contribution of software and software-intensive digital systems to overall system risk, in a manner which is entirely compatible and integrated with the format of a "standard" Probabilistic Risk Assessment (PRA), as currently documented and applied for NASA missions and applications. The CSRM also provides a risk-informed path and criteria for conducting organized and systematic digital system and software testing so that, within this risk-informed paradigm, the achievement of a quantitatively defined level of safety and mission success assurance may be targeted and demonstrated. The framework is based on the concept of context-dependent software risk scenarios and on the modeling of such scenarios via the use of traditional PRA techniques - i.e., event trees and fault trees - in combination with more advanced modeling devices such as the Dynamic Flowgraph Methodology (DFM) or other dynamic logic-modeling representations. The scenarios can be synthesized and quantified in a conditional logic and probabilistic formulation. The application of the CSRM method documented in this report refers to the MiniAERCam system designed and developed by the NASA Johnson Space Center.

  1. The Barselina Project Phase 4 Summary report. Ignalina Unit 2 Probabilistic Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, Gunnar [ES-Konsult AB, Stockholm (Sweden); Hellstroem, P. [RELCON AB, Solna (Sweden); Zheltobriuch, G.; Bagdonas, A. [Ignalina Power Plant, Visaginas (Lithuania)

    1996-12-01

    The Barselina Project was initiated in the summer of 1991. The project is a multilateral co-operation between Lithuania, Russia and Sweden. The long range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. The Swedish BWR Barsebaeck is used as reference plant and the Lithuanian RBMK Ignalina as application plant. During phase 3, from March, 1993 to June, 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 (INPP-2) was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. To increase the realism of the risk model a set of deterministic analyses were performed and plant/RBMK-specific data bases were developed and used. A general concept for analysing this type of reactor was developed. During phase 4, July 1994 to September 1996, the PSA was further developed, taking into account plant changes, improved modeling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The updated model is quantified and new results and conclusions are evaluated.

  2. Selection of important initiating events for Level 1 probabilistic safety assessment study at Puspati TRIGA Reactor

    International Nuclear Information System (INIS)

    Maskin, M.; Charlie, F.; Hassan, A.; Prak Tom, P.; Ramli, Z.; Mohamed, F.

    2016-01-01

    Highlights: • Identifying possible important initiating events (IEs) for Level 1 probabilistic safety assessment performed on research nuclear reactor. • Methods in screening and grouping IEs are addressed. • Focusing only on internal IEs due to random failures of components. - Abstract: This paper attempts to present the results in identifying possible important initiating events (IEs) as comprehensive as possible to be applied in the development of Level-1 probabilistic safety assessment (PSA) study. This involves the approaches in listing and the methods in screening and grouping IEs, by focusing only on the internal IEs due to random failures of components and human errors with full power operational conditions and reactor core as the radioactivity source. Five approaches were applied in listing the IEs and each step of the methodology was described and commented. The criteria in screening and grouping the IEs were also presented. The results provided the information on how the Malaysian PSA team applied the approaches in selecting the most probable IEs as complete as possible in order to ensure the set of IEs was identified systematically and as representative as possible, hence providing confidence to the completeness of the PSA study. This study is perhaps one of the first to address classic comprehensive steps in identifying important IEs to be used in a Level-1 PSA study.

  3. A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Aldemir, Tunc

    2013-01-01

    Highlights: ► Dynamic methodologies for probabilistic safety assessment (PSA) are surveyed. ► These methodologies overcome the limitations of the traditional approach to PSA. ► They are suitable for PSA using a best estimate plus uncertainty approach. ► They are highly computation intensive and produce very large number of scenarios. ► Use of scenario clustering can assist the analysis of the results. -- Abstract: Dynamic methodologies for probabilistic safety assessment (PSA) are defined as those which use a time-dependent phenomenological model of system evolution along with its stochastic behavior to account for possible dependencies between failure events. Over the past 30 years, numerous concerns have been raised in the literature regarding the capability of the traditional static modeling approaches such as the event-tree/fault-tree methodology to adequately account for the impact of process/hardware/software/firmware/human interactions on the stochastic system behavior. A survey of the types of dynamic PSA methodologies proposed to date is presented, as well as a brief summary of an example application for the PSA modeling of a digital feedwater control system of an operating pressurized water reactor. The use of dynamic methodologies for PSA modeling of passive components and phenomenological uncertainties are also discussed.

  4. Application of probabilistic fracture mechanics to reactor pressure vessel safety assessment

    International Nuclear Information System (INIS)

    Venturini, V.; Pitner, P.

    1995-06-01

    Among all the components of a PWR (Pressurized Water Reactor) nuclear power plant, the reactor vessel is of major importance for safety. The integrity of this structure must be guaranteed in all circumstances, even in the case of the most severe accidents, and its mechanical state can be decisive for the lifetime of the plant. The brittle rupture would be the most important of all potential hazards if the irradiation effects were not consistent with predictions. The interest of having a reliable and precise method of evaluating the available safety margins and the integrity of this component led Electricite de France (EDF) to carry out a probabilistic fracture mechanics analysis. The probabilistic model developed by integration of the uncertainties in the usual fracture mechanics equations is presented. A special focus is made on the problem of coupling thermo-mechanical finite element calculations and reliability analysis. The use of a finite element code can be associated with prohibitive computation times when it is invoked numerous times during simulations sequences or complex iterative procedures. The response surface method is used. It provides an approximation of the response from a reduced number of original data. The global approach is illustrated on an example corresponding to a specific accidental transient. A validation of the obtained results is also carried out through the comparison with an equivalent model without coupling. (author)

  5. Outcomes of an international initiative for harmonization of low power and shutdown probabilistic safety assessment

    Directory of Open Access Journals (Sweden)

    Manna Giustino

    2010-01-01

    Full Text Available Many probabilistic safety assessment studies completed to the date have demonstrated that the risk dealing with low power and shutdown operation of nuclear power plants is often comparable with the risk of at-power operation, and the main contributors to the low power and shutdown risk often deal with human factors. Since the beginning of the nuclear power generation, human performance has been a very important factor in all phases of the plant lifecycle: design, commissioning, operation, maintenance, surveillance, modification, decommissioning and dismantling. The importance of this aspect has been confirmed by recent operating experience. This paper provides the insights and conclusions of a workshop organized in 2007 by the IAEA and the Joint Research Centre of the European Commission, on Harmonization of low power and shutdown probabilistic safety assessment for WWER nuclear power plants. The major objective of the workshop was to provide a comparison of the approaches and the results of human reliability analyses and gain insights in the enhanced handling of human factors.

  6. Study of applicable methods on safety verification of disposal facilities and waste packages

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Three subjects about safety verification on the disposal of low level radioactive waste were investigated in FY. 2012. For radioactive waste disposal facilities, specs and construction techniques of covering with soil to prevent possible destruction caused by natural events (e.g. earthquake) were studied to consider verification methods for those specs. For waste packages subject to near surface pit disposal, settings of scaling factor and average radioactivity concentration (hereafter referred to as ''SF'') on container-filled and solidified waste packages generated from Kashiwazaki Kariwa Nuclear Power Station Unit 1-5, setting of cesium residual ratio of molten solidified waste generated from Tokai and Tokai No.2 Power Stations, etc. were studied. Those results were finalized in consideration of the opinion from advisory panel, and publicly opened as JNES-EV reports. In FY 2012, five JNES reports were published and these have been used as standards of safety verification on waste packages. The verification method of radioactive wastes subject to near-surface trench disposal and intermediate depth disposal were also studied. For radioactive wastes which will be returned from overseas, determination methods of radioactive concentration, heat rate and hydrogen generation rate of CSD-C were established. Determination methods of radioactive concentration and heat rate of CSD-B were also established. These results will be referred to verification manuals. (author)

  7. 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

  8. Uncertainty and sensitivity analysis on probabilistic safety assessment of an experimental facility

    International Nuclear Information System (INIS)

    Burgazzi, L.

    2000-01-01

    The aim of this work is to perform an uncertainty and sensitivity analysis on the probabilistic safety assessment of the International Fusion Materials Irradiation Facility (IFMIF), in order to assess the effect on the final risk values of the uncertainties associated with the generic data used for the initiating events and component reliability and to identify the key quantities contributing to this uncertainty. The analysis is conducted on the expected frequency calculated for the accident sequences, defined through the event tree (ET) modeling. This is in order to increment credit to the ET model quantification, to calculate frequency distributions for the occurrence of events and, consequently, to assess if sequences have been correctly selected on the probability standpoint and finally to verify the fulfillment of the safety conditions. Uncertainty and sensitivity analysis are performed using respectively Monte Carlo sampling and an importance parameter technique. (author)

  9. Review of probabilistic safety assessments: insights and recommendations regarding further developments

    International Nuclear Information System (INIS)

    Spitzer, C.

    1996-01-01

    Probabilistic Safety Assessments (PSAs) performed by utilities in the framework of Periodic Safety Reviews for German nuclear power plants are reviewed by TUeV Suedwest. Insights gained and recommendations concerning the necessity and focus of further developments and applications according to practical requests for the performance and assessment of PSAs within regulatory procedures are presented in this paper. Further on, recommendations are made in order to ensure the validity of the results of PSAs necessary in order to achieve the goals thereof. Beside some general points of view the emphasis of the paper is on methodological aspects with respect to evaluation methods and assessment of common cause failures as well as human reliability assessment

  10. Development of probabilistic methods for safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Schott, H.; Berg, H.P.

    1998-01-01

    Since its introduction by the German Risk Study, Probabilistic Safety Assessment (PSA) has developed in Germany to a valuable tool in regulatory decision-making. Plant specific PSAs of Level 1+ are now conducted for all nuclear power plants in the frame of Periodic Safety Reviews. This paper is devoted to the description or key elements set out in the regulatory guidelines for PSA-Level 1+ and the corresponding technical documents and the further development of PSA methodology in the Federal Republic of Germany. In the course of the next years it is intended to make progress in the modeling of common cause failures, human reliability evaluation, reduction of uncertainties in PSA modeling techniques and data estimation, analysis of low power and shut down states as well as in reaching a mature methodology for inclusion of external events into the analysis. (author)

  11. Human Reliability in Probabilistic Safety Assessments; Fiabilidad Humana en los Analisis Probabilisticos de Seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Nunez Mendez, J

    1989-07-01

    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.

  12. Consideration of probabilistic safety objectives in OECD/NEA member countries: Short overview and update

    International Nuclear Information System (INIS)

    Versteeg, M.F.; Andrews, R.M.

    1994-01-01

    Almost every member country of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) uses probabilistic safety criteria (PSC), in one way or another, for the safety assessment of nuclear power plants. The choice of the PSC, their applicability, and whether or not these PSC are used in a formal and/or legal way, is dependent on the political and regulatory situation. The spectrum of utilization includes the use as design requirements and the use as a regulatory and licensing tool be the authorities. The paper summarises the various PSC applied to the assessment of nuclear power plant in the OECD member countries and presents in more detail the use of PSC on the public health level in the Netherlands, United Kingdom and USA. 10 refs, 1 fig., 6 tabs

  13. Impact of support system failure limitations on probabilistic safety assessment and in regulatory decision making

    International Nuclear Information System (INIS)

    Bickel, J.H.

    1990-01-01

    When used as a tool for safety decision making, Probabilistic Safety Assessment (PSA) is as effective as it realistically characterizes the overall frequency and consequences of various types of system and component failures. If significant support system failure events are omitted from consideration, the PSA process omits the characterization of possible unique contributors to core damage risk, possibly underestimates the frequency of core damage, and reduces the future utility of the PSA as a decision making tool for the omitted support system. This paper is based on a review of several recent US PSA studies and the author's participation in several International Atomic Energy Agency (IAEA) sponsored peer reviews. 21 refs., 2 figs., 1 tab

  14. Human actions in the pre-operational probabilistic safety analysis of Juragua Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ferro, R.

    1995-01-01

    Human error is one of the main contributors to the biggest industrial disasters that the world has suffered in the last years. Safety probabilistic analysis techniques allow to consider, in the some study, the contribution of a facility's mechanical and human components safety, this guaranteeing a move integral assessment of these two factors although the PSA study of Juragua Nuclear Power Plant is carried out at a preoperational stage which causes important information limitations fos assessment of human reliability some considerations and suppositions in order to conduct treatment of human actions this stage were adopted. The present work describes the projected targets, approach applied and results obtained from the lakes of human reliability of this study

  15. Marked point process framework for living probabilistic safety assessment and risk follow-up

    International Nuclear Information System (INIS)

    Arjas, Elja; Holmberg, Jan

    1995-01-01

    We construct a model for living probabilistic safety assessment (PSA) by applying the general framework of marked point processes. The framework provides a theoretically rigorous approach for considering risk follow-up of posterior hazards. In risk follow-up, the hazard of core damage is evaluated synthetically at time points in the past, by using some observed events as logged history and combining it with re-evaluated potential hazards. There are several alternatives for doing this, of which we consider three here, calling them initiating event approach, hazard rate approach, and safety system approach. In addition, for a comparison, we consider a core damage hazard arising in risk monitoring. Each of these four definitions draws attention to a particular aspect in risk assessment, and this is reflected in the behaviour of the consequent risk importance measures. Several alternative measures are again considered. The concepts and definitions are illustrated by a numerical example

  16. Probabilistic safety assessments of nuclear power plants for low power and shutdown modes

    International Nuclear Information System (INIS)

    2000-03-01

    Within the past several years the results of nuclear power plant operating experience and performance of probabilistic safety assessments (PSAs) for low power and shutdown operating modes have revealed that the risk from operating modes other than full power may contribute significantly to the overall risk from plant operations. These early results have led to an increased focus on safety during low power and shutdown operating modes and to an increased interest of many plant operators in performing shutdown and low power PSAs. This publication was developed to provide guidance and insights on the performance of PSA for shutdown and low power operating modes. The preparation of this publication was initiated in 1994. Two technical consultants meetings were conducted in 1994 and one in February 1999 in support of the development of this report

  17. The Use and Development of Probabilistic Safety Assessment in NEA Member Countries

    International Nuclear Information System (INIS)

    2002-01-01

    The mission of the CSNI is to assist Member countries in maintaining and further developing the scientific and technical knowledge base required to assess the safety of nuclear reactors and fuel cycle facilities. The mission of the Working Group on Risk Assessment (WGRisk) is to advance the understanding and utilisation of Probabilistic Safety Assessment (PSA) in ensuring continued safety of nuclear installations in Member countries. In pursuing this goal, the Working Group shall recognize the different methodologies for identifying contributors to risk and assessing their importance. While the Working Group shall continue to focus on the more mature PSA methodologies for Level 1, Level 2, internal, external, shutdown, etc. It shall also consider the applicability and maturity of PSA methods for considering evolving issues such as human reliability, software reliability, ageing issues, etc., as appropriate. This report provides descriptions of the current status of PSA programmes in Member countries including basic background information, guidelines, various PSA applications, major results in recent studies, PSA based plant modifications and research and development topics. While the compilation is a not complete compilation it provides a 'snapshot' of the current situation in the Member countries and hence it provides reference information and various insights to both the PSA practician and others involved in the nuclear industry. The terms PSA (Probabilistic Safety Assessment) and PRA (Probabilistic Risk Assessment) are utilised to denote this subject. In each of the chapters the objective is to present a 'snapshot' of the current status. The main issues considered in the different chapters are Background Information, Quantitative Safety Guidelines, Status of PSA Programmes, PSA Applications, PSA Related Research and Development and PSA Based Plant Modifications. It is important to note that the information contained in this report represents current practices in

  18. Risk-Based Predictive Maintenance for Safety-Critical Systems by Using Probabilistic Inference

    Directory of Open Access Journals (Sweden)

    Tianhua Xu

    2013-01-01

    Full Text Available Risk-based maintenance (RBM aims to improve maintenance planning and decision making by reducing the probability and consequences of failure of equipment. A new predictive maintenance strategy that integrates dynamic evolution model and risk assessment is proposed which can be used to calculate the optimal maintenance time with minimal cost and safety constraints. The dynamic evolution model provides qualified risks by using probabilistic inference with bucket elimination and gives the prospective degradation trend of a complex system. Based on the degradation trend, an optimal maintenance time can be determined by minimizing the expected maintenance cost per time unit. The effectiveness of the proposed method is validated and demonstrated by a collision accident of high-speed trains with obstacles in the presence of safety and cost constrains.

  19. Probabilistic safety analyses. Status and further development of methods and models, applications

    International Nuclear Information System (INIS)

    Berg, H.P.; Schott, H.

    1992-12-01

    The report describes the topics of the deterministic and probabilistic approach. The PSA is used in order to investigate event sequences beyond design limits; in particular the expected frequency of core melting is important. The basis of PSA is described including its limits. Moreover, the current state of the art of science and technology in the field of PSA including the so-called 'living PSA' are explained. Some measures which result in order to improve the safety of a nuclear power plant from the German Risk-Study are shown. An overview is given on the status of PSA in periodic safety reviews in German nuclear power plants. Moreover, the main topics of running investigations are presented. (orig.) [de

  20. Human Reliability in Probabilistic Safety Assessments; Fiabilidad Humana en los Analisis Probabilisticos de Seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Nunez Mendez, J.

    1989-07-01

    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.

  1. Application of probabilistic methods for sizing of safety factors in studies on defect harm fullness

    International Nuclear Information System (INIS)

    Ardillon, E.; Pitner, P.

    1996-01-01

    The design rules that are currently under application in nuclear engineering recommend the use of deterministic analysis methods. Probabilistic methods allow the uncertainties inherent in input variables of the analytical model to be taken into account owing to data provided by operation feedback so as to better evaluate the link between the deterministic margins adopted and the actual risk level. In the Resistance R/Loading L elementary case where the variables are Gaussian, there is an explicit relation between the required safety level and the partial safety coefficients which affect each variable. In the complex case of a flawed pipe subjected to various modes of ruin where many random variables are not Gaussian, one can obtain implicit relations. These relations allow a certain flexibility when choosing the coefficients, which poses the problem of their optimum calibration. The choice of coefficients based upon the coordinates of the ''most probable failure point'' illustrates this approach. (authors). 7 refs., 5 figs., 2 tabs

  2. Probabilistic Causal Analysis for System Safety Risk Assessments in Commercial Air Transport

    Science.gov (United States)

    Luxhoj, James T.

    2003-01-01

    Aviation is one of the critical modes of our national transportation system. As such, it is essential that new technologies be continually developed to ensure that a safe mode of transportation becomes even safer in the future. The NASA Aviation Safety Program (AvSP) is managing the development of new technologies and interventions aimed at reducing the fatal aviation accident rate by a factor of 5 by year 2007 and by a factor of 10 by year 2022. A portfolio assessment is currently being conducted to determine the projected impact that the new technologies and/or interventions may have on reducing aviation safety system risk. This paper reports on advanced risk analytics that combine the use of a human error taxonomy, probabilistic Bayesian Belief Networks, and case-based scenarios to assess a relative risk intensity metric. A sample case is used for illustrative purposes.

  3. Plant Operation Station for HTR-PM Low Power and Shutdown operation Probabilistic safety analysis

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan

    2014-01-01

    Full range Probabilistic safety analysis (PSA) is one of key conditions for nuclear power plant (NPP) licensing according to the requirement of nuclear safety regulatory authority. High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) has developed construction design and prepared for the charging license application. So after the normal power operation PSA submitted for review, the Low power and Shutdown operation Probabilistic safety analysis (LSPSA) also begin. The results of LSPSA will together with prior normal power PSA results to demonstrate the safety level of HTR-PM NPP Plant Operation Station (POS) is one of important terms in LSPSA. The definition of POS lays the foundation for LSPSA modeling. POS provides initial and boundary conditions for the following event tree and fault tree model development. The aim of this paper is to describe the state-of-the-art of POS definition for HTR-PM LSPSA. As for the first attempt to the high temperature gas cooled reactor module plant, the methodology and procedure of POS definition refers to the LWR LSPSA guidance, and adds to plant initial status analysis due to the HTR-PM characteristics. A specific set of POS grouping vectors is investigate and suggested for HTR-PM NPP, which reflects the characteristics of plant modularization and on-line refueling. As a result, seven POSs are given according to the grouping vectors at the end of the paper. They will be used to the LSPSA modelling and adjusted if necessary. The papers ’work may provide reference to the analogous NPP LSPSA. (author)

  4. Model Verification and Validation Concepts for a Probabilistic Fracture Assessment Model to Predict Cracking of Knife Edge Seals in the Space Shuttle Main Engine High Pressure Oxidizer

    Science.gov (United States)

    Pai, Shantaram S.; Riha, David S.

    2013-01-01

    Physics-based models are routinely used to predict the performance of engineered systems to make decisions such as when to retire system components, how to extend the life of an aging system, or if a new design will be safe or available. Model verification and validation (V&V) is a process to establish credibility in model predictions. Ideally, carefully controlled validation experiments will be designed and performed to validate models or submodels. In reality, time and cost constraints limit experiments and even model development. This paper describes elements of model V&V during the development and application of a probabilistic fracture assessment model to predict cracking in space shuttle main engine high-pressure oxidizer turbopump knife-edge seals. The objective of this effort was to assess the probability of initiating and growing a crack to a specified failure length in specific flight units for different usage and inspection scenarios. The probabilistic fracture assessment model developed in this investigation combined a series of submodels describing the usage, temperature history, flutter tendencies, tooth stresses and numbers of cycles, fatigue cracking, nondestructive inspection, and finally the probability of failure. The analysis accounted for unit-to-unit variations in temperature, flutter limit state, flutter stress magnitude, and fatigue life properties. The investigation focused on the calculation of relative risk rather than absolute risk between the usage scenarios. Verification predictions were first performed for three units with known usage and cracking histories to establish credibility in the model predictions. Then, numerous predictions were performed for an assortment of operating units that had flown recently or that were projected for future flights. Calculations were performed using two NASA-developed software tools: NESSUS(Registered Trademark) for the probabilistic analysis, and NASGRO(Registered Trademark) for the fracture

  5. Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

    International Nuclear Information System (INIS)

    Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

    2009-01-01

    A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

  6. Safety evaluation by living probabilistic safety assessment. Procedures and applications for planning of operational activities and analysis of operating experience

    International Nuclear Information System (INIS)

    Johanson, Gunnar; Holmberg, J.

    1994-01-01

    Living Probabilistic Safety Assessment (PSA) is a daily safety management system and it is based on a plant-specific PSA and supporting information systems. In the living use of PSA, plant status knowledge is used to represent actual plant safety status in monitoring or follow-up perspective. The PSA model must be able to express the risk at a given time and plant configuration. The process, to update the PSA model to represent the current or planned configuration and to use the model to evaluate and direct the changes in the configuration, is called living PSA programme. The main purposes to develop and increase the usefulness of living PSA are: Long term safety planning: To continue the risk assessment process started with the basic PSA by extending and improving the basic models and data to provide a general risk evaluation tool for analyzing the safety effects of changes in plant design and procedures. Risk planning of operational activities: To support the operational management by providing means for searching optimal operational maintenance and testing strategies from the safety point of view. The results provide support for risk decision making in the short term or in a planning mode. The operational limits and conditions given by technical specifications can be analyzed by evaluating the risk effects of alternative requirements in order to balance the requirements with respect to operational flexibility and plant economy. Risk analysis of operating experience: To provide a general risk evaluation tool for analyzing the safety effects of incidents and plant status changes. The analyses are used to: identify possible high risk situations, rank the occurred events from safety point of view, and get feedback from operational events for the identification of risk contributors. This report describes the methods, models and applications required to continue the process towards a living use of PSA. 19 tabs, 20 figs

  7. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    International Nuclear Information System (INIS)

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events

  8. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

  9. Best estimate probabilistic safety assessment results for the Westinghouse Advanced Loop Tester (WALT)

    International Nuclear Information System (INIS)

    Wang, Guoqiang; Xu, Yiban; Oelrich, Robert L. Jr.; Byers, William A.; Young, Michael Y.; Karoutas, Zeses E.

    2011-01-01

    The nuclear industry uses the probabilistic safety assessment (PSA) technique to improve safety decision making and operation. The methodology evaluates the system reliability, which is defined as the probability of system success, and the postulated accident/problematic scenarios of systems for the nuclear power plants or other facilities. The best estimate probabilistic safety assessment (BE-PSA) method of evaluating system reliability and postulated problematic scenarios will produce more detailed results of interest, such as best estimated reliability analysis and detailed thermal hydraulic calculations using a sub-channel or Computational Fluid Dynamics (CFD) code. The methodology is typically applied to reactors, but can also be applied to any system such as a test facility. In this paper, a BE-PSA method is introduced and used for evaluating the Westinghouse Advanced Loop Tester (WALT). The WALT test loop at the George Westinghouse Science and Technology Center (STC), which was completed in October 2005, is designed to be utilized to model the top grid span of a hot rod in a fuel assembly under the Pressurizer Water Reactor (PWR) normal operating conditions. In order to safely and successfully operate the WALT test loop and correctly use the WALT experimental data, it is beneficial to perform a probabilistic safety assessment and analyze the thermal hydraulic results for the WALT loop in detail. Since October 2005, a number of test runs have been performed on the WALT test facility designed and fabricated by Westinghouse Electric Company LLC. This paper briefly describes the BE-PSA method and performs BE-PSA for the WALT loop. Event trees linked with fault trees embedding thermal hydraulic analysis models, such as sub-channel and/or CFD models, were utilized in the analyses. Consequently, some selected useful experimental data and analysis results are presented for future guidance on WALT and/or other similar test facilities. For example, finding and

  10. Use of probabilistic safety assessment in the regulatory process. Report of the technical committee meeting. Working material

    International Nuclear Information System (INIS)

    1994-01-01

    A Technical Committee Meeting (TCM) was organized between 5-8 December 1994 to discuss and review the international situation in connection with the use made, whether formally or informally, by regulatory bodies of probabilistic safety assessment (PSA) in the course of their work, and the related question of the use and value of adopting probabilistic safety criteria (PSC) as an aid to judging the results of PSAs. The document includes the output from the four working groups, as well as 11 papers from the 12 papers presented to the TCM. A separate abstract was prepared for each paper. Refs, figs, tabs

  11. Probabilistic and deterministic safety study of the transportation of liquefied gases in the vicinity of a nuclear site

    International Nuclear Information System (INIS)

    Gobert, T.; Lannoy, A.

    1982-01-01

    The safety analyses for nuclear power plants devotes special attention to the evaluation of hazards which may be induced by industrial activity in the environment of nuclear sites. For instance, explosion of a drifting gas cloud resulting from an accidental release of liquefied gas may jeopardize the plant safety. The paper presents the methodology, both probabilistic and deterministic, followed by Electricite de France to evaluate these risks. It particularly shows that the probabilistic approach is strongly linked with the definition of ''design basis accidents'' and the evaluation of their effects

  12. Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

    International Nuclear Information System (INIS)

    Ruebel, Andre; Meleshyn, Artur

    2014-08-01

    The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

  13. Safety technical considerations on the 2012 periodic safety verification of the Beznau nuclear power plant

    International Nuclear Information System (INIS)

    2016-12-01

    According to nuclear legislation, the owner of an operational license for a nuclear power plant has to provide a periodic safety verification (PSU) every 10 years. The 'North Eastern Power Plants' company (NOK), today AXPO Power AG already performed such a PSU for the Beznau-2 nuclear reactor block (KKB2) in 2002. The Beznau-1 nuclear reactor block (KKB1) received its definitive operational license in October 1970, after test operation during 7 months. After the license for test operation received on July 16 th , 1971, the operational license of KKB2 was renewed several times, each time for a certain period of validity. In 1991, NOK requested a definitive operational license for KKB2, but in 1994 the Swiss Federal Council lengthened the license for only 10 years. Moreover, it laid down that NOK has to periodically report on the safety of the facility. With its letter of August 23 rd , 1998, the Federal Office of Energy defined the documents to be produced for the PSU. The extent of the PSU was defined in such a way that many documents concern the whole power plant, i.e. both nuclear reactor blocks. On December 3 rd , 2004, the Swiss Federal Council granted KKB2 an operational license of limited validity. The present report reviews the 2012 PSU, which covers the time interval from January 1 st , 2002, to December 31 st , 2011, from the point of view of safety. It contains documents for the evaluation of both reactor blocks at KKB. The Beznau interim storage pool was also taken into consideration; it is situated on the KKB site, but, according to a decision of the Swiss Federal Council of May 23 rd , 1991, it has an independent operational license. The evaluation of ageing surveillance takes the whole operational period of the facility into account, i.e. the ageing mechanisms acting as from the beginning of the operation. Moreover, important developments that occurred after the surveillance time interval have been taken into account, especially the status

  14. Probabilistic safety assessment technology for commercial nuclear power plant security evaluation

    International Nuclear Information System (INIS)

    Liming, J.K.; Johnson, D.H.; Dykes, A.A.

    2004-01-01

    Commercial nuclear power plant physical security has received much more intensive treatment and regulatory attention since September 11, 2001. In light of advancements made by the nuclear power industry in the field of probabilistic safety assessment (PSA) for its power plants over that last 30 years, and given the many examples of successful applications of risk-informed regulation at U. S. nuclear power plants during recent years, it may well be advisable to apply a 'risk-informed' approach to security management at nuclear power plants from now into the future. In fact, plant PSAs developed in response to NRC Generic Letter 88-20 and related requirements are used to help define target sets of critical plant safety equipment in our current security exercises for the industry. With reasonable refinements, plant PSAs can be used to identify, analyze, and evaluate reasonable and prudent approaches to address security issues and associated defensive strategies at nuclear power plants. PSA is the ultimate scenario-based approach to risk assessment, and thus provides a most powerful tool in identifying and evaluating potential risk management decisions. This paper provides a summary of observations of factors that are influencing or could influence cost-effective or 'cost-reasonable' security management decision-making in the current political environment, and provides recommendations for the application of PSA tools and techniques to the nuclear power plant operational safety response exercise process. The paper presents a proposed framework for nuclear power plant probabilistic terrorist risk assessment that applies these tools and techniques. (authors)

  15. Probabilistic Design

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard; Burcharth, H. F.

    This chapter describes how partial safety factors can be used in design of vertical wall breakwaters and an example of a code format is presented. The partial safety factors are calibrated on a probabilistic basis. The code calibration process used to calibrate some of the partial safety factors...

  16. Elements of the safety case for the Morsleben repository based on probabilistic modelling

    International Nuclear Information System (INIS)

    Wollrath, J.; Niemeyer, M.; Resele, G.; Becker, D.A.; Hirsekorn, P.

    2008-01-01

    The Morsleben nuclear waste repository (ERAM) for low- and intermediate-level mainly short-lived waste is located in a former salt mine. The closure concept was developed in parallel and interacting with the safety assessment. The safety concept is based on extensive backfilling with salt concrete complemented with seals between the main disposal areas and the rest of the mine building. Thus, the entire system exhibits a barrier effect through a partially redundant combination of several processes. However, in the formal safety assessment no credit is taken from the barrier effect of the extensive backfill. In the safety assessments, the different possibilities of system development, the resulting array of potential fluid movement and a large number of potential radionuclide migration pathways are mapped in the bandwidth of derived parameters. The calculated response of the system to parameter variations is non-linear. Different processes may compete and compensate each other. Hence, the common practice to choose a conservative parameter set for the safety assessment is a priori impossible. The safety assessment has been performed independently by two groups with different computer models, for the same closure concept and the same basic parameters but independent conceptual approaches. Both groups have performed deterministic and probabilistic dose calculations. The results match well; the differences can be explained on basis of the model approaches. Although a large bandwidth is considered for a number of parameters the maximum radiation exposure remains clearly below the applicable dose limit for nearly all calculations, demonstrating the robustness of the system. These aspects significantly contribute to confidence building in the Safety Case for ERAM. (authors)

  17. Human reliability analysis in probabilistic safety assessment for nuclear power plants. A Safety Practice. A publication within the NUSS programme

    International Nuclear Information System (INIS)

    1995-01-01

    Probabilistic safety assessment (PSA) is playing an increasingly important role in the safe operation of nuclear power plants throughout the world. In order to establish a consistent framework for conducting PSA studies, for promoting technology transfer of the state of the art, and for encouraging uniformity in the way PSA is carried out, the IAEA is preparing a set of publications which gives guidance on various aspects of PSA. This document presents a practical approach for incorporating human reliability analysis (HRA) into PSA. It describes the steps needed and the documentation that should be provided both to support the PSA itself and to ensure effective communication of important information arising from the studies. It also describes a framework for analysing those human actions which could affect safety and for relating such human influences to specific parts of a PSA. This Safety Practice also addresses the limitations of PSA in taking account of human factors in relation to safety and risk. Refs, figs and tabs

  18. Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

    International Nuclear Information System (INIS)

    Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

    2014-01-01

    Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

  19. Probabilistic safety assessment for instrumentation and control systems in nuclear power plants: an overview

    International Nuclear Information System (INIS)

    Lu, Lixuan; Jiang, Jin

    2004-01-01

    Deregulation in the electricity market has resulted in a number of challenges in the nuclear power industry. Nuclear power plants must find innovative ways to remain competitive by reducing operating costs without jeopardizing safety. Instrumentation and Control (I and C) systems not only play important roles in plant operation, but also in reducing the cost of power generation while maintaining and/or enhancing safety. Therefore, it is extremely important that I and C systems are managed efficiently and economically. With the increasing use of digital technologies, new methods are needed to solve problems associated with various aspects of digital I and C systems. Probabilistic Safety Assessment (PSA) has proved to be an effective method for safety analysis and risk-based decisions, even though challenges are still present. This paper provides an overview of PSA applications in three areas of digital I and C systems in nuclear power plants. These areas are Graded Quality Assurance, Surveillance Testing, and Instrumentation and Control System Design. In addition, PSA application in the regulation of nuclear power plants that adopt digital I and C systems is also investigated. (author)

  20. Problems of probabilistic safety assessment after Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    Sugiyama, Naoki

    2011-01-01

    Probabilistic safety assessment (PSA) methodology to assure nuclear safety is had great expectations of lessons learned from Fukushima Daiichi nuclear power plant (NPP) accident and on the other hand this accident made actualized technical problems of PSA. Effectiveness of current PSA methodology for risk assessment was confirmed by comparing the accident development with accident scenario of PSA and equipment failure rate. From a viewpoint of nuclear safety objective and defense in depth approach of IAEA, technical problems of PSA were (1) extension of PSA for spent fuel pool and waste disposal system as well as level 3PSA for broader environmental contamination and (2) overlapping of accident scenario of plural unit site, balance of high quality plant management and preceding negation, treatment of uncertainty of external events, severe accident measure and human reliability analysis and reflection of disaster prevention capability to level 3PSA. In order to upgrade PSA technology, six proposals were described for nuclear safety and defense in depth, comprehensive evaluation scope and catch-up of latest technology, necessity of strategic preparation of PSA standard, human resources fostering and risk communication. (T. Tanaka)

  1. 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

  2. Probabilistic safety assessment support for the maintenance rule at Duke Power Company

    International Nuclear Information System (INIS)

    Brewer, H. Duncan; Canady, Ken S.

    1999-01-01

    The Nuclear Regulatory Commission (NRC) published the Maintenance Rule on July 10, 1991 with an implementation date of July 10, 1996 . Maintenance rule implementation at the Duke Power Company has used probabilistic safety assessment (PSA) insights to help focus the monitoring of structures, systems and components (SSC) performance and to ensure that maintenance is effectively performed. This paper describes how the probabilistic risk assessment (PRA) group at the Duke Power Company provides support for the maintenance rule by performing the following tasks: (1) providing a member of the expert panel; (2) determining the risk-significant SSCs; (3) establishing SSC performance criteria for availability and reliability; (4) evaluating past performance and its impact on core damage risk as part of the periodic assessment; (5) providing input to the PRA matrix; (6) providing risk analyses of combinations of SSCs out of service; (7) providing support for the SENTINEL program; and (8) providing support for PSA training. These tasks are not simply tied to the initial implementation of the rule. The maintenance rule must be kept consistent with the current design and operation of the plant. This will require that the PRA models and the many PSA calculations performed to support the maintenance rule are kept up-to-date. Therefore, support of the maintenance rule will be one of the primary roles of the PSA group for the remainder of the life of the plant

  3. Evaluation of safety of hypobaric decompressions and EVA from positions of probabilistic theory

    Science.gov (United States)

    Nikolaev, V. P.

    Formation and subsequent evolution of gas bubbles in blood and tissues of subjects exposed to decompression are casual processes in their nature. Such character of bubbling processes in a body predetermines probabilistic character of decompression sickness (DCS) incidence in divers, aviators and astronauts. Our original probabilistic theory of decompression safety is based on stochastic models of these processes and on the concept of critical volume of a free gas phase in body tissues. From positions of this theory, the probability of DCS incidence during single-stage decompressions and during hypobaric decompressions under EVA in particular, is defined by the distribution of possible values of nucleation efficiency in "pain" tissues and by its critical significance depended on the parameters of a concrete decompression. In the present study the following is shown: 1) the dimensionless index of critical nucleation efficiency for "pain" body tissues is a more adequate index of decompression stress in comparison with Tissue Ratio, TR; 2) a priory the decompression under EVA performed according to the Russian protocol is more safe than decompression under EVA performed in accordance with the U.S. protocol; 3) the Russian space suit operated at a higher pressure and having a higher "rigidity" induces a stronger inhibition of mechanisms of cavitation and gas bubbles formation in tissues of a subject located in it, and by that provides a more considerable reduction of the DCS risk during real EVA performance.

  4. Methodology and results of the seismic probabilistic safety assessment of Krsko nuclear power plant

    International Nuclear Information System (INIS)

    Vermaut, M.K.; Monette, P.; Campbell, R.D.

    1995-01-01

    A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Krsko plant. The methodology adopted is the seismic PSA (Probabilistic Safety Assessment). The Krsko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of the site hazard, calculation of plant structures response including soil-structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Also relay chatter analysis and soil stability studies were performed. The seismic PSA described here is limited to the analysis of CDF (level I PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Krsko NPP but are not further described in this paper. The results of the seismic PSA study indicate that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable to that of most US and Western Europe NPPs. (author)

  5. Functional verification of a safety class controller for NPPs using a UVM register Model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu Chull [Dept. of Applied Computer Engineering, Dankook University, Cheonan (Korea, Republic of)

    2014-06-15

    A highly reliable safety class controller for NPPs (Nuclear Power Plants) is mandatory as even a minor malfunction can lead to disastrous consequences for people, the environment or the facility. In order to enhance the reliability of a safety class digital controller for NPPs, we employed a diversity approach, in which a PLC-type controller and a PLD-type controller are to be operated in parallel. We built and used structured testbenches based on the classes supported by UVM for functional verification of the PLD-type controller designed for NPPs. We incorporated a UVM register model into the testbenches in order to increase the controllability and the observability of the DUT(Device Under Test). With the increased testability, we could easily verify the datapaths between I/O ports and the register sets of the DUT, otherwise we had to perform black box tests for the datapaths, which is very cumbersome and time consuming. We were also able to perform constrained random verification very easily and systematically. From the study, we confirmed the various advantages of using the UVM register model in verification such as scalability, reusability and interoperability, and set some design guidelines for verification of the NPP controllers.

  6. 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

  7. A probabilistic analysis method to evaluate the effect of human factors on plant safety

    International Nuclear Information System (INIS)

    Ujita, H.

    1987-01-01

    A method to evaluate the effect of human factors on probabilistic safety analysis (PSA) is developed. The main features of the method are as follows: 1. A time-dependent multibranch tree is constructed to treat time dependency of human error probability. 2. A sensitivity analysis is done to determine uncertainty in the PSA due to branch time of human error occurrence, human error data source, extraneous act probability, and human recovery probability. The method is applied to a large-break, loss-of-coolant accident of a boiling water reactor-5. As a result, core melt probability and risk do not depend on the number of time branches, which means that a small number of branches are sufficient. These values depend on the first branch time and the human error probability

  8. Use of the Safety probabilistic analysis for the risk monitor before maintenance

    International Nuclear Information System (INIS)

    Gonzalez C, M.

    2004-01-01

    In this work the use of the Safety Probabilistic Analysis (APS) of the Laguna Verde Power plant to quantify the risk before maintenance is presented. Beginning to describe the nature of the Rule of Maintenance and their risk evaluations, it is planned about the paper of the APS for that purpose, and a systematic form to establish the reaches for this use open of the model is delineated. The work provides some technique details of the implantation methods of the APS like risk monitor, including the form of introducing the systems, trains and components to the user, as well as the fitness to the models and improvements to the used platform. There are covered some of the measures taken to achieve the objectives of preserving the base model approved, to facilitate the periodic realize, and to achieve acceptable times of execution for their efficient use. (Author)

  9. Procedures for conducting probabilistic safety assessments of nuclear power plants (Level 1)

    International Nuclear Information System (INIS)

    1992-01-01

    This report provides guidance for conducting a Level 1 of probabilistic safety assessment (PSA), that is a PSA concerned with events leading to core damage. The scope of this report is confined to internal initiating events (excluding internal fires and floods). A particular aim is to promote a standardized framework, terminology and form of documentation for PSAs so as to facilitate external review of the results of such studies. The report is divided into the following major sections: management and organization; identification of sources of radioactive releases and accident initiators; accident sequence modelling; data assessment and parameter estimation; accident sequence quantification; documentation of the analysis: display and interpretation of result. 45 refs, 7 figs, 23 tabs

  10. Biasing transition rate method based on direct MC simulation for probabilistic safety assessment

    Institute of Scientific and Technical Information of China (English)

    Xiao-Lei Pan; Jia-Qun Wang; Run Yuan; Fang Wang; Han-Qing Lin; Li-Qin Hu; Jin Wang

    2017-01-01

    Direct Monte Carlo (MC) simulation is a powerful probabilistic safety assessment method for accounting dynamics of the system.But it is not efficient at simulating rare events.A biasing transition rate method based on direct MC simulation is proposed to solve the problem in this paper.This method biases transition rates of the components by adding virtual components to them in series to increase the occurrence probability of the rare event,hence the decrease in the variance of MC estimator.Several cases are used to benchmark this method.The results show that the method is effective at modeling system failure and is more efficient at collecting evidence of rare events than the direct MC simulation.The performance is greatly improved by the biasing transition rate method.

  11. Containment response analysis for the PSA (Probabilistic Safety Assessment) of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.

    1997-01-01

    This work is part of the probabilistic safety assessment actually under development for the CAREM-25 nuclear power station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during severe accidents. Then plan damage states are defined. Furthermore, containment damage states are also defined, and containment event trees are built for each plant damage state. Those sequences considered representative from the annual probability (those which exceed or probability of IE-09 per year, are used to quantify the combinations of plant damage states/containment damage states, based on the estimation of a vulnerability matrix. (author) [es

  12. MASCOT and MOP programs for probabilistic safety assessment. Pt. E. MOP (Version 3A) user guide

    International Nuclear Information System (INIS)

    Agg, P.J.; Hopper, M.J.; Sinclair, J.E.; Sumner, P.J.

    1994-04-01

    MOP is a post-processor for the probabilistic safety assessment program MASCOT, which models the consequences of the disposal of radioactive waste. This document provides a general description of the capabilities of the MOP program, together with a comprehensive guide to the MOP user command language. MOP is able to calculate and present various statistical measures of the modelled radiological consequences, in both printed and graphical form. The results of intermediate analyses can be saved from one MOP job to the next, and this allows MOP to be used as many times as desired to process the results of the same MASCOT job. MOP can work with the quantities passed to it from the MASCOT job or with new quantities, defined and calculated according to individual requirements. This is usually done by transforming the MASCOT quantities using algebraic expressions. (Author)

  13. Application of thermal comfort theory in probabilistic safety assessment of a nuclear power plant

    International Nuclear Information System (INIS)

    Zhou Tao; Sun Canhui; Li Zhenyang; Wang Zenghui

    2011-01-01

    Human factor errors in probabilistic safety assessment (PSA) of a nuclear power plant (NPP) can be prevented using thermal comfort analysis. In this paper, the THERP + HCR model is modified by using PMV (Predicted Mean Vote) and PPD (Predicted Percentage Dissatisfied) index system, so as to obtain the operator cognitive reliability,and to reflect and analyze human perception, thermal comfort status,and cognitive ability in a specific NPP environment. The mechanism of human factors in the PSA is analyzed by operators of skill, rule and knowledge types. The THERP + HCR model modified by thermal comfort theory can reflect the conditions in actual environment, and optimize reliability analysis of human factors. Improving human thermal comfort for different types of operators reduces adverse factors due to human errors, and provides a safe and optimum decision-making for NPPs. (authors)

  14. Human reliability analysis for probabilistic safety assessments - review of methods and issues

    International Nuclear Information System (INIS)

    Srinivas, G.; Guptan, Rajee; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

    2011-01-01

    It is well known that the two major events in World Nuclear Power Plant Operating history, namely the Three Mile Island and Chernobyl, were Human failure events. Subsequent to these two events, several significant changes have been incorporated in Plant Design, Control Room Design and Operator Training to reduce the possibility of Human errors during plant transients. Still, human error contribution to Risk in Nuclear Power Plant operations has been a topic of continued attention for research, development and analysis. Probabilistic Safety Assessments attempt to capture all potential human errors with a scientifically computed failure probability, through Human Reliability Analysis. Several methods are followed by different countries to quantify the Human error probability. This paper reviews the various popular methods being followed, critically examines them with reference to their criticisms and brings out issues for future research. (author)

  15. Results of the probabilistic safety assessment to the cobalt-therapy process

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Ferro, R.; Lozano, B.; De la Fuente Puch, Andres; Dumenigo Gonzalez, Cruz; Troncoso, M.; Perez, Y.; Alemany, J.; Leon, L.; Amador, R.; Lazo, R.; Labrador, F.; Blanco, A.; Betancourt, L.; Crespo, D.; Silvestre, I.

    2004-01-01

    This paper presents the results of the Probabilistic Safety Assessment (PSA) to the Cobalt Therapy Treatment Process in the Oncological Unit of Pinar del Rio city to evaluate occupational, public and medical exposures during cobalt therapy treatment. Equipment's Failures Modes and Human Error were evaluated for each system and treatment stage aimed at obtaining an exhaustive list of the deviations with a reasonable probability to occur and may produce significant adverse outcomes. The lowest exposures probabilities correspond to the public exposures during the treatment process; around 10-10 per year, being the workers exposures around 10-4 per year. Regarding the patient, exposures frequencies vary in dependence of the extent to which the error affect individual treatment, individual patients, or all the patients treated on a specific unit

  16. Procedures for conducting common cause failure analysis in probabilistic safety assessment

    International Nuclear Information System (INIS)

    1992-05-01

    The principal objective of this report is to supplement the procedure developed in Mosleh et al. (1988, 1989) by providing more explicit guidance for a practical approach to common cause failures (CCF) analysis. The detailed CCF analysis following that procedure would be very labour intensive and time consuming. This document identifies a number of options for performing the more labour intensive parts of the analysis in an attempt to achieve a balance between the need for detail, the purpose of the analysis and the resources available. The document is intended to be compatible with the Agency's Procedures for Conducting Probabilistic Safety Assessments for Nuclear Power Plants (IAEA, 1992), but can be regarded as a stand-alone report to be used in conjunction with NUREG/CR-4780 (Mosleh et al., 1988, 1989) to provide additional detail, and discussion of key technical issues

  17. Status, experience and future prospects for the development of probabilistic safety criteria

    International Nuclear Information System (INIS)

    1989-09-01

    During 27-31 January 1986 the IAEA held a Technical Committee Meeting on ''Status, Experience, and Future Prospects for the Development of Probabilistic Safety Criteria''. Participation included representation of essentially all countries with major developments in the area as well as the Nuclear Energy Agency of the OECD and CEC. Though it has to be recognized that in such a short time period it is impossible to resolve or even analyse all aspects of this complex issue, the present situation, the main problems and the directions for future work clearly emerged. This report was prepared by the members of the Technical Committee based on the opinions expressed and on the information available at the time of the meeting. The report also contains 20 papers presented at the meeting by participants. A separate abstract was prepared for each of these 20 papers. Refs, figs and tabs

  18. Comparison of plant-specific probabilistic safety assessments and lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Balfanz, H.P. [TUeV Nord, Hamburg (Germany); Berg, H.P. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany); Steininger, U. [TUeV Energie- und Systemtechnik GmbH, Unternehmensgruppe TUeV Sueddeutschland, Muenchen (Germany)

    2001-11-01

    Probabilistic safety assessments (PSA) have been performed for all German nuclear power plants in operation. These assessments are mainly based on the recent German PSA guide and an earlier draft, respectively. However, comparison of these PSA show differences in the results which are discussed in this paper. Lessons learned from this comparison and further development of the PSA methodology are described. (orig.) [German] Probabilistische Sicherheitsanalysen (PSA) sind fuer alle in Betrieb befindlichen deutschen Kernkraftwerke durchgefuehrt worden. Diese Analysen basierten in der Regel auf dem aktuellen deutschen PSA-Leitfaden bzw. einem frueheren Entwurf. Ein Vergleich dieser PSA zeigt Unterschiede in den Ergebnissen, die in diesem Beitrag diskutiert werden. Erfahrungen und Erkenntnisse, die aus diesem Vergleich abgeleitet werden koennen, und weitere Entwicklungen der PSA-Methoden werden beschrieben. (orig.)

  19. Development of specific data of plant for a safety probabilistic analysis

    International Nuclear Information System (INIS)

    Gonzalez C, M.; Nelson E, P.

    2004-01-01

    In this work the development of specific data of plant is described for the Safety Probabilistic Analysis (APS) of the Laguna Verde Central. The description of those used methods concentrate on the obtention of rates of failure of the equipment and frequencies of initiator events modeled in the APS, making mention to other types of data that also appeal to specific sources of the plant. The method to obtain the rates of failure of the equipment takes advantage the information of failures of components and unavailability of systems obtained entreaty in execution with the Maintenance Rule (1OCFR50.65). The method to develop the frequencies of initiators take in account the registered operational experience as reportable events. In both cases the own experience is combined with published generic data using Bayesian realized techniques. Details are provided about the gathering of information, the confirmations of consistency and adjustment necessities, presenting examples of the obtained results. (Author)

  20. Attributes of Full Scope Level 1 Probabilistic Safety Assessment (PSA) for Applications in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2016-10-01

    This publication supersedes IAEA-TECDOC-1511, Determining the Quality of Probabilistic Safety Assessment (PSA) for Applications in Nuclear Power Plants (published in 2006), which provided detailed information on technical features of a restricted scope PSA aimed at analysing only internal initiating events caused by random component failures and human errors, and accident sequences that may lead to reactor core damage during operation. The present publication extends the scope of the PSA to cover a broader range of internal and external hazards, and low power and shutdown modes of nuclear power plant operation. In addition, some PSA aspects relevant to lessons learned from the accident at the Fukushima Daiichi nuclear power plant are also considered

  1. Probabilistic safety and risk assessments in the field of nuclear technology - Mode of operation, possibilities and limits

    International Nuclear Information System (INIS)

    Mertens, J.

    1993-01-01

    In this study probabilistic safety and risk assessments in the field of nuclear energy are explained. Mainly qualitative results and conclusions are presented. Explanations for often discussed aspects of such analysis reveal the procedure and reasonable limits of application. The mentioned literature contains detailed results. (orig./DG) [de

  2. Comparison of event tree, fault tree and Markov methods for probabilistic safety assessment and application to accident mitigation

    International Nuclear Information System (INIS)

    James, H.; Harris, M.J.; Hall, S.F.

    1992-01-01

    Probabilistic safety assessment (PSA) is used extensively in the nuclear industry. The main stages of PSA and the traditional event tree method are described. Focussing on hydrogen explosions, an event tree model is compared to a novel Markov model and a fault tree, and unexpected implication for accident mitigation is revealed. (author)

  3. Application of the methodology of safety probabilistic analysis to the modelling the emergency feedwater system of Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Troncoso, M.; Oliva, G.

    1993-01-01

    The application of the methodology developed in the framework of the national plan of safety probabilistic analysis (APS) to the emergency feed water system for the failures of small LOCAS and external electrical supply loss in the nuclear power plant is illustrated in this work. The facilities created by the ARCON code to model the systems and its documentation are also expounded

  4. Safety during sea transport of radioactive materials. Probabilistic safety analysis of package fro sea surface fire accident

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi; Obara, Isonori; Akutsu, Yukio; Aritomi, Masanori

    2000-01-01

    The ships carrying irradiated nuclear fuel, plutonium and high level radioactive wastes(INF materials) are designed to keep integrity of packaging based on the various safety and fireproof measures, even if the ship encounters a maritime fire accident. However, granted that the frequency is very low, realistic severe accidents should be evaluated. In this paper, probabilistic safety assessment method is applied to evaluate safety margin for severe sea fire accidents using event tree analysis. Based on our separate studies, the severest scenario was estimated as follows; an INF transport ship collides with oil tanker and induces a sea surface fire. Probability data such as ship's collision, oil leakage, ignition, escape from fire region, operations of cask cooling system and water flooding systems were also introduced from above mentioned studies. The results indicate that the probability of which packages cannot keep their integrity during the sea surface fire accident is very low and sea transport of INF materials is carried out very safely. (author)

  5. 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

  6. The Role of Probabilistic Design Analysis Methods in Safety and Affordability

    Science.gov (United States)

    Safie, Fayssal M.

    2016-01-01

    For the last several years, NASA and its contractors have been working together to build space launch systems to commercialize space. Developing commercial affordable and safe launch systems becomes very important and requires a paradigm shift. This paradigm shift enforces the need for an integrated systems engineering environment where cost, safety, reliability, and performance need to be considered to optimize the launch system design. In such an environment, rule based and deterministic engineering design practices alone may not be sufficient to optimize margins and fault tolerance to reduce cost. As a result, introduction of Probabilistic Design Analysis (PDA) methods to support the current deterministic engineering design practices becomes a necessity to reduce cost without compromising reliability and safety. This paper discusses the importance of PDA methods in NASA's new commercial environment, their applications, and the key role they can play in designing reliable, safe, and affordable launch systems. More specifically, this paper discusses: 1) The involvement of NASA in PDA 2) Why PDA is needed 3) A PDA model structure 4) A PDA example application 5) PDA link to safety and affordability.

  7. Probabilistic Safety Analysis of High Speed and Conventional Lines Using Bayesian Networks

    Energy Technology Data Exchange (ETDEWEB)

    Grande Andrade, Z.; Castillo Ron, E.; O' Connor, A.; Nogal, M.

    2016-07-01

    A Bayesian network approach is presented for probabilistic safety analysis (PSA) of railway lines. The idea consists of identifying and reproducing all the elements that the train encounters when circulating along a railway line, such as light and speed limit signals, tunnel or viaduct entries or exits, cuttings and embankments, acoustic sounds received in the cabin, curves, switches, etc. In addition, since the human error is very relevant for safety evaluation, the automatic train protection (ATP) systems and the driver behavior and its time evolution are modelled and taken into account to determine the probabilities of human errors. The nodes of the Bayesian network, their links and the associated probability tables are automatically constructed based on the line data that need to be carefully given. The conditional probability tables are reproduced by closed formulas, which facilitate the modelling and the sensitivity analysis. A sorted list of the most dangerous elements in the line is obtained, which permits making decisions about the line safety and programming maintenance operations in order to optimize them and reduce the maintenance costs substantially. The proposed methodology is illustrated by its application to several cases that include real lines such as the Palencia-Santander and the Dublin-Belfast lines. (Author)

  8. Probabilistic safety analysis for fire events for the NPP Isar 2

    International Nuclear Information System (INIS)

    Schmaltz, H.; Hristodulidis, A.

    2007-01-01

    The 'Probabilistic Safety Analysis for Fire Events' (Fire-PSA KKI2) for the NPP Isar 2 was performed in addition to the PSA for full power operation and considers all possible events which can be initiated due to a fire. The aim of the plant specific Fire-PSA was to perform a quantitative assessment of fire events during full power operation, which is state of the art. Based on simplistic assumptions referring to the fire induced failures, the influence of system- and component-failures on the frequency of the core damage states was analysed. The Fire-PSA considers events, which will result due to fire-induced failures of equipment on the one hand in a SCRAM and on the other hand in events, which will not have direct operational effects but because of the fire-induced failure of safety related installations the plant will be shut down as a precautionary measure. These events are considered because they may have a not negligible influence on the frequency of core damage states in case of failures during the plant shut down because of the reduced redundancy of safety related systems. (orig.)

  9. Human performance analysis in the frame of probabilistic safety assessment of research reactors

    International Nuclear Information System (INIS)

    Farcasiu, Mita; Nitoi, Mirela; Apostol, Minodora; Turcu, I.; Florescu, Gh.

    2005-01-01

    Full text: The analysis of operating experience has identified the importance of human performance in reliability and safety of research reactors. In Probabilistic Safety Assessment (PSA) of nuclear facilities, human performance analysis (HPA) is used in order to estimate human error contribution to the failure of system components or functions. HPA is a qualitative and quantitative analysis of human actions identified for error-likely situations or accident-prone situations. Qualitative analysis is used to identify all man-machine interfaces that can lead to an accident, types of human interactions which may mitigate or exacerbate the accident, types of human errors and performance shaping factors. Quantitative analysis is used to develop estimates of human error probability as effects of human performance in reliability and safety. The goal of this paper is to accomplish a HPA in the PSA frame for research reactors. Human error probabilities estimated as results of human actions analysis could be included in system event tree and/or system fault tree. The achieved sensitivity analyses determine human performance sensibility at systematically variations both for dependencies level between human actions and for operator stress level. The necessary information was obtained from operating experience of research reactor TRIGA from INR Pitesti. The required data were obtained from generic data bases. (authors)

  10. Probabilistic safety analysis on an SBWR 72 hours after the initiating event

    International Nuclear Information System (INIS)

    Dominguez Bautista, M.T.; Peinador Veira, M.

    1996-01-01

    Passive plants, including SBWRs, are designed to carry out safety functions with passive systems during the first 72 hours after the initiation event with no need for manual actions or external support. After this period, some recovery actions are required to enable the passive systems to continue performing their safety functions. The study was carried out by the INITEC-Empresarios Agrupados Joint Venture within the framework of the international group collaborating with GE on this project. Its purpose has been to assess, by means of probabilistic criteria, the importance to safety of each of these support actions, in order to define possible requirements to be considered in the design in respect of said recovery actions. In brief, the methodology developed for this objective consists of (1) quantifying success event trees from the PSA up to 72 hours, (2) determining the actions required in each sequence to maintain Steady State after 72 hours, (3) identifying available alternative core cooling methods in each sequence, (4) establishing the approximate (order of magnitude) realizability of each alternative method, (5) calculating the frequency of core damage as a function of the failure probability of post-72-hour actions and (6) analysing the importance of post-72-hour actions. The results of this analysis permit the establishment, right from the conceptual design phase, of the requirements that will arise to ensure these actions in the long term, enhancing their reliability and preventing the accident from continuing beyond this period. (Author)

  11. Probabilistic safety assessment based expert systems in support of dynamic risk assessment

    International Nuclear Information System (INIS)

    Varde, P.V.; Sharma, U.L.; Marik, S.K.; Raina, V.K.; Tikku, A.C.

    2006-01-01

    Probabilistic Safety Assessment (PSA) studies are being performed, world over as part of integrated risk assessment for Nuclear Power Plants and in many cases PSA insight is utilized in support of decision making. Though the modern plants are built with inherent safety provisions, particularly to reduce the supervisory requirements during initial period into the accident, it is always desired to develop an efficient user friendly real-time operator advisory system for handling of plant transients/emergencies which would be of immense benefit for the enhancement of operational safety of the plant. This paper discusses an integrated approach for the development of operator support system. In this approach, PSA methodology and the insight obtained from PSA has been utilized for development of knowledge based or rule based experts system. While Artificial Neural Network (ANN) approach has been employed for transient identification, rule-base expert system shell environment was used for the development of diagnostic module in this system. Attempt has been made to demonstrate that this approach offers an efficient framework for addressing requirements related to handling of real-time/dynamic scenario. (author)

  12. Application of probabilistic methods to safety R and D and design choices

    International Nuclear Information System (INIS)

    Gavigan, F.X.; Griffith, J.D.

    1977-01-01

    The Liquid Metal Fast Breeder Reactor (LMFBR) safety program is committed to identifying and exploiting areas in which probabilistic methods can be developed and used in making reactor safety R and D choices and optimizing designs of safety systems. Emphasis will be placed on a positive approach of solidifying and expanding our knowledge. This will provide the groundwork for a consensus on FBR risk. The management structure which will be used is based on a mechanistic approach to an LMFBR Core Disruptive Accident (CDA) with risk partitioned into ''Lines of Assurance,'' i.e., independent, phenomenologically-based barriers which will impede or mitigate the progression and consequences of accident sequences. Quantitative determination of the probability of breach of these barriers through the completion of work identified for each Line of Assurance will allow the quantification of the contribution to risk reduction associated with the success of each barrier. This process can lead to better use of resources by channeling R and D in directions which promise the greatest potential for reducing risk and by identifying an orderly approach to the development and demonstration of design features which will keep LMFBR risks at an acceptable level

  13. Probabilistic safety assessment for instrumentation and control systems in nuclear power plants. A literature survey

    International Nuclear Information System (INIS)

    Lu, Lixuan; Jiang, Jin

    2003-01-01

    Deregulation in electricity market will create a great deal of challenges for Nuclear Power Plants (NPP). To stay competitive, NPP will need to find new ways to reduce their operation costs. In NPP, Instrumentation and Control (I and C) systems play an important role in reducing the cost of producing electricity while maintaining and/or enhancing safety. Therefore, it is extremely important that one should manage the I and C systems more efficiently and economically. Meanwhile, obsolescence problem associated with I and C systems encouraged the usage of advanced digital techniques in I and C systems. Thus, new methodologies are needed to analyze the reliability and determine the maintenance strategy for the digital I and C systems. Probabilistic Safety Assessment (PSA) has been probed to be a promising method to deal with this issue. This paper provides a literature survey on the development of digital I and C systems in NPP, followed by a detailed review of PSA including its benefits, limitations and the future direction of its development. Most importantly, potential applications of PSA in various aspects of I and C systems are brought into perspective throughout the paper. Furthermore, the applicability of PSA in the regulation of safety-related I and C systems is demonstrated. Detailed information on PSA applications in 1) the resource allocation for I and C systems: 2) the determination of surveillance testing strategies; and 3) I and C system designs, is provided. (author)

  14. Probabilistic analysis of safety of a production plant of hydrogen using nuclear energy

    International Nuclear Information System (INIS)

    Flores F, A.; Nelson E, P.F.; Francois L, J.L.

    2005-01-01

    The present work makes use of the Probabilistic Safety analysis to evaluate and to quantify the safety in a plant producer of hydrogen coupled to a nuclear reactor of high temperature, the one which is building in Japan. It is had the description of systems and devices of the HTTR, the pipe diagrams and instrumentation of the plant, as well as the rates of generic faults for the components of the plant. The first step was to carry out a HAZOP study (Hazard and Operability Study) with the purpose of obtaining the initiator events; once obtained these, it was developed a tree of events by each initiator event and for each system it was developed a fault tree; the data used for the quantification of the failure probability of the systems were obtained starting from several generic sources of information. In each tree of events different final states were obtained and it stops each one, their occurrence frequency. The construction and evaluation of the tree of events and of failures one carries out with the SAPHIRE program. The results show the safety of the shutdown system of the HTTR and they allow to suggest modifications to the auxiliary system of refrigeration and to the heat exchanger helium/water pressurized. (Author)

  15. 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)

  16. Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

    Energy Technology Data Exchange (ETDEWEB)

    R.M. Dolin

    2003-03-03

    This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

  17. Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

    International Nuclear Information System (INIS)

    Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

    2005-01-01

    This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

  18. What price safety. A probabilistic cost-benefit evaluaton of existing engineered safety features

    International Nuclear Information System (INIS)

    O'Donnell, E.P.

    1978-01-01

    The paper provides a method for performing quantitative cost-benefit evaluations for nuclear safety concerns involving accidents of low probability and potentially large consequences. It presents an application of the method to ECCS, containment, emergency power system and hydrogen recombiner system. This evaluation provides a valuable assessment of the relative cost effectiveness of these features in reducing accident risk. It also provides insight into the sensitivity of cost-benefit calculations to the manner in which safety features are sequantially added in design. (author)

  19. Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

    1994-01-01

    Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

  20. [Verification of bacteriological safety of PCM 40 air conditioner].

    Science.gov (United States)

    Dumas, J L; Ducel, G; Rouge, J C

    1991-01-01

    This study assessed the bacteriological safety of the bedside air conditioner PCM 40 (Howorth Airtech), used for prevention of intraoperative hypothermia, by blowing filtered warm air through a special mattress. The 3 microns bacterial filter of the device released 2,968 +/- 5,618 particles of diameter less than 3 microns per m3 of room air, containing 78,798 +/- 37,243 of such particles per m3. The amount of bacteries in the air pulsed from the mattress was 30 +/- 41 cfu/m3 vs 120 cfu/m3 in the ambient air and in the hot air supply tubing it reached 6 +/- 5 cfu/m3 vs 175 +/- 77 cfu/m3. It is concluded that bacteriological data do not contra-indicate the use of this air conditioner in the operating theater. The only limitations for use are the position (prone or lateral position) and type of surgery (neurosurgery).

  1. Radiological Operational Safety Verification for LILW Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology, SNU, Seoul (Korea, Republic of); Jeong, Seung Young; Kim, Byung Soo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    The successful implementation of radioactive waste repository program depends on scientific and technical aspects of excellent safety strategy as well as on societal aspects such as stakeholder acceptance and confidence. Monitoring is considered as key element in serving both ends. It covers all stages of the disposal process from site selection to institutional monitoring after the repository is closed. Basically, the purpose of the monitoring of radioactive waste disposal facility is not to reveal any increase of radioactivity due to the repository, but to provide reassurance and confirmation that the repository is fulfilling its passive safety purpose as an initial disposal concept and that long-term safety driven by regulatory requirements is ensured throughout the entire lifetime of disposal facility including post-closure phase. Five principal objectives of monitoring of geological disposal are summarized by IAEA-TECDOC-1208 as follows 1) Supporting management decisions in a staged programme of repository development: 2) Strengthening understanding of system behavior: 3) Societal decision making: 4) Accumulating an environmental database: 5) Nuclear safeguards (if repository contains fissile material, i.e., spent fuel or plutonium-rich waste) Based on the results of detailed studies of the above objectives and related phenomena, 6 categories of potential monitoring parameters are determined as follows: (1) degradation of repository structures, (2) behavior of the waste package and its associated buffer material, (3) near field chemical interactions between introduced materials, groundwater and host rock, (4) chemical and physical changes to the surrounding geosphere, (5) provision of an environmental database, and (6) nuclear safeguards. Typical monitoring parameters include temperature (heat), water level, pore-water and moisture content (groundwater), rock pressure, fractures, displacement and deformation (stress), water quality chemistry and dissolved

  2. Improving operational safety management through probabilistic safety assessment on personal computers

    International Nuclear Information System (INIS)

    1988-10-01

    The Technical Committee Meeting considered the current effort in the implementation and use of PSA information for day-to-day operational safety management on Personal Computers. Due to the very recent development of the necessary hardware and software for Personal Computers, the application of PSA information for day-to-day operational safety management on PCs is essentially still in a pioneering stage. There is at present only one such system for end users existing, the PRISIM (Plant Risk Status Information Management) program for which a limited practical application experience is available. Others are still in the development stage. The main aim of the Technical Committee Meeting was to discuss the present status of PSA based systems for operational safety management support on small computers, to consider practical aspects when implementing these systems into a nuclear installation and to address problems related to the further work in the area. A separate abstract was prepared for the summary of the Technical Committee Meeting and for the 8 papers presented by the participants. Refs, figs and tabs

  3. Development of a computational database for application in Probabilistic Safety Analysis of nuclear research reactors

    International Nuclear Information System (INIS)

    Macedo, Vagner dos Santos

    2016-01-01

    The objective of this work is to present the computational database that was developed to store technical information and process data on component operation, failure and maintenance for the nuclear research reactors located at the Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares, IPEN), in São Paulo, Brazil. Data extracted from this database may be applied in the Probabilistic Safety Analysis of these research reactors or in less complex quantitative assessments related to safety, reliability, availability and maintainability of these facilities. This database may be accessed by users of the corporate network, named IPEN intranet. Professionals who require the access to the database must be duly registered by the system administrator, so that they will be able to consult and handle the information. The logical model adopted to represent the database structure is an entity-relationship model, which is in accordance with the protocols installed in IPEN intranet. The open-source relational database management system called MySQL, which is based on the Structured Query Language (SQL), was used in the development of this work. The PHP programming language was adopted to allow users to handle the database. Finally, the main result of this work was the creation a web application for the component reliability database named PSADB, specifically developed for the research reactors of IPEN; furthermore, the database management system provides relevant information efficiently. (author)

  4. Reliability data update using condition monitoring and prognostics in probabilistic safety assessment

    Directory of Open Access Journals (Sweden)

    Hyeonmin Kim

    2015-03-01

    Full Text Available Probabilistic safety assessment (PSA has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs. Even though the nature of PSA seeks realistic results, there are still “conservative” aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP, utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs and probability of basic events (BEs. Two illustrative examples will be introduced: (1 how the failure probability of a passive system can be evaluated under different plant conditions and (2 how the IE frequency for a steam generator tube rupture (SGTR can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF depending on operational conditions.

  5. PROSA PRObabilistic Safety Assessment: Dutch summary of the ECN/RIVM/RGD final report

    International Nuclear Information System (INIS)

    Prij, J.; Laheij, G.M.H.; Oostrom, M.; Van Rheenen, W.; Uffink, G.J.M.; Uijt de Haag, P.; Wildenborg, A.F.B.

    1994-05-01

    In the PROSA project the safety of radioactive waste in salt caverns is investigated systematically. PROSA is carried out within the framework of the phase 1A program of the Committee Land Storage (OPLA, abbreviated in Dutch) and is a follow-up of the safety study VEOS. PROSA is focused on improving some aspects of VEOS, in particular the systematic selection of scenarios and determining and calculating the uncertainties. For the scenario selection a system has been developed that takes into account the multi-barrier system and all the possible FEPs (features, events and processes). As a result of the method 22 scenarios were identified. For seven scenarios the radiological consequences have been analyzed by means of a computer model that differs from the model, applied in the VEOS study. The parameters, necessary for the analyses are determined by means of the sources VEOS, PAGIS and PACOMA. The stochastic parameters for the groundwater compartment are calculated with MiniBIOS analyses. Probabilistic calculations were made for the subrosion scenarios, and deterministic calculations are made for the water intrusion scenarios. Of the human intrusion scenarios it appeared that the calculated risk is much lower than has been calculated in VEOS. From the calculated results of the sensitivity and uncertainty analysis it appeared that there is a very large distribution of risks. 10 figs., 10 tabs

  6. RA-6 reactor's probabilistic safety evaluation. Identification and selection of starting events

    International Nuclear Information System (INIS)

    Kay, J.; Chiossi, C.; Felizia, E.; Vallerga, H.; Kalejman, G.; Navarro, R.; Caruso, G.J.

    1987-01-01

    A summary of the 'Identification and selection of starting events' stage of the previous probabilistic safety evaluation of RA-6 reactor is presented. This evaluation was performed to verify if the safety criteria required for the licensing of RA-6 are met and to promote the diffusion of its meaning and usefulness with educational purposes. At this stage the starting events of RA-6 are determined and the probability that such events occur is calculated. The identification and selection of starting events is performed in two steps: determination of proposed starting events and determination of postulated starting events. The proposed starting events are determined by means of the master logic diagram (MLD) method, while the postulated starting events are obtained by grouping the proposed starting events. The simplifying hypothesis required for the application of MLD to the reactor are also formulated. The probability that the proposed and postulated starting events occur is afterwards calculated, adopting different fault models, in accordance with the nature of events that are considered. Conservative hypothesis on the characteristics of these events and the uncertainty of parameter values of those models are also formulated. The numerical values of the above mentioned probabilities are obtained by giving the parameters suitable values that are extracted from specialized publications. (Author)

  7. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    Energy Technology Data Exchange (ETDEWEB)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

    1994-05-01

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

  8. Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

    International Nuclear Information System (INIS)

    Torabi, T.; Bettig, R.; Iliescu, P.; Robinson, J.; Santamaura, P.; Skorupska, B.; Tyagi, A.K.; Vencel, I.

    2013-01-01

    The Enhanced CANDU 6(EC6) is a 700 MWe reactor, which has evolved from the well-established CANDU line of reactors, which are heavy-water moderated, and heavy-water cooled horizontal pressure tube reactors, using natural uranium fuel. The EC6 design retains the generic CANDU design features, while incorporating innovations and state-of-the-art technologies to ensure competitiveness with other design with respect to operation, performance and economics. A design assist probabilistic safety assessment (PSA) was conducted during the design change phase of the project. The purpose of the assessment was to assess internal events during at-power operation and identify the design improvements and additional features needed to comply with the latest regulatory requirements in Canada and compete with other reactor designs, internationally. The PSA results show that the EC6 plant response to the postulated initiating events is well balanced, and the design meets its safety objectives. This paper summarizes the results and insights gained during the development of the PSA models for at-power internal events. (author)

  9. Reliability data update using condition monitoring and prognostics in probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeon Min; Lee, Sang Hwan; Park, Jun Seok; Kim, Hyung Dae; Chang, Yoon Suk; Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2015-03-15

    Probabilistic safety assessment (PSA) has had a significant role in quantitative decision making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still 'conservative' aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

  10. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W.; Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J.

    1994-05-01

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases (open-quotes burpsclose quotes) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity

  11. Application of Integrated Verification Approach to FPGA-based Safety-Critical I and C System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Ibrahim; Heo, Gyunyoung [Kyunghee Univ., Yongin (Korea, Republic of); Jung, Jaecheon [KEPCO, Ulsan (Korea, Republic of)

    2016-10-15

    Safety-critical instrumentation and control (I and C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. Generally in FPGA design verification, the designers make use of verification techniques by writing the test benches which involved various stages of verification activities of register-transfer level (RTL), gate-level, and place and route. Writing the test benches is considerably time consuming and require a lot of efforts to achieve a satisfied desire results. Furthermore, performing the verification at each stage is a major bottleneck and demanded much activities and time. In addition, verification is conceivably, the most difficult and complicated aspect of any design. Therefore, in view of these, this work applied an integrated verification approach to the verification of FPGA-based I and C system in NPP that simultaneously verified the whole design modules using MATLAB/Simulink HDL Co-simulation models. Verification is conceivably, the most difficult and complicated aspect of any design, and an FPGA design is not an exception. Therefore, in this work, we introduced and discussed how an application of integrated verification technique to the verification and testing of FPGA-based I and C system design in NPP can facilitate the verification processes, and verify the entire design modules of the system simultaneously using MATLAB/Simulink HDL co-simulation models. In conclusion, the results showed that, the integrated verification approach through MATLAB/Simulink models, if applied to any design to be verified, could speed up the design verification and reduce the V and V tasks.

  12. Application of Integrated Verification Approach to FPGA-based Safety-Critical I and C System of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ahmed, Ibrahim; Heo, Gyunyoung; Jung, Jaecheon

    2016-01-01

    Safety-critical instrumentation and control (I and C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. Generally in FPGA design verification, the designers make use of verification techniques by writing the test benches which involved various stages of verification activities of register-transfer level (RTL), gate-level, and place and route. Writing the test benches is considerably time consuming and require a lot of efforts to achieve a satisfied desire results. Furthermore, performing the verification at each stage is a major bottleneck and demanded much activities and time. In addition, verification is conceivably, the most difficult and complicated aspect of any design. Therefore, in view of these, this work applied an integrated verification approach to the verification of FPGA-based I and C system in NPP that simultaneously verified the whole design modules using MATLAB/Simulink HDL Co-simulation models. Verification is conceivably, the most difficult and complicated aspect of any design, and an FPGA design is not an exception. Therefore, in this work, we introduced and discussed how an application of integrated verification technique to the verification and testing of FPGA-based I and C system design in NPP can facilitate the verification processes, and verify the entire design modules of the system simultaneously using MATLAB/Simulink HDL co-simulation models. In conclusion, the results showed that, the integrated verification approach through MATLAB/Simulink models, if applied to any design to be verified, could speed up the design verification and reduce the V and V tasks

  13. Probabilistic Safety Assessment: An Effective Tool to Support “Systemic Approach” to Nuclear Safety and Analysis of Human and Organizational Aspects

    International Nuclear Information System (INIS)

    Kuzmina, I.

    2016-01-01

    The Probabilistic Safety Assessment (PSA) represents a comprehensive conceptual and analytical tool for quantitative evaluation of risk of undesirable consequences from nuclear facilities and drawing on qualitative insights for nuclear safety. PSA considers various technical, human, and organizational factors in an integral manner thus explicitly pursuing a true ‘systemic approach’ to safety and enabling holistic insights for further safety improvement. Human Reliability Analysis (HRA) is one of the major tasks within PSA. The poster paper provides an overview of the objectives and scope of PSA and HRA and discusses on further needs in the area of HRA. (author)

  14. Probabilistic safety analysis for nuclear fuel cycle facilities, an exemplary application for a fuel fabrication plant

    International Nuclear Information System (INIS)

    Gmal, B.; Gaenssmantel, G.; Mayer, G.; Moser, E.F.

    2013-01-01

    In order to assess the risk of complex technical systems, the application of the Probabilistic Safety Assessment (PSA) in addition to the Deterministic Safety Analysis becomes of increasing interest. Besides nuclear installations this applies to e. g. chemical plants. A PSA is capable of expanding the basis for the risk assessment and of complementing the conventional deterministic analysis, by which means the existing safety standards of that facility can be improved if necessary. In the available paper, the differences between a PSA for a nuclear power plant and a nuclear fuel cycle facility (NFCF) are discussed in shortness and a basic concept for a PSA for a nuclear fuel cycle facility is described. Furthermore, an exemplary PSA for a partial process in a fuel assembly fabrication facility is described. The underlying data are partially taken from an older German facility, other parts are generic. Moreover, a selected set of reported events corresponding to this partial process is taken as auxiliary data. The investigation of this partial process from the fuel fabrication as an example application shows that PSA methods are in principle applicable to nuclear fuel cycle facilities. Here, the focus is on preventing an initiating event, so that the system analysis is directed to the modeling of fault trees for initiating events. The quantitative results of this exemplary study are given as point values for the average occurrence frequencies. They include large uncertainties because of the limited documentation and data basis available, and thus have only methodological character. While quantitative results are given, further detailed information on process components and process flow is strongly required for robust conclusions with respect to the real process. (authors)

  15. Treatment of external hazards in probabilistic safety assessment for nuclear power plants. A Safety Practice

    International Nuclear Information System (INIS)

    1995-01-01

    This Safety Practice provides guidance on conducting a PSA for external hazards in nuclear power plants. Emphasis is placed on the procedural steps of the PSA rather then on the details of corresponding methods. The publication is intended to assist technical persons managing or performing PSAs. A particular aim is to promote a standardized framework, terminology and form of documentation for external hazards PSA so as to facilitate external review of the results of such studies. For those specialists who are already involved in related studies, such as a Level 1 PSA dealing only with internal events, this publication provides an indication of how additional external events could be integrated into an existing PSA. 16 refs, 5 figs, 2 tabs

  16. Spent fuel verification options for final repository safeguards in Finland. A study on verification methods, their feasibility and safety aspects

    International Nuclear Information System (INIS)

    Hautamaeki, J.; Tiitta, A.

    2000-12-01

    The verification possibilities of the spent fuel assemblies from the Olkiluoto and Loviisa NPPs and the fuel rods from the research reactor of VTT are contemplated in this report. The spent fuel assemblies have to be verified at the partial defect level before the final disposal into the geologic repository. The rods from the research reactor may be verified at the gross defect level. Developing a measurement system for partial defect verification is a complicated and time-consuming task. The Passive High Energy Gamma Emission Tomography and the Fork Detector combined with Gamma Spectrometry are the most potential measurement principles to be developed for this purpose. The whole verification process has to be planned to be as slick as possible. An early start in the planning of the verification and developing the measurement devices is important in order to enable a smooth integration of the verification measurements into the conditioning and disposal process. The IAEA and Euratom have not yet concluded the safeguards criteria for the final disposal. E.g. criteria connected to the selection of the best place to perform the verification. Measurements have not yet been concluded. Options for the verification places have been considered in this report. One option for a verification measurement place is the intermediate storage. The other option is the encapsulation plant. Crucial viewpoints are such as which one offers the best practical possibilities to perform the measurements effectively and which would be the better place in the safeguards point of view. Verification measurements may be needed both in the intermediate storages and in the encapsulation plant. In this report also the integrity of the fuel assemblies after wet intermediate storage period is assessed, because the assemblies have to stand the handling operations of the verification measurements. (orig.)

  17. Method and practice on safety software verification and validation for digital reactor protection system

    International Nuclear Information System (INIS)

    Li Duo; Zhang Liangju; Feng Junting

    2010-01-01

    The key issue arising from digitalization of reactor protection system for Nuclear Power Plant (NPP) is in essence, how to carry out Verification and Validation (V and V), to demonstrate and confirm the software is reliable enough to perform reactor safety functions. Among others the most important activity of software V and V process is unit testing. This paper discusses the basic concepts on safety software V and V and the appropriate technique for software unit testing, focusing on such aspects as how to ensure test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed herein was successfully used in the work of unit testing on safety software of a digital reactor protection system. (author)

  18. Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Ebisawa, K.; Abe, K.; Muramatsu, K.; Itoh, M.; Kohno, K.; Tanaka, T.

    1994-01-01

    This paper presents a method for evaluating 'response factors' of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design resonse to actual response. This method has the following characteristic features: (1) The components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups. (orig.)

  19. Study on quantification method based on Monte Carlo sampling for multiunit probabilistic safety assessment models

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kye Min [KHNP Central Research Institute, Daejeon (Korea, Republic of); Han, Sang Hoon; Park, Jin Hee; Lim, Ho Gon; Yang, Joon Yang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of)

    2017-06-15

    In Korea, many nuclear power plants operate at a single site based on geographical characteristics, but the population density near the sites is higher than that in other countries. Thus, multiunit accidents are a more important consideration than in other countries and should be addressed appropriately. Currently, there are many issues related to a multiunit probabilistic safety assessment (PSA). One of them is the quantification of a multiunit PSA model. A traditional PSA uses a Boolean manipulation of the fault tree in terms of the minimal cut set. However, such methods have some limitations when rare event approximations cannot be used effectively or a very small truncation limit should be applied to identify accident sequence combinations for a multiunit site. In particular, it is well known that seismic risk in terms of core damage frequency can be overestimated because there are many events that have a high failure probability. In this study, we propose a quantification method based on a Monte Carlo approach for a multiunit PSA model. This method can consider all possible accident sequence combinations in a multiunit site and calculate a more exact value for events that have a high failure probability. An example model for six identical units at a site was also developed and quantified to confirm the applicability of the proposed method.

  20. PSAPACK 4.2. A code for probabilistic safety assessment level 1. User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    Only limited use has been made until now of the large amount of information contained in probabilistic safety assessments (PSAs). This is mainly due to the complexity of the PSA reports and the difficulties in obtaining intermediate results and in performing updates and recalculations. Moreover, PSA software was developed for mainframe computers, and the files of information such as fault trees and accident sequences were intended for the use of the analysts carrying out PSA studies or other skilled PSA practitioners. The increasing power and availability of personal computers (PCs) and developments in recent years in both hardware and software have made it possible to develop PSA software for use in PCs. Furthermore, the operational characteristics of PCs make them attractive not only for performing PSAs but also for updating the results and in using them in day-to-day applications. The IAEA has therefore developed in co-operation with its Member States, a software package (PSAPACK) for PCs for use in performing a Level 1 PSA and for easy interrogation of the results. Figs.

  1. Common problems in the elicitation and analysis of expert opinion affecting probabilistic safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M.A.; Booker, J.M.

    1990-01-01

    Expert opinion is frequently used in probabilistic safety assessment (PSA), particularly in estimating low probability events. In this paper, we discuss some of the common problems encountered in eliciting and analyzing expert opinion data and offer solutions or recommendations. The problems are: that experts are not naturally Bayesian. People fail to update their existing information to account for new information as it becomes available, as would be predicted by the Bayesian philosophy; that experts cannot be fully calibrated. To calibrate experts, the feedback from the known quantities must be immediate, frequent, and specific to the task; that experts are limited in the number of things that they can mentally juggle at a time to 7 {plus minus} 2; that data gatherers and analysts can introduce bias by unintentionally causing an altering of the expert's thinking or answers; that the level of detail the data, or granularity, can affect the analyses; and the conditioning effect poses difficulties in gathering and analyzing of the expert data. The data that the expert gives can be conditioned on a variety of factors that can affect the analysis and the interpretation of the results. 31 refs.

  2. Incidents in nuclear research reactor examined by deterministic probability and probabilistic safety analysis

    International Nuclear Information System (INIS)

    Lopes, Valdir Maciel

    2010-01-01

    This study aims to evaluate the potential risks submitted by the incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency, IAEA, were used, the Incident Report System for Research Reactor and Research Reactor Data Base. For this type of assessment was used the Probabilistic Safety Analysis (PSA), within a confidence level of 90% and the Deterministic Probability Analysis (DPA). To obtain the results of calculations of probabilities for PSA, were used the theory and equations in the paper IAEA TECDOC - 636. The development of the calculations of probabilities for PSA was used the program Scilab version 5.1.1, free access, executable on Windows and Linux platforms. A specific program to get the results of probability was developed within the main program Scilab 5.1.1., for two distributions Fischer and Chi-square, both with the confidence level of 90%. Using the Sordi equations and Origin 6.0 program, were obtained the maximum admissible doses related to satisfy the risk limits established by the International Commission on Radiological Protection, ICRP, and were also obtained these maximum doses graphically (figure 1) resulting from the calculations of probabilities x maximum admissible doses. It was found that the reliability of the results of probability is related to the operational experience (reactor x year and fractions) and that the larger it is, greater the confidence in the outcome. Finally, a suggested list of future work to complement this paper was gathered. (author)

  3. PSAPACK 4.2. A code for probabilistic safety assessment level 1. User's manual

    International Nuclear Information System (INIS)

    1995-01-01

    Only limited use has been made until now of the large amount of information contained in probabilistic safety assessments (PSAs). This is mainly due to the complexity of the PSA reports and the difficulties in obtaining intermediate results and in performing updates and recalculations. Moreover, PSA software was developed for mainframe computers, and the files of information such as fault trees and accident sequences were intended for the use of the analysts carrying out PSA studies or other skilled PSA practitioners. The increasing power and availability of personal computers (PCs) and developments in recent years in both hardware and software have made it possible to develop PSA software for use in PCs. Furthermore, the operational characteristics of PCs make them attractive not only for performing PSAs but also for updating the results and in using them in day-to-day applications. The IAEA has therefore developed in co-operation with its Member States, a software package (PSAPACK) for PCs for use in performing a Level 1 PSA and for easy interrogation of the results. Figs

  4. Wind Power in Mexico: Simulation of a Wind Farm and Application of Probabilistic Safety Analysis

    Directory of Open Access Journals (Sweden)

    C. Martín del Campo–Márquez

    2009-10-01

    Full Text Available The most important aspects of wind energy in Mexico, including the potential for generating electricity and the major projects planned are presented here. Inparticular, the generation costs are compared to those of other energy sources. The results from the simulation of a 100 MWwind farm in the Tehuantepec Isthmus are also presented. In addition, the environmental impacts related to the wind farm in the mentioned zone are analyzed. Finally, some benefits of using Probabilistic Safety Analysis are discussed with respect to evaluating the risks associated with events that can occur in wind parks, being especially useful for design and maintenance of the parks and the wind turbines themselves. In particular, an event tree was developed to analyze possible accident sequences that could occur when the wind speed is too great. Also, fault trees were developed for each mitigating system considered, in order to determine the relative importance of the wind generator components to the failure sequences, in order to evaluate the yield of suggested improvements and the optimization of maintenance programs.

  5. Development of a computational database for probabilistic safety assessment of nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Vagner S.; Oliveira, Patricia S. Pagetti de; Andrade, Delvonei Alves de, E-mail: vagner.macedo@usp.br, E-mail: patricia@ipen.br, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The objective of this work is to describe the database being developed at IPEN - CNEN / SP for application in the Probabilistic Safety Assessment of nuclear research reactors. The database can be accessed by means of a computational program installed in the corporate computer network, named IPEN Intranet, and this access will be allowed only to professionals previously registered. Data updating, editing and searching tasks will be controlled by a system administrator according to IPEN Intranet security rules. The logical model and the physical structure of the database can be represented by an Entity Relationship Model, which is based on the operational routines performed by IPEN - CNEN / SP users. The web application designed for the management of the database is named PSADB. It is being developed with MySQL database software and PHP programming language is being used. Data stored in this database are divided into modules that refer to technical specifications, operating history, maintenance history and failure events associated with the main components of the nuclear facilities. (author)

  6. A min cut-set-wise truncation procedure for importance measures computation in probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Duflot, Nicolas [Universite de technologie de Troyes, Institut Charles Delaunay/LM2S, FRE CNRS 2848, 12, rue Marie Curie, BP2060, F-10010 Troyes cedex (France)], E-mail: nicolas.duflot@areva.com; Berenguer, Christophe [Universite de technologie de Troyes, Institut Charles Delaunay/LM2S, FRE CNRS 2848, 12, rue Marie Curie, BP2060, F-10010 Troyes cedex (France)], E-mail: christophe.berenguer@utt.fr; Dieulle, Laurence [Universite de technologie de Troyes, Institut Charles Delaunay/LM2S, FRE CNRS 2848, 12, rue Marie Curie, BP2060, F-10010 Troyes cedex (France)], E-mail: laurence.dieulle@utt.fr; Vasseur, Dominique [EPSNA Group (Nuclear PSA and Application), EDF Research and Development, 1, avenue du Gal de Gaulle, 92141 Clamart cedex (France)], E-mail: dominique.vasseur@edf.fr

    2009-11-15

    A truncation process aims to determine among the set of minimal cut-sets (MCS) produced by a probabilistic safety assessment (PSA) model which of them are significant. Several truncation processes have been proposed for the evaluation of the probability of core damage ensuring a fixed accuracy level. However, the evaluation of new risk indicators as importance measures requires to re-examine the truncation process in order to ensure that the produced estimates will be accurate enough. In this paper a new truncation process is developed permitting to estimate from a single set of MCS the importance measure of any basic event with the desired accuracy level. The main contribution of this new method is to propose an MCS-wise truncation criterion involving two thresholds: an absolute threshold in addition to a new relative threshold concerning the potential probability of the MCS of interest. The method has been tested on a complete level 1 PSA model of a 900 MWe NPP developed by 'Electricite de France' (EDF) and the results presented in this paper indicate that to reach the same accuracy level the proposed method produces a set of MCS whose size is significantly reduced.

  7. A min cut-set-wise truncation procedure for importance measures computation in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Duflot, Nicolas; Berenguer, Christophe; Dieulle, Laurence; Vasseur, Dominique

    2009-01-01

    A truncation process aims to determine among the set of minimal cut-sets (MCS) produced by a probabilistic safety assessment (PSA) model which of them are significant. Several truncation processes have been proposed for the evaluation of the probability of core damage ensuring a fixed accuracy level. However, the evaluation of new risk indicators as importance measures requires to re-examine the truncation process in order to ensure that the produced estimates will be accurate enough. In this paper a new truncation process is developed permitting to estimate from a single set of MCS the importance measure of any basic event with the desired accuracy level. The main contribution of this new method is to propose an MCS-wise truncation criterion involving two thresholds: an absolute threshold in addition to a new relative threshold concerning the potential probability of the MCS of interest. The method has been tested on a complete level 1 PSA model of a 900 MWe NPP developed by 'Electricite de France' (EDF) and the results presented in this paper indicate that to reach the same accuracy level the proposed method produces a set of MCS whose size is significantly reduced.

  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. Development of a computational database for probabilistic safety assessment of nuclear research reactors

    International Nuclear Information System (INIS)

    Macedo, Vagner S.; Oliveira, Patricia S. Pagetti de; Andrade, Delvonei Alves de

    2015-01-01

    The objective of this work is to describe the database being developed at IPEN - CNEN / SP for application in the Probabilistic Safety Assessment of nuclear research reactors. The database can be accessed by means of a computational program installed in the corporate computer network, named IPEN Intranet, and this access will be allowed only to professionals previously registered. Data updating, editing and searching tasks will be controlled by a system administrator according to IPEN Intranet security rules. The logical model and the physical structure of the database can be represented by an Entity Relationship Model, which is based on the operational routines performed by IPEN - CNEN / SP users. The web application designed for the management of the database is named PSADB. It is being developed with MySQL database software and PHP programming language is being used. Data stored in this database are divided into modules that refer to technical specifications, operating history, maintenance history and failure events associated with the main components of the nuclear facilities. (author)

  10. Probabilistic earthquake risk assessment as a tool to improve safety and explanatory adequacy

    International Nuclear Information System (INIS)

    Itoi, Tatsuya

    2015-01-01

    This paper explains the concept of probabilistic earthquake risk assessment, mainly from the viewpoint as a tool to improve safety and explanatory adequacy. The definition of risk is the expected value of undesirable effect in an engineering meaning that is likely to occur in the future, and it is defined in risk management as the triplet of scenario (what can happen), frequency, and impact. As for the earthquake risk assessment of a nuclear power plant, the fragility of structure / system / component (SSC) against earthquake (so-called earthquake fragility) is assessed, and by combining with the earthquake hazard that has been separately obtained, the occurrence frequency and impact of the accident are obtained. From the view of the authors, earthquake risk assessment is for the purpose of decision-making, and is not intended to calculate the probability in a scientifically rigorous manner. For ensuring the quality of risk assessment, the table of 'Expert utilization standards for the evaluation of epistemological uncertainty' is used. Sole quantitative risk assessment is not necessarily sufficient for risk management. It would be important to find how to build the 'framework for comprehensive decision-making.' (A.O.)

  11. Use of OECD/NEA Data Project Products in Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Cherkas, G.; Raducu, Gheorghe; Riznic, J.; Yalaoui, S.; Huang, Hui-Wen; Holy, Jaroslav; Holmberg, Jan-Erik; Sandberg, Jorma; Balmain, Michel; Bonnevialle, Anne-Marie; Curnier, Florence; Georgescu, Gabriel; Lanore, Jeanne-Marie; Lindner, Arndt; Fujimoto, Haruo; Ahn, Kwang-Il; Hwang, Taesuk; Jang, Seung-Cheol; Husarcek, Jan; Kovacs, Zoltan; Vazquez, Teresa; Johanson, Gunnar; Liwaang, Bo; Nyman, Ralph; Dang, Vinh; Schoen, Gerhard; Brook, Kevin; Hamblen, David; Siu, Nathan; Sturzebecher, Karl; Tobin, Margaret; Wood, Jeff; Amri, Abdallah; Breest, Axel

    2014-01-01

    The Nuclear Energy Agency (NEA)/Committee for the Safety of Nuclear Installations' (CSNI) Working Group on Risk Assessment (WGRISK) is tasked with supporting the improved use of Probabilistic Safety Assessment (PSA) in risk informed regulation and safety management through the analysis of results and the development of perspectives regarding potentially important risk contributors and associated risk reduction strategies. The task consists of the following major activities: Development, distribution, and completion of survey questionnaires; Analysis of survey questionnaire results at a task workshop; Preparation of the final task report. The main objectives of this task, as proposed by WGRISK and approved by CSNI, are the following: - Identification and characterization of the current uses of OECD data project products and data in support of PSA. In this context, the term 'products' refers to data analysis results, technical reports, and other project outputs. - Identification and characterization of technical and programmatic characteristics that either support or impede use of data project products in PSA. This includes an assessment of which PSA parameters could be potentially estimated from the various data project products and gaps between available product information and PSA data needs. - Identification of recommendations for enhancing the usefulness of data project products and the coordination between WGRISK and the data projects. This task report consists of the following sections: - Chapter 1 Provides a general overview of motivation and approach used for this task. - Chapter 2 Describes scope and objectives of the task. - Chapter 3 Provides an overview of the ICDE, FIRE, OPDE/CODAP, and COMPSIS data projects. For each project, the project objectives, project history, data collection methodology and quality assurance, project status, example PSA Applications, and information related to project participation is provided. - Chapter 4 Describes the

  12. Towards the Verification of Safety-critical Autonomous Systems in Dynamic Environments

    Directory of Open Access Journals (Sweden)

    Adina Aniculaesei

    2016-12-01

    Full Text Available There is an increasing necessity to deploy autonomous systems in highly heterogeneous, dynamic environments, e.g. service robots in hospitals or autonomous cars on highways. Due to the uncertainty in these environments, the verification results obtained with respect to the system and environment models at design-time might not be transferable to the system behavior at run time. For autonomous systems operating in dynamic environments, safety of motion and collision avoidance are critical requirements. With regard to these requirements, Macek et al. [6] define the passive safety property, which requires that no collision can occur while the autonomous system is moving. To verify this property, we adopt a two phase process which combines static verification methods, used at design time, with dynamic ones, used at run time. In the design phase, we exploit UPPAAL to formalize the autonomous system and its environment as timed automata and the safety property as TCTL formula and to verify the correctness of these models with respect to this property. For the runtime phase, we build a monitor to check whether the assumptions made at design time are also correct at run time. If the current system observations of the environment do not correspond to the initial system assumptions, the monitor sends feedback to the system and the system enters a passive safe state.

  13. Implementation of a risk assessment tool based on a probabilistic safety assessment developed for radiotherapy practices

    International Nuclear Information System (INIS)

    Paz, A.; Godinez, V.; Lopez, R.

    2010-10-01

    The present work describes the implementation process and main results of the risk assessment to the radiotherapy practices with Linear Accelerators (Linac), with cobalt 60, and with brachytherapy. These evaluations were made throughout the risk assessment tool for radiotherapy practices SEVRRA (risk evaluation system for radiotherapy), developed at the Mexican National Commission in Nuclear Safety and Safeguards derived from the outcome obtained with the Probabilistic Safety Analysis developed at the Ibero-American Regulators Forum for these radiotherapy facilities. The methodology used is supported by risk matrices method, a mathematical tool that estimates the risk to the patient, radiation workers and public from mechanical failures, mis calibration of the devices, human mistakes, and so. The initiating events are defined as those undesirable events that, together with other failures, can produce a delivery of an over-dose or an under-dose of the medical prescribed dose, to the planned target volume, or a significant dose to non prescribed human organs. Initiating events frequency and reducer of its frequency (actions intended to avoid the accident) are estimated as well as robustness of barriers to those actions, such as mechanical switches, which detect and prevent the accident from occurring. The spectrum of the consequences is parameterized, and the actions performed to reduce the consequences are identified. Based on this analysis, a software tool was developed in order to simplify the evaluations to radiotherapy installations and it has been applied as a first step forward to some Mexican installations, as part of a national implementation process, the final goal is evaluation of all Mexican facilities in the near future. The main target and benefits of the SEVRRA implementation are presented in this paper. (Author)

  14. Implementation of a risk assessment tool based on a probabilistic safety assessment developed for radiotherapy practices

    Energy Technology Data Exchange (ETDEWEB)

    Paz, A.; Godinez, V.; Lopez, R., E-mail: abpaz@cnsns.gob.m [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2010-10-15

    The present work describes the implementation process and main results of the risk assessment to the radiotherapy practices with Linear Accelerators (Linac), with cobalt 60, and with brachytherapy. These evaluations were made throughout the risk assessment tool for radiotherapy practices SEVRRA (risk evaluation system for radiotherapy), developed at the Mexican National Commission in Nuclear Safety and Safeguards derived from the outcome obtained with the Probabilistic Safety Analysis developed at the Ibero-American Regulators Forum for these radiotherapy facilities. The methodology used is supported by risk matrices method, a mathematical tool that estimates the risk to the patient, radiation workers and public from mechanical failures, mis calibration of the devices, human mistakes, and so. The initiating events are defined as those undesirable events that, together with other failures, can produce a delivery of an over-dose or an under-dose of the medical prescribed dose, to the planned target volume, or a significant dose to non prescribed human organs. Initiating events frequency and reducer of its frequency (actions intended to avoid the accident) are estimated as well as robustness of barriers to those actions, such as mechanical switches, which detect and prevent the accident from occurring. The spectrum of the consequences is parameterized, and the actions performed to reduce the consequences are identified. Based on this analysis, a software tool was developed in order to simplify the evaluations to radiotherapy installations and it has been applied as a first step forward to some Mexican installations, as part of a national implementation process, the final goal is evaluation of all Mexican facilities in the near future. The main target and benefits of the SEVRRA implementation are presented in this paper. (Author)

  15. Results of the Safety probabilistic analysis of Level 2 of the CNSNS

    International Nuclear Information System (INIS)

    Lopez M, R.; Godinez S, V.

    2004-01-01

    The National Commission of Nuclear Safety and Safeguards (CNSNS) it has concluded the one develop of their Probabilistic Analysis of Safety (APS) of Level 2. The reach of the study it considers internal events to full power and it was developed on the base of the methodology of the NUREG-1150, for what you it was built an Event Tree of the Progression of the Accident (APET) to analyze the 25 States of Damage to the Plant (PDS) obtained of the APS Nl of the CNSNS. In the APET are considered the phenomenology of severe accidents, the performance of mitigation systems and actions of the operator that could modify the evolution of a severe accident in the CNLV, as well as the diverse modes of failure of the primary container and it identifies the trajectories of liberation of radioactive material to the exterior. The conditional probabilities of failure of the primary container were obtained and it was characterized the time so much to which happens the liberation of radioactive material as the quantity of the term liberated source. Also, to establish the times and parameters of the evolution of accidents were selected representative accident sequences of the diverse accident types and their conditions were simulated by means of the MELCOR computer code. Also it was developed a code of parametric compute type XSOR, specific for Laguna Verde, with which it was carried out the estimate of the term source in each one of the release trajectories. In this work the main characteristic ones are presented and results of the APS N2 developed in the CNSNS and they are compared against the model and results of the EIP of the CNLV. (Author)

  16. Development and application of a living probabilistic safety assessment tool: Multi-objective multi-dimensional optimization of surveillance requirements in NPPs considering their ageing

    International Nuclear Information System (INIS)

    Kančev, Duško; Čepin, Marko; Gjorgiev, Blaže

    2014-01-01

    The benefits of utilizing the probabilistic safety assessment towards improvement of nuclear power plant safety are presented in this paper. Namely, a nuclear power plant risk reduction can be achieved by risk-informed optimization of the deterministically-determined surveillance requirements. A living probabilistic safety assessment tool for time-dependent risk analysis on component, system and plant level is developed. The study herein focuses on the application of this living probabilistic safety assessment tool as a computer platform for multi-objective multi-dimensional optimization of the surveillance requirements of selected safety equipment seen from the aspect of the risk-informed reasoning. The living probabilistic safety assessment tool is based on a newly developed model for calculating time-dependent unavailability of ageing safety equipment within nuclear power plants. By coupling the time-dependent unavailability model with a commercial software used for probabilistic safety assessment modelling on plant level, the frames of the new platform i.e. the living probabilistic safety assessment tool are established. In such way, the time-dependent core damage frequency is obtained and is further on utilized as first objective function within a multi-objective multi-dimensional optimization case study presented within this paper. The test and maintenance costs are designated as the second and the incurred dose due to performing the test and maintenance activities as the third objective function. The obtained results underline, in general, the usefulness and importance of a living probabilistic safety assessment, seen as a dynamic probabilistic safety assessment tool opposing the conventional, time-averaged unavailability-based, probabilistic safety assessment. The results of the optimization, in particular, indicate that test intervals derived as optimal differ from the deterministically-determined ones defined within the existing technical specifications

  17. Verification and validation of the safety parameter display system for nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Yuanfang

    1993-05-01

    During the design and development phase of the safety parameter display system for nuclear power plant, a verification and validation (V and V) plan has been implemented to improve the quality of system design. The V and V activities are briefly introduced, which were executed in four stages of feasibility research, system design, code development and system integration and regulation. The evaluation plan and the process of implementation as well as the evaluation conclusion of the final technical validation for this system are also presented in detail

  18. International exchange on nuclear safety related expert systems: The role of software verification and validation

    International Nuclear Information System (INIS)

    Sun, B.K.H.

    1996-01-01

    An important lesson learned from the Three Mile Island accident is that human errors can be significant contributors to risk. Recent advancement in computer hardware and software technology helped make expert system techniques potentially viable tools for improving nuclear power plant safety and reliability. As part of the general man-machine interface technology, expert systems have recently become increasingly prominent as a potential solution to a number of previously intractable problems in many phases of human activity, including operation, maintenance, and engineering functions. Traditional methods for testing and analyzing analog systems are no longer adequate to handle the increased complexity of software systems. The role of Verification and Validation (V and V) is to add rigor to the software development and maintenance cycle to guarantee the high level confidence needed for applications. Verification includes the process and techniques for confirming that all the software requirements in one stage of the development are met before proceeding on to the next stage. Validation involves testing the integrated software and hardware system to ensure that it reliably fulfills its intended functions. Only through a comprehensive V and V program can a high level of confidence be achieved. There exist many different standards and techniques for software verification and validation, yet they lack uniform approaches that provides adequate levels of practical guidance which can be used by users for nuclear power plant applications. There is a need to unify different approaches for addressing software verification and validation and to develop practical and cost effective guidelines for user and regulatory acceptance. (author). 8 refs

  19. Probabilistic methods of optimization of scheduled tests for heat equipment of safety systems of reactor at full power

    International Nuclear Information System (INIS)

    Bilej, D.V.; Fridman, N.A.; Kolykhanov, V.N.; Skalozubov, V.I.

    2004-01-01

    This article generalises the basic results of a long-term teamwork with respect to a scientific and technical substantiation of perfection of the regulations of safe operation power units with VVER. This perfection is concerning a periodicity and volumes of tests of safety systems when a reactor works at full power. The article shows that the application of the probabilistic approaches connected to minimisation of a risk criterion function is an effective methodical base for the optimisation. For certain safety systems of serial power units with VVER 1000 the results of calculated substantiations are presented

  20. Development of a tool of probabilistic safety analysis for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Hidalgo H, F.E.; Fran N, P.

    2007-01-01

    It is developing a tool to explain in a simple way in that it consists the Probabilistic Safety Analysis (APS) and at the same time to facilitate the comparison among the different designs of advanced nuclear reactors starting from their safety systems. This tool for teaching contemplates all the workspaces in an APS, but it is deepened only in what is the development of accident sequences and systems models. At the moment its have incorporated three types of advanced reactors, ABWR, ESBWR, and the HTGR and they are compared among if and with a BWR like that of Laguna Verde. This tool is carried out in Visual Basic code because it is a platform that can be used in any Windows atmosphere and for their easy programming. The system includes a tree of events developed for this purpose for a research HTGR built in Japan (HTTR) to have a point of comparison of the same one with other reactors of previous generations. It is that in the fourth generation reactors the measure of frequency of core damage doesn't make the same sense that for reactors of previous generations, which is due to its passive safety systems and its design type of the fuel, that which makes indispensable the development of another type of risk measure. The tree of events is presented for the initiator event 'the rupture of the main pipe' that causes the depressurization of the HTTR reactor. In this article it was concluded that it is necessary to evaluate the accident until reaching to the liberation of fission products that one knows in APS like an APS study level 1 and level 2 together. The final states developed starting from the possible phenomena that happen in these scenarios are presented. For this, its are considered flaws of all the mitigation systems that intervene in this accident. The tree of events developed for this work and the definition of the final states contributes to the development of as carrying out an APS for fourth generation reactors, with the purpose of developing an APS

  1. CNE (Embalse nuclear power plant): probabilistic safety study. Loss of service water. Probabilistic evaluation and analysis through events sequence

    International Nuclear Information System (INIS)

    Couto, A.J.; Perez, S.S.

    1987-01-01

    This work is part of a study on the service water systems of the Embalse nuclear power plant from a safety point of view. The faults of service water systems of high and low pressure that can lead to situations threatening the plant safety were analyzed in a previous report. The event 'total loss of low pressure service water' causes the largest number of such conditions. Such event is an operational incident that can lead to an accident situation due to faults in the required process systems or by omission of a procedure. The annual frequency of the event 'total loss of low pressure service water' is calculated. The main contribution comes from pump failure. The evaluation of the accident sequences shows that the most direct way to the liberation of fission products is the loss of steam generators as heat sink. The contributions to small and large LOCA and electric supply loss are analyzed. The sequence that leads to tritium release through boiling of moderator is also evaluated. (Author)

  2. Verification of MCNP6.2 for Nuclear Criticality Safety Applications

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-10

    Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.

  3. Improving Patient Safety With Error Identification in Chemotherapy Orders by Verification Nurses.

    Science.gov (United States)

    Baldwin, Abigail; Rodriguez, Elizabeth S

    2016-02-01

    The prevalence of medication errors associated with chemotherapy administration is not precisely known. Little evidence exists concerning the extent or nature of errors; however, some evidence demonstrates that errors are related to prescribing. This article demonstrates how the review of chemotherapy orders by a designated nurse known as a verification nurse (VN) at a National Cancer Institute-designated comprehensive cancer center helps to identify prescribing errors that may prevent chemotherapy administration mistakes and improve patient safety in outpatient infusion units. This article will describe the role of the VN and details of the verification process. To identify benefits of the VN role, a retrospective review and analysis of chemotherapy near-miss events from 2009-2014 was performed. A total of 4,282 events related to chemotherapy were entered into the Reporting to Improve Safety and Quality system. A majority of the events were categorized as near-miss events, or those that, because of chance, did not result in patient injury, and were identified at the point of prescribing.

  4. Verification and validation issues for digitally-based NPP safety systems

    International Nuclear Information System (INIS)

    Ets, A.R.

    1993-01-01

    The trend toward standardization, integration and reduced costs has led to increasing use of digital systems in reactor protection systems. While digital systems provide maintenance and performance advantages, their use also introduces new safety issues, in particular with regard to software. Current practice relies on verification and validation (V and V) to ensure the quality of safety software. However, effective V and V must be done in conjunction with a structured software development process and must consider the context of the safety system application. This paper present some of the issues and concerns that impact on the V and V process. These include documentation of systems requirements, common mode failures, hazards analysis and independence. These issues and concerns arose during evaluations of NPP safety systems for advanced reactor designs and digital I and C retrofits for existing nuclear plants in the United States. The pragmatic lessons from actual systems reviews can provide a basis for further refinement and development of guidelines for applying V and V to NPP safety systems. (author). 14 refs

  5. Verification of FPGA-Signal using the test board which is applied to Safety-related controller

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Youn-Hu; Yoo, Kwanwoo; Lee, Myeongkyun; Yun, Donghwa [SOOSAN ENS, Seoul (Korea, Republic of)

    2016-10-15

    This article aims to provide the verification method for BGA-type FPGA of Programmable Logic Controller (PLC) developed as Safety Class. The logic of FPGA in the control device with Safety Class is the circuit to control overall logic of PLC. Saftety-related PLC must meet the international standard specifications. With this reason, we use V and V according to an international standard in order to secure high reliability and safety. By using this, we are supposed to proceed to a variety of verification courses for extra reliability and safety analysis. In order to have efficient verification of test results, we propose the test using the newly changed BGA socket which can resolve the problems of the conventional socket on this paper. The Verification of processes is divided into verification of Hardware and firmware. That processes are carried out in the unit testing and integration testing. The proposed test method is simple, the effect of cost reductions by batch process. In addition, it is advantageous to measure the signal from the Hi-speed-IC due to its short length of the pins and it was plated with the copper around it. Further, it also to prevent abrasion on the IC ball because it has no direct contact with the PCB. Therefore, it can be actually applied is to the BGA package test and we can easily verify logic as well as easily checking the operation of the designed data.

  6. Evaluation of the probabilistic safety assessment portfolio for NSD. Plan of work

    International Nuclear Information System (INIS)

    Gould, J.

    1999-01-01

    The aim is to use the research portfolio evaluation protocol developed by HSL to evaluate the Probabilistic Safety Assessment (PSA) portfolio, both to draw conclusions about the PSA portfolio and as a pilot study to show the suitability of the evaluation protocol. The objectives of the work are: (1) To collect sufficient information to carry out a preliminary review of the portfolio; (2) o produce a plan of work detailing the time and costs to carry out a full evaluation of the PSA portfolio; (3) to evaluate the PSA portfolio of research; (4) to produce a report of the evaluation of the PSA portfolio; (5) if necessary, to make changes to the methodology in light of the experience gained in the evaluation of the PSA research portfolio. This report completes objectives 1 and 2. It details the plan of work for the evaluation of the PSA research portfolio. The plan has shown that the evaluation of the PSA research portfolio has many difficulties to overcome. It is suitable as a pilot study to show the suitability of the portfolio evaluation protocol and will provide valuable information that can be used to improve it. The evaluation of the PSA portfolio will require a considerable amount of time and effort to complete. The task analysis has shown it to be of the order of Pound Sterling 25k and to take two months to complete after this preliminary data collection. The plan to evaluate the PSA research portfolio detailed in this report should be carried out and the lessons learned by carrying out this pilot study should be used to improve the evaluation protocol

  7. CNE (Embalse nuclear power plant): probabilistic safety study. Electric power supply. Events sequence

    International Nuclear Information System (INIS)

    Figueroa, N.

    1987-01-01

    The plant response to the occurrence of the starting event 'total loss of electric power supply to class IV and class III' is analyzed. This involves the study of automatical actions of safety and process systems as well as the operator actions. The probabilistic evaluation of starting event frequency is performed through fault-tree techniques. The frequency of occurrence 'loss of electric power supply to class IV (λIV = 0.56/year) and the probability of failure to demand of 'reserve' generating groups (Pd III 6.79 x 10 -3 ) contribute to the mentioned frequency. As soon as the starting event occurs, the reactor power must be reduced to 0%, the fuel must be cooled through the thermo siphon and decay heat has to be removed. The events sequence analysis leads to the conclusion that the non shutting down of the reactor with any of the shutdown systems is 'incredible' (10 -6 /year). In all cases the fuel is cooled by building the thermo siphon except when a substantial inventory loss exist due to a closure failure of some valve of pressure and inventory control system. The order of magnitude of the failure of decay heat removal through the steam generators is 4 x 10 -4 . This removal would be assured by the emergency water system. Therefore, the frequency of the sequence of possible core meltdown, when the reactor does not shut down is: λ = 5 x 10 -9 /year and for the failure of heat removal: λ = 2 x 10 -6 /year. (Author)

  8. Safety of long-distance pipelines. Probabilistic and deterministic aspects; Sicherheit von Rohrfernleitungen. Probabilistik und Deterministik im Vergleich

    Energy Technology Data Exchange (ETDEWEB)

    Hollaender, Robert [Leipzig Univ. (Germany). Inst. fuer Infrastruktur und Ressourcenmanagement

    2013-03-15

    The Committee for Long-Distance Pipelines (Berlin, Federal Republic of Germany) reported on the relation between deterministic and probabilistic approaches in order to contribute to a better understanding of the safety management of long-distance pipelines. The respective strengths and weaknesses as well as the deterministic and probabilistic fundamentals of the safety management are described. The comparison includes fundamental aspects, but is essentially determined by the special character of the technical plant 'long-distance pipeline' as an infrastructure project in the area. This special feature results to special operation conditions and related responsibilities. However, our legal system 'long-distance pipeline' does not grant the same legal position in comparison to other infrastructural facilities such as streets and railways. Thus, the question whether and in what manner the impacts from the land-use in the environment of long-distance pipelines have to be considered is again and again the initial point for the discussion on probabilistic and deterministic approaches.

  9. Operational verification of a framework for the probabilistic nowcasting of river discharge in small and medium size basins

    Directory of Open Access Journals (Sweden)

    F. Silvestro

    2012-03-01

    Full Text Available Forecasting river discharge is a very important issue for the prediction and monitoring of ground effects related to severe precipitation events. The meteorological forecast systems are unable to predict precipitation on small spatial (few km and temporal (hourly scales. For these reasons the issuing of reliable flood forecasts is not feasible in those regions where the basin's response to rainfall events is very fast and can generate flash floods. This problem can be tackled by using rainfall nowcasting techniques based on radar observations coupled with hydrological modeling. These procedures allow the forecasting of future streamflow with a few hours' notice. However, to account for the short-term uncertainties in the evolution of fine scale precipitation field, a probabilistic approach to rainfall nowcasting is needed. These uncertainties are then propagated from rainfall to runoff through a distributed hydrological model producing a set of equi-probable discharge scenarios to be used for the flood nowcasting with time horizons of a few hours. Such a hydrological nowcasting system is presented here and applied to some case studies. A first evaluation of its applicability in an operational context is provided and the opportunity of using the results quantitatively is discussed.

  10. Ageing effects modelling in probabilistic safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Nitoi, M.; Turcu, I.; Florescu, G.; Apostol, M.; Farcasiu, M.; Pavelescu, M.

    2005-01-01

    Ageing management has become a major concern for many responsible organizations during the last years, because as the operating power plants have got older, they may have the tendency to become less safe. The effects of age-related degradation of plant components, systems and structures are necessary to be assessed in order to assure a continuous safe operation of nuclear power plants. The Probabilistic Safety Analysis (PSA) is an efficient system analysis method which is used to assess the risk of operation of nuclear power plants. In the assessment of risk level for a plant, most of the PSA studies generally didn't take into account the ageing effects, and uses a time averaged unavailability. By incorporation of ageing effects, the results enable an identification of the components that have the greatest effect on risk if their failure rates increase due to ageing effects modelling. In this paper, it was assessed the impact on Class IV Electrical Power System unavailability of the assumed increase in components failure probability caused by components ageing. The electrical system was chosen for the study because there are a lot of cables and for these types of equipment there is no planned preventive or corrective maintenance, and they are originally designed to reach the end of plant life with an adequate safety margin. To quantify the effects of age-related degradation on components, the linear ageing model was used. In this model, the failure rate of a component λ (t) is expressed as a sum of two independent failure rates, one associated with random failure, λ 0 , and the other associated with failures due to aging α, so: λ(t) = λ 0 + αt. The basic events were coded using a computer code similar to CAFTA, developed in INR Pitesti. For the reliability data allocation for basic events a intern data base was used. This data base contains data from the following generic data bases: IAEA Component Reliability Data for use in PSA, Point Lepreau Component

  11. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

  12. Verification and testing of the RTOS for safety-critical embedded systems

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Na Young [Seoul National University, Seoul (Korea, Republic of); Kim, Jin Hyun; Choi, Jin Young [Korea University, Seoul (Korea, Republic of); Sung, Ah Young; Choi, Byung Ju [Ewha Womans University, Seoul (Korea, Republic of); Lee, Jang Soo [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Development in Instrumentation and Control (I and C) technology provides more convenience and better performance, thus, adopted in many fields. To adopt newly developed technology, nuclear industry requires rigorous V and V procedure and tests to assure reliable operation. Adoption of digital system requires verification and testing of the OS for licensing. Commercial real-time operating system (RTOS) is targeted to apply to various, unpredictable needs, which makes it difficult to verify. For this reason, simple, application-oriented realtime OS is developed for the nuclear application. In this work, we show how to verify the developed RTOS at each development lifecycle. Commercial formal tool is used in specification and verification of the system. Based on the developed model, software in C language is automatically generated. Tests are performed for two purposes; one is to identify consistency between the verified model and the generated code, the other is to find errors in the generated code. The former assumes that the verified model is correct, and the latter incorrect. Test data are generated separately to satisfy each purpose. After we test the RTOS software, we implement the test board embedded with the developed RTOS and the application software, which simulates the safety critical plant protection function. Testing to identify whether the reliability criteria is satisfied or not is also designed in this work. It results in that the developed RTOS software works well when it is embedded in the system.

  13. Verification and testing of the RTOS for safety-critical embedded systems

    International Nuclear Information System (INIS)

    Lee, Na Young; Kim, Jin Hyun; Choi, Jin Young; Sung, Ah Young; Choi, Byung Ju; Lee, Jang Soo

    2003-01-01

    Development in Instrumentation and Control (I and C) technology provides more convenience and better performance, thus, adopted in many fields. To adopt newly developed technology, nuclear industry requires rigorous V and V procedure and tests to assure reliable operation. Adoption of digital system requires verification and testing of the OS for licensing. Commercial real-time operating system (RTOS) is targeted to apply to various, unpredictable needs, which makes it difficult to verify. For this reason, simple, application-oriented realtime OS is developed for the nuclear application. In this work, we show how to verify the developed RTOS at each development lifecycle. Commercial formal tool is used in specification and verification of the system. Based on the developed model, software in C language is automatically generated. Tests are performed for two purposes; one is to identify consistency between the verified model and the generated code, the other is to find errors in the generated code. The former assumes that the verified model is correct, and the latter incorrect. Test data are generated separately to satisfy each purpose. After we test the RTOS software, we implement the test board embedded with the developed RTOS and the application software, which simulates the safety critical plant protection function. Testing to identify whether the reliability criteria is satisfied or not is also designed in this work. It results in that the developed RTOS software works well when it is embedded in the system

  14. Verification of criticality safety in on-site spent fuel storage systems

    International Nuclear Information System (INIS)

    Rasmussen, R.W.

    1989-01-01

    On February 15, 1984, Duke Power Company received approval for a two-region, burnup credit, spent fuel storage rack design at both Units 1 and 2 of the McGuire Nuclear Station. Duke also hopes to obtain approval by January of 1990 for a dry spent fuel storage system at the Oconee Nuclear Station, which will incorporate the use of burnup credit in the criticality analysis governing the design of the individual storage units. While experiences in burnup verification for criticality safety for their dry storage system at Oconee are in the future, the methods proposed for burnup verification will be similar to those currently used at the McGuire Nuclear Station in the two-region storage racks installed in both pools. In conclusion, the primary benefit of the McGuire rerack effort has obviously been the amount of storage expansion it provided. A total increase of about 2,000 storage cells was realized, 1,000 of which were the result of pursuing the two-region rather than the conventional poison rack design. Less impacting, but equally as important, however, has been the experience gained during the planning, installation, and operation of these storage racks. This experience should prove useful for future rerack efforts likely to occur at Duke's Catawba Nuclear Station as well as for the current dry storage effort underway for the Oconee Nuclear Station

  15. The use of probabilistic safety assessment (PSA) based maintenance indicators to increase the availability of safety related systems in nuclear power plants

    International Nuclear Information System (INIS)

    Kirchsteiger, C.

    1991-04-01

    This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain nuclear power plant. Plant specific equipment history and maintenance work on data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant components as well as the detection of the dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system-based and maintenance-related performance indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

  16. The use of probabilistic safety assessment based maintenance indicators to increase the availability of safety related systems in nuclear power plants

    International Nuclear Information System (INIS)

    Kirchsteiger, C.

    1991-04-01

    This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain Nuclear Power Plant. Plant specific equipment history and maintenance work order data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis) in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system based and maintenance related Performance Indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

  17. A computational method for probabilistic safety assessment of I and C systems and human operators in nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Man Cheol; Seong, Poong Hyun

    2006-01-01

    To make probabilistic safety assessment (PSA) more realistic, the improvements of human reliability analysis (HRA) are essential. But, current HRA methods have many limitations including the lack of considerations on the interdependency between instrumentation and control (I and C) systems and human operators, and lack of theoretical basis for situation assessment of human operators. To overcome these limitations, we propose a new method for the quantitative safety assessment of I and C systems and human operators. The proposed method is developed based on the computational models for the knowledge-driven monitoring and the situation assessment of human operators, with the consideration of the interdependency between I and C systems and human operators. The application of the proposed method to an example situation demonstrates that the quantitative description by the proposed method for a probable scenario well matches with the qualitative description of the scenario. It is also demonstrated that the proposed method can probabilistically consider all possible scenarios and the proposed method can be used to quantitatively evaluate the effects of various context factor on the safety of nuclear power plants. In our opinion, the proposed method can be used as the basis for the development of advanced HRA methods

  18. General-purpose heat source safety verification test series: SVT-11 through SVT-13

    International Nuclear Information System (INIS)

    George, T.G.; Pavone, D.

    1986-05-01

    The General-Purpose Heat Source (GPHS) is a modular component of the radioisotope thermoelectric generator that will provide power for the Galileo and Ulysses (formerly ISPM) space missions. The GPHS provides power by transmitting the heat of 238 Pu α-decay to an array of thermoelectric elements. Because the possibility of an orbital abort always exists, the heat source was designed and constructed to minimize plutonia release in any accident environment. The Safety Verification Test (SVT) series was formulated to evaluate the effectiveness of GPHS plutonia containment after atmospheric reentry and Earth impact. The first two reports (covering SVT-1 through SVT-10) described the results of flat, side-on, and angular module impacts against steel targets at 54 m/s. This report describes flat-on module impacts against concrete and granite targets, at velocities equivalent to or higher than previous SVTs

  19. The choice between two designs for the safety-injection system of a pressurized-water reactor, using probabilistic methods

    International Nuclear Information System (INIS)

    Villemeur, Alain

    1982-01-01

    A probabilistic study has been carried out to compare two designs for the safety-injection circuit of a pressurized-water reactor. It appears that unavailability of the circuit after an accident involving loss of coolant decreases little when one moves from a 2-line to a 3-line system. These results are compared with the disadvantages arising from increased redundancy, and in particular the increased cost of the installations. The 2-line circuit appears the optimum one on the basis of cost and reliability criteria. It has been chosen for the 1300-MWe units [fr

  20. 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

  1. The importance of probabilistic evaluations in connection with risk analyses according to technical safety laws

    International Nuclear Information System (INIS)

    Mathiak, E.

    1984-01-01

    The nuclear energy sector exemplifies the essential importance to be attached to the practical application of probabilistic evaluations (e.g. probabilistic reliability analyses) in connection with the legal risk assessment of technical systems and installations. The study is making use of a triad risk analysis and tries to reconcile the natural science and legal points of view. Without changing the definitions of 'risk' and 'hazard' in the legal sense of their meaning the publication discusses their reconcilation with the laws of natural science, their interpretation and application in view of the latter. (HSCH) [de

  2. Probabilistic evaluation of scenarios in long-term safety analyses. Results of the project ISIBEL; Probabilistische Bewertung von Szenarien in Langzeitsicherheitsanalysen. Ergebnisse des Vorhabens ISIBEL

    Energy Technology Data Exchange (ETDEWEB)

    Buhmann, Dieter; Becker, Dirk-Alexander; Laggiard, Eduardo; Ruebel, Andre; Spiessl, Sabine; Wolf, Jens

    2016-07-15

    In the frame of the project ISIBEL deterministic analyses on the radiological consequences of several possible developments of the final repository were performed (VSG: preliminary safety analysis of the site Gorleben). The report describes the probabilistic evaluation of the VSG scenarios using uncertainty and sensitivity analyses. It was shown that probabilistic analyses are important to evaluate the influence of uncertainties. The transfer of the selected scenarios in computational cases and the used modeling parameters are discussed.

  3. Procedures for conducting probabilistic safety assessments of nuclear power plants (level 2). Accident progression, containment analysis and estimation of accident source terms

    International Nuclear Information System (INIS)

    1995-01-01

    The present publication on Level 2 PSA is based on a compilation and review of practices in various Member States. It complements Safety Series No. 50-P-4, issued in 1992, on Procedures for Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 1). Refs, figs and tabs

  4. Verification of a primary-to-secondary leaking safety procedure in a nuclear power plant using coloured Petri nets

    International Nuclear Information System (INIS)

    Nemeth, E.; Bartha, T.; Fazekas, Cs.; Hangos, K.M.

    2009-01-01

    This paper deals with formal and simulation-based verification methods of a PRImary-to-SEcondary leaking (abbreviated as PRISE) safety procedure. The PRISE safety procedure controls the draining of the contaminated water in a faulty steam generator when a non-compensable leaking from the primary to the secondary circuit occurs. Because of the discrete nature of the verification, a Coloured Petri Net (CPN) representation is proposed for both the procedure and the plant model. We have proved by using a non-model-based strategy that the PRISE safety procedure is safe, there are no dead markings in the state space, and all transitions are live; being either impartial or fair. Further analysis results have been obtained using a model-based verification approach. We created a simple, low dimensional, nonlinear dynamic model of the primary circuit in a VVER-type pressurized water nuclear power plant for the purpose of the model-based verification. This is in contrast to the widely used safety analysis that requires an accurate detailed model. Our model also describes the relevant safety procedures, as well as all of the major leaking-type faults. We propose a novel method to transform this model to a CPN form by discretization. The composed plant and PRISE safety procedure system has also been analysed by simulation using CPN analysis tools. We found by the model-based analysis-using both single and multiple faults-that the PRISE safety procedure initiates the draining when the PRISE event occurs, and no false alarm will be initiated

  5. KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jang Yeol; Lee, Jang Soo; Eom, Heung Seop

    1997-07-01

    This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs.

  6. KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

    International Nuclear Information System (INIS)

    Kim, Jang Yeol; Lee, Jang Soo; Eom, Heung Seop.

    1997-07-01

    This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs

  7. On the functional failures concept and probabilistic safety margins: challenges in application for evaluation of effectiveness of shutdown systems - 15318

    International Nuclear Information System (INIS)

    Serghiuta, D.; Tholammakkil, J.

    2015-01-01

    The use of level-3 reliability approach and the concept of functional failure probability could provide the basis for defining a safety margin metric which would include a limit for the probability of functional failure, in line with the definition of a reliability-based design. It can also allow a quantification of level of confidence, by explicit modeling and quantification of uncertainties, and provide a better framework for representation of actual design and optimization of design margins within an integrated probabilistic-deterministic model. This paper reviews the attributes and challenges in application of functional failure concept in evaluation of risk-informed safety margins using as illustrative example the case of CANDU reactors shutdown systems effectiveness. A risk-informed formulation is first introduced for estimation of a reasonable limit for the functional failure probability using a Swiss cheese model. It is concluded that more research is needed in this area and a deterministic - probabilistic approach may be a reasonable intermediate step for evaluation of functional failure probability at the system level. The views expressed in this paper are those of the authors and do not necessarily reflect those of CNSC, or any part thereof. (authors)

  8. Estimation of the loss of Offsite power frequency for the probabilistic safety assessment of the Juragua NPP

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Valhuerdi Debesa, C.

    1996-01-01

    The loss offsite power is defined as the interruption of the preferred power supply to the essential and non essential switchgear buses necessitating or resulting in the use of emergency AC power supply. Because many safety system required for reactor core decay heat removal and containment heat removal depend on AC power, a loss of offsite power, if emergency power supply (diesel generators) fails, could be severe accidents The purpose of this work was to determine, for the Probabilistic Safety Assessment of the Juragua NPP, the causes, frequency and duration relationships of the loss of offsite power. A description is presented of the different factor that determine the occurrence of this event and the characteristics for the Juragua NPP

  9. The use of probabilistic safety analysis in design and operation -- Lessons learned from Sizewell B. Annex 14

    International Nuclear Information System (INIS)

    Buttery, N.E.

    2002-01-01

    Probabilistic Safety Assessments (PSAs) have been used extensively in the design and licensing of Sizewell B. This paper outlines the role of PSA in the UK licensing process and describes how it has been applied to Sizewell B during both the pre-construction and pre-operational phases. From this experience a 'Living PSA' has been formulated which continues be used to support operation. The application of PSA to Sizewell B has demonstrated that it is a powerful tool with potential for future use. Its strengths and limitations as a tool need to recognised by both users and regulators. It is not a fully mechanistic means of ensuring design safety, but is an important aid to decision making. It also has the potential to allow risk judgements to be taken in conjunction with commercial and environmental issues. (author)

  10. Development and Verification of the Computer Codes for the Fast Reactors Nuclear Safety Justification

    International Nuclear Information System (INIS)

    Kisselev, A.E.; Mosunova, N.A.; Strizhov, V.F.

    2015-01-01

    The information on the status of the work on development of the system of the nuclear safety codes for fast liquid metal reactors is presented in paper. The purpose of the work is to create an instrument for NPP neutronic, thermohydraulic and strength justification including human and environment radiation safety. The main task that is to be solved by the system of codes developed is the analysis of the broad spectrum of phenomena taking place on the NPP (including reactor itself, NPP components, containment rooms, industrial site and surrounding area) and analysis of the impact of the regular and accidental releases on the environment. The code system is oriented on the ability of fully integrated modeling of the NPP behavior in the coupled definition accounting for the wide range of significant phenomena taking place on the NPP under normal and accident conditions. It is based on the models that meet the state-of-the-art knowledge level. The codes incorporate advanced numerical methods and modern programming technologies oriented on the high-performance computing systems. The information on the status of the work on verification of the separate codes of the system of codes is also presented. (author)

  11. Validation and Verification of Future Integrated Safety-Critical Systems Operating under Off-Nominal Conditions

    Science.gov (United States)

    Belcastro, Christine M.

    2010-01-01

    Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents and reducing them will require a holistic integrated intervention capability. Future onboard integrated system technologies developed for preventing loss of vehicle control accidents must be able to assure safe operation under the associated off-nominal conditions. The transition of these technologies into the commercial fleet will require their extensive validation and verification (V and V) and ultimate certification. The V and V of complex integrated systems poses major nontrivial technical challenges particularly for safety-critical operation under highly off-nominal conditions associated with aircraft loss-of-control events. This paper summarizes the V and V problem and presents a proposed process that could be applied to complex integrated safety-critical systems developed for preventing aircraft loss-of-control accidents. A summary of recent research accomplishments in this effort is also provided.

  12. Methodology for the application of probabilistic safety assessment techniques (PSA) to the cobalt-therapy units in Cuba

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Ferro Fernandez, R.; Troncoso Fleitas, M.; Lozano Lima, B.; Fuente Puch, A. de la; Perez Reyes, Y.; Dumenigo Gonzalez, C.

    2001-01-01

    The applications of PSA techniques in the nuclear power plants during the last two decades and the positive results obtained for decision making in relation with safety, as a complement to deterministic methods, have increased their use in the rest of the nuclear applications. At present a large set of documents from international institutions can be found summarizing the investigations carried out in this field and promoting their use in radioactive facilities. Although still without a mandatory character, the new regulations on radiological safety also promote the complete or partial application of the PSA techniques in the safety assessment of the radiological practices. Also the IAEA, through various programs in which Cuba has been inserted, is taking a group of actions so that the nuclear community will encourage the application of the probabilistic risk methods for the evaluations and decision making with respect to safety. However, the fact that in no radioactive installation has a complete PSA study been carried out, makes that certain methodological aspects require to be improved and modified for the application of these techniques. This work presents the main elements for the use of PSA in the evaluation of the safety of cobalt-therapy units in Cuba. Also presented, as part of the results of the first stage of the Study, are the Guidelines that are being applied in a Research Contract with the Agency by the authors themselves, who belong to the CNSN, together with other specialists from the Cuban Ministry of Public Health. (author) [es

  13. Proceedings of the Specialist Meeting on the Seismic Probabilistic Safety Assessment of Nuclear Facilities

    International Nuclear Information System (INIS)

    2007-01-01

    The main objectives of the Meeting were to review recent advances in the methodology of Seismic Probabilistic Safety Assessment (SPSA), to discuss practical applications, to review the current state of the art, and to identify methodological issues where further research would be beneficial in enhancing the usefulness of the methodology. Applications of the Seismic Margin Assessment methodology (SMA), a methodology related to SPSA, were also discussed. One specific objective was to compare the situation today with the situation at the time of the 1999 Tokyo workshop, and to develop a set of findings and recommendations that would update those from that earlier workshop. There was a consensus at the Specialists Meeting that SPSA is now in widespread use throughout the nuclear-power industry worldwide, by the operating nuclear power plants (NPPs) themselves, by the various national regulatory agencies, and by the designers of new NPPs. It was also widely agreed that it can systematically accomplish several very important objectives; specifically, it can contribute: - To understanding the seismic risk arising from NPPs. - To understanding the safety significance of seismic design shortfalls. - To prioritizing seismic safety improvements. - To evaluating and improving seismic regulations. - To modifying the seismic regulatory/licensing basis of an individual NPP. Compared to the situation in 1999, when the first Workshop was held in Tokyo, there have been significant expansions in the use of SPSA in many different areas. Some countries provided detailed information on their regulatory framework for using seismic PSA. Many other countries also provided some information in their papers as background for conducting SPSA. During the Meeting, a small number of important weaknesses in SPSA methodology were identified. None of these are new, all having been widely recognized for many years. However, for some of the weaknesses, extensive discussions during the Meeting provided

  14. Proceedings of the Specialist Meeting on the Seismic Probabilistic Safety Assessment of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-11-14

    The main objectives of the Meeting were to review recent advances in the methodology of Seismic Probabilistic Safety Assessment (SPSA), to discuss practical applications, to review the current state of the art, and to identify methodological issues where further research would be beneficial in enhancing the usefulness of the methodology. Applications of the Seismic Margin Assessment methodology (SMA), a methodology related to SPSA, were also discussed. One specific objective was to compare the situation today with the situation at the time of the 1999 Tokyo workshop, and to develop a set of findings and recommendations that would update those from that earlier workshop. There was a consensus at the Specialists Meeting that SPSA is now in widespread use throughout the nuclear-power industry worldwide, by the operating nuclear power plants (NPPs) themselves, by the various national regulatory agencies, and by the designers of new NPPs. It was also widely agreed that it can systematically accomplish several very important objectives; specifically, it can contribute: - To understanding the seismic risk arising from NPPs. - To understanding the safety significance of seismic design shortfalls. - To prioritizing seismic safety improvements. - To evaluating and improving seismic regulations. - To modifying the seismic regulatory/licensing basis of an individual NPP. Compared to the situation in 1999, when the first Workshop was held in Tokyo, there have been significant expansions in the use of SPSA in many different areas. Some countries provided detailed information on their regulatory framework for using seismic PSA. Many other countries also provided some information in their papers as background for conducting SPSA. During the Meeting, a small number of important weaknesses in SPSA methodology were identified. None of these are new, all having been widely recognized for many years. However, for some of the weaknesses, extensive discussions during the Meeting provided

  15. Development of a Probabilistic Flood Hazard Assessment (PFHA) for the nuclear safety

    Science.gov (United States)

    Ben Daoued, Amine; Guimier, Laurent; Hamdi, Yasser; Duluc, Claire-Marie; Rebour, Vincent

    2016-04-01

    The purpose of this study is to lay the basis for a probabilistic evaluation of flood hazard (PFHA). Probabilistic assessment of external floods has become a current topic of interest to the nuclear scientific community. Probabilistic approaches complement deterministic approaches by exploring a set of scenarios and associating a probability to each of them. These approaches aim to identify all possible failure scenarios, combining their probability, in order to cover all possible sources of risk. They are based on the distributions of initiators and/or the variables caracterizing these initiators. The PFHA can characterize the water level for example at defined point of interest in the nuclear site. This probabilistic flood hazard characterization takes into account all the phenomena that can contribute to the flooding of the site. The main steps of the PFHA are: i) identification of flooding phenomena (rains, sea water level, etc.) and screening of relevant phenomena to the nuclear site, ii) identification and probabilization of parameters associated to selected flooding phenomena, iii) spreading of the probabilized parameters from the source to the point of interest in the site, v) obtaining hazard curves and aggregation of flooding phenomena contributions at the point of interest taking into account the uncertainties. Within this framework, the methodology of the PFHA has been developed for several flooding phenomena (rain and/or sea water level, etc.) and then implemented and tested with a simplified case study. In the same logic, our study is still in progress to take into account other flooding phenomena and to carry out more case studies.

  16. Optimized probabilistic calibration of safety coefficients in defect severity assessments; Dimensionnement probabiliste optimise des coefficients de securite dans les etudes de nocivite de defauts

    Energy Technology Data Exchange (ETDEWEB)

    Ardillon, E.; Pitner, P.; Barthelet, B. [Electricite de France, Direction des Etudes at Recherches, 92 - Clamart (France)

    1997-12-31

    The construction codes currently used in nuclear engineering recommend analysis methods and criteria consistent with a deterministic approach. Since 1993, in the framework of work related to the RSEM codes, the EFMT Branch has launched a probabilistic approach to establish a link between the current `deterministic` rules and failure risk assessments for the structures considered. There is an explicit link between the two approaches in the elementary strength/load case where the variables are Gaussian. This case provides the basis for the proposed methodology. In the complex case discussed in this paper, involving cracked piping with numerous non-Gaussian inputs, for a given failure mode, there is an implicit relationship between the target reliability level and the partial safety coefficients attached to each variable. The mean flaw size is the intermediate parameter used to make this link and allows flexibility in the choice of coefficients, thereby raising the question of optimized calibration. The approach is illustrated by the choice of coefficients based on the coordinates of the most probable failure point, resulting in a single set of coefficients adapted to the immediate vicinity of a given situation. In cases where the criterion must guarantee a given reliability level for a number of different operating situations, no set of coefficients can entirely guarantee the target reliability level. So, an optimized set of coefficients has to be selected, ensuring a reliability level as uniform as possible over the scope considered. This paper compares an initial coefficient proposal with a choice based on the design point method. The intermediate variable in assessing the reliability level is the mean flaw size, which would seem compatible with problems encountered under operating conditions. In addition, realistic risk assessment requires validation of the main variable distribution assumptions. We give an example of adjustment of distribution assumptions to

  17. The Integrated Safety Management System Verification Enhancement Review of the Plutonium Finishing Plant (PFP)

    International Nuclear Information System (INIS)

    BRIGGS, C.R.

    2000-01-01

    The primary purpose of the verification enhancement review was for the DOE Richland Operations Office (RL) to verify contractor readiness for the independent DOE Integrated Safety Management System Verification (ISMSV) on the Plutonium Finishing Plant (PFP). Secondary objectives included: (1) to reinforce the engagement of management and to gauge management commitment and accountability; (2) to evaluate the ''value added'' benefit of direct public involvement; (3) to evaluate the ''value added'' benefit of direct worker involvement; (4) to evaluate the ''value added'' benefit of the panel-to-panel review approach; and, (5) to evaluate the utility of the review's methodology/adaptability to periodic assessments of ISM status. The review was conducted on December 6-8, 1999, and involved the conduct of two-hour interviews with five separate panels of individuals with various management and operations responsibilities related to PFP. A semi-structured interview process was employed by a team of five ''reviewers'' who directed open-ended questions to the panels which focused on: (1) evidence of management commitment, accountability, and involvement; and, (2) consideration and demonstration of stakeholder (including worker) information and involvement opportunities. The purpose of a panel-to-panel dialogue approach was to better spotlight: (1) areas of mutual reinforcement and alignment that could serve as good examples of the management commitment and accountability aspects of ISMS implementation, and, (2) areas of potential discrepancy that could provide opportunities for improvement. In summary, the Review Team found major strengths to include: (1) the use of multi-disciplinary project work teams to plan and do work; (2) the availability and broad usage of multiple tools to help with planning and integrating work; (3) senior management presence and accessibility; (4) the institutionalization of worker involvement; (5) encouragement of self-reporting and self

  18. The importance of Probabilistic Safety Assessment in the careful study of risks involved to new nuclear power plant projects

    International Nuclear Information System (INIS)

    Mata, Jônatas F.C. da; Mesquita, Amir Z.

    2017-01-01

    The Fukushima Daiichi nuclear accident in Japan in 2011 has raised public fears about the actual safety of nuclear power plants in several countries. The response to this concern by government agencies and private companies has been objective and pragmatic in order to guarantee best practices in the design, construction, operation and decommissioning phases of nuclear reactors. In countries where the nucleo-electric matrix is consolidated, such as the United States, France and the United Kingdom, the safety assessment is carried out considering deterministic and probabilistic criteria. In the licensing stages of new projects, it is necessary to analyze and simulate the behavior of the nuclear power plant, when subjected to conditions that can lead to sequences of accidents. Each initiator event is studied and simulated through computational models, which allow the description and estimation of possible physical phenomena occurring in nuclear reactors. Probabilistic Safety Assessment (PSA) is fundamental in this process, as it studies in depth the sequences of events that can lead to the fusion of the nucleus of the nuclear reactor. Such sequences should be quantified in terms of probability of occurrence and your possible consequences, and organized through techniques such as Fault Tree Analysis and Event Tree Analysis. For these simulations, specialized computer codes for each type of phenomenon should be used, as well as databases based on experience gained in the operation of similar nuclear reactors. The present work will describe, in an objective way, the procedures for the realization of PSA and its applicability to the assurance of the operational reliability of the nuclear reactors, as well as a brief comparative between the approaches used in some countries traditionally users of thermonuclear energy and Brazil. By means of this analysis, it can be concluded that nuclear power is increasingly reliable and safe, being able to provide the necessary

  19. The importance of Probabilistic Safety Assessment in the careful study of risks involved to new nuclear power plant projects

    Energy Technology Data Exchange (ETDEWEB)

    Mata, Jônatas F.C. da, E-mail: jonatasfmata@yahoo.com.br [Universidade do Estado de Minas Gerais (UEMG), João Monlevade, MG (Brazil); Mesquita, Amir Z., E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The Fukushima Daiichi nuclear accident in Japan in 2011 has raised public fears about the actual safety of nuclear power plants in several countries. The response to this concern by government agencies and private companies has been objective and pragmatic in order to guarantee best practices in the design, construction, operation and decommissioning phases of nuclear reactors. In countries where the nucleo-electric matrix is consolidated, such as the United States, France and the United Kingdom, the safety assessment is carried out considering deterministic and probabilistic criteria. In the licensing stages of new projects, it is necessary to analyze and simulate the behavior of the nuclear power plant, when subjected to conditions that can lead to sequences of accidents. Each initiator event is studied and simulated through computational models, which allow the description and estimation of possible physical phenomena occurring in nuclear reactors. Probabilistic Safety Assessment (PSA) is fundamental in this process, as it studies in depth the sequences of events that can lead to the fusion of the nucleus of the nuclear reactor. Such sequences should be quantified in terms of probability of occurrence and your possible consequences, and organized through techniques such as Fault Tree Analysis and Event Tree Analysis. For these simulations, specialized computer codes for each type of phenomenon should be used, as well as databases based on experience gained in the operation of similar nuclear reactors. The present work will describe, in an objective way, the procedures for the realization of PSA and its applicability to the assurance of the operational reliability of the nuclear reactors, as well as a brief comparative between the approaches used in some countries traditionally users of thermonuclear energy and Brazil. By means of this analysis, it can be concluded that nuclear power is increasingly reliable and safe, being able to provide the necessary

  20. Electric and mechanical basic parameters to elaborate a process for a technical verification of safety related design modifications

    International Nuclear Information System (INIS)

    Lamuno Fernandez, Mercedes; La Roca Mallofre, GISEL; Bano Azcon, Alberto

    2010-01-01

    This paper presents a systematic process to check a design in order to achieve all the requirements that regulations demand. Nuclear engineers must verify that a design is done according to the safety requirements, and this paper presents how we have elaborated a process to improve the technical project verification. For a faster, better and easier verification process, here we summarize how to select the electric and mechanical basic parameters, which ensure the correct project verification of safety related design modifications. This process considers different aspects, which guarantee that the design preserves the availability, reliability and functional capability of the Structures, Systems and Components needed to operate the Nuclear Power Station with security. Electric and mechanical reference parameters are identified and discussed as well as others related ones, which are critical to safety. The implementation procedure to develop tasks performed in any company that has a quality plan is a requirement. On the engineering business, it is important not to use the personal criteria to do a technical analysis of a project; although, many times it is the checker's criteria and knowledge responsibility to ensure the correct development of a design modification. Then, the checker capabilities are the basis of the modification verification. This kind of procedure's development is not easy, because in an engineering project with important technical contents, there are multiple scenarios, but lots of them have a common basis. If we can identify the technical common basis of these projects, we will make good project verification but there are many difficulties we can encounter along this process. (authors)

  1. Use of cut-off values as meaningfulness limits in probabilistic studies and its effect on NPPs risk assessment and safety improvement

    International Nuclear Information System (INIS)

    Petrangeli, G.; Valeri, A.; Zaffiro, C.

    1991-01-01

    This paper discusses the use of cut-off values in probabilistic risk assessment/probabilistic safety assessment (PRA/PSA) of nuclear power plants (NPPs), in order to explore under which conditions this practice may help improve the meaningfulness of the results of the analyses and safety of plants, and how it may affect the assessment of risk. Reference is made, in particular, to some past practical applications, also taken from the experience of the authors within the frame of the Italian licensing process. The paper describes the Italian probabilistic criteria which use probabilistic targets and cut-off values to assess safety and identify plant safety improvements. The rationale of the approach is also discussed in the paper and results of sample applications are illustrated. The paper concludes that the use of cut-off values, if properly implemented, could be productive to improve the plant safety as it helps the analyst to focus on a restricted field of analysis, ignoring lower probability and less known events. It also points out that cut-off values should be considered as living numbers to be lowered and even eliminated as soon as significant advancements are made, through research and operational experience, in the knowledge of the pertinent events

  2. Use of probabilistic safety assessment in supporting regulatory authority`s work; Todennaekoeisyyspohjaisen turvallisuusanalyysin kaeyttoe viranomaistyoen tukena

    Energy Technology Data Exchange (ETDEWEB)

    Julin, A

    1995-11-01

    The aim of the study was to examine possibilities to use probabilistic safety assessment (PSA) more effectively in regulatory control of nuclear power plants. The structure, results and evaluation methods of PSA along with the necessary equations and principles, which could be used in utilising level 1 PSA results in decision making, have been introduced. The presented examples describe the ways PSA has been utilised abroad and particularly in Finnish Centre for Radiation and Nuclear Safety (STUK). The examples calculated in the study are based on the SPSA code and the PSA model of Olkiluoto nuclear power plant (TVO). The examples compare component safety classes versus safety importance and the risk of continued operation versus shutdown alternative in residual heat removal system failures. In addition to this allowed outage times, as calculated by PSA, were compared to allowed outage times according to technical specifications. The last 9 years operating experiences of TVO II was also examined by analysing the risk importance of significant component failures and operational disturbances. The analysis showed that the contribution of component failures and operational disturbances to the overall core damage risk during the studied time period was only 5 per cent. It appeared that the rare, significant initiating events provide the main contribution to the total cumulative risk. (57 refs., 22 figs., 17 tabs.).

  3. Peer review of the Barselina Level 1 probabilistic safety assessment of the Ignalina Nuclear Power Plant, Unit 2

    International Nuclear Information System (INIS)

    McKay, S.L.; Coles, G.A.

    1995-01-01

    The Barselina Project is a Swedish-funded, cooperative effort among Lithuania, Russia and Sweden to transfer Western probabilistic safety assessment (PSA) methodology to the designers/operators of Ignalina Nuclear Power Plant (INPP). The overall goal is to use the PSA as a tool for assessing plant operational safety. The INPP is a two-unit, Former Soviet Union-designed nuclear facility located in Lithuania. The results of this PSA will ultimately be used to identify plant-specific improvements in system design and the conduct of facility operations, allowing improved operational safety. Pacific Northwest Laboratory (PNL) was asked to perform an independent expert peer review of the Barselina PSA. This report documents the findings of this review. This review, financed with nuclear safety assistance funds through the US Agency for International Development (USAID) and the US Department of Energy (DOE), satisfies Task II of the PNL peer review of the Barselina project. The objective is to provide an independent, in-proce ss examination of the Barselina Level 1 PSA of Ignalina Nuclear Power Plant, Unit 2. The review consisted of an investigation of the project documentation, interviews, and extensive discussions with the PSA staff during critical stages of the project. PNL assessed the readability, completeness, consistency, validity, and applicability of the PSA. The major aspects explored were its purpose, major assumptions, analysis/modeling, results, and interpretation. It was not within the scope of this review to perform plant walkdowns or to review material other than the PSA documentation

  4. Report of the IPERS (International Peer Review Service) pre-review mission for the Cernavoda nuclear power plant probabilistic safety evaluation (CPSE - PHASE B) in Romania 31 October to 3 November 1994

    International Nuclear Information System (INIS)

    1994-01-01

    This report presents the results of the IAEA international peer review services pre-review mission which reviewed the status of the present version of the Cernavoda probabilistic safety evaluation, a Level 1 internal events Probabilistic Safety Assessment for the Cernavoda, Unit 1, nuclear power plant. 2 refs

  5. Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea.

  6. Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

    International Nuclear Information System (INIS)

    Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun

    2015-01-01

    IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea

  7. Epistemic and aleatory uncertainties in integrated deterministic and probabilistic safety assessment: Tradeoff between accuracy and accident simulations

    International Nuclear Information System (INIS)

    Karanki, D.R.; Rahman, S.; Dang, V.N.; Zerkak, O.

    2017-01-01

    The coupling of plant simulation models and stochastic models representing failure events in Dynamic Event Trees (DET) is a framework used to model the dynamic interactions among physical processes, equipment failures, and operator responses. The integration of physical and stochastic models may additionally enhance the treatment of uncertainties. Probabilistic Safety Assessments as currently implemented propagate the (epistemic) uncertainties in failure probabilities, rates, and frequencies; while the uncertainties in the physical model (parameters) are not propagated. The coupling of deterministic (physical) and probabilistic models in integrated simulations such as DET allows both types of uncertainties to be considered. However, integrated accident simulations with epistemic uncertainties will challenge even today's high performance computing infrastructure, especially for simulations of inherently complex nuclear or chemical plants. Conversely, intentionally limiting computations for practical reasons would compromise accuracy of results. This work investigates how to tradeoff accuracy and computations to quantify risk in light of both uncertainties and accident dynamics. A simple depleting tank problem that can be solved analytically is considered to examine the adequacy of a discrete DET approach. The results show that optimal allocation of computational resources between epistemic and aleatory calculations by means of convergence studies ensures accuracy within a limited budget. - Highlights: • Accident simulations considering uncertainties require intensive computations. • Tradeoff between accuracy and accident simulations is a challenge. • Optimal allocation between epistemic & aleatory computations ensures the tradeoff. • Online convergence gives an early indication of computational requirements. • Uncertainty propagation in DDET is examined on a tank problem solved analytically.

  8. Modelling software failures of digital I and C in probabilistic safety analyses based on the TELEPERM registered XS operating experience

    International Nuclear Information System (INIS)

    Jockenhoevel-Barttfeld, Mariana; Taurines Andre; Baeckstroem, Ola; Holmberg, Jan-Erik; Porthin, Markus; Tyrvaeinen, Tero

    2015-01-01

    Digital instrumentation and control (I and C) systems appear as upgrades in existing nuclear power plants (NPPs) and in new plant designs. In order to assess the impact of digital system failures, quantifiable reliability models are needed along with data for digital systems that are compatible with existing probabilistic safety assessments (PSA). The paper focuses on the modelling of software failures of digital I and C systems in probabilistic assessments. An analysis of software faults, failures and effects is presented to derive relevant failure modes of system and application software for the PSA. The estimations of software failure probabilities are based on an analysis of the operating experience of TELEPERM registered XS (TXS). For the assessment of application software failures the analysis combines the use of the TXS operating experience at an application function level combined with conservative engineering judgments. Failure probabilities to actuate on demand and of spurious actuation of typical reactor protection application are estimated. Moreover, the paper gives guidelines for the modelling of software failures in the PSA. The strategy presented in this paper is generic and can be applied to different software platforms and their applications.

  9. The End-To-End Safety Verification Process Implemented to Ensure Safe Operations of the Columbus Research Module

    Science.gov (United States)

    Arndt, J.; Kreimer, J.

    2010-09-01

    The European Space Laboratory COLUMBUS was launched in February 2008 with NASA Space Shuttle Atlantis. Since successful docking and activation this manned laboratory forms part of the International Space Station(ISS). Depending on the objectives of the Mission Increments the on-orbit configuration of the COLUMBUS Module varies with each increment. This paper describes the end-to-end verification which has been implemented to ensure safe operations under the condition of a changing on-orbit configuration. That verification process has to cover not only the configuration changes as foreseen by the Mission Increment planning but also those configuration changes on short notice which become necessary due to near real-time requests initiated by crew or Flight Control, and changes - most challenging since unpredictable - due to on-orbit anomalies. Subject of the safety verification is on one hand the on orbit configuration itself including the hardware and software products, on the other hand the related Ground facilities needed for commanding of and communication to the on-orbit System. But also the operational products, e.g. the procedures prepared for crew and ground control in accordance to increment planning, are subject of the overall safety verification. In order to analyse the on-orbit configuration for potential hazards and to verify the implementation of the related Safety required hazard controls, a hierarchical approach is applied. The key element of the analytical safety integration of the whole COLUMBUS Payload Complement including hardware owned by International Partners is the Integrated Experiment Hazard Assessment(IEHA). The IEHA especially identifies those hazardous scenarios which could potentially arise through physical and operational interaction of experiments. A major challenge is the implementation of a Safety process which owns quite some rigidity in order to provide reliable verification of on-board Safety and which likewise provides enough

  10. On the use of data and judgment in probabilistic risk and safety analysis

    International Nuclear Information System (INIS)

    Kaplan, S.

    1986-01-01

    This paper reviews the line of thought of a nuclear plant probabilistic risk analysis (PRA) identifying the points where data and judgement enter. At the ''bottom'' of the process, data and judgment are combined, using one and two stage Bayesian methods, to express what is known about the element of variables. Higher in the process, we see the use of judgment in identifying scenarios and developing almost models and specifying initiating event categories. Finally, we discuss the judgments involved in deciding to do a PRA and in applying the results. (orig.)

  11. General-Purpose Heat Source development: Safety Verification Test Program. Bullet/fragment test series

    Energy Technology Data Exchange (ETDEWEB)

    George, T.G.; Tate, R.E.; Axler, K.M.

    1985-05-01

    The radioisotope thermoelectric generator (RTG) that will provide power for space missions contains 18 General-Purpose Heat Source (GPHS) modules. Each module contains four /sup 238/PuO/sub 2/-fueled clads and generates 250 W/sub (t)/. Because a launch-pad or post-launch explosion is always possible, we need to determine the ability of GPHS fueled clads within a module to survive fragment impact. The bullet/fragment test series, part of the Safety Verification Test Plan, was designed to provide information on clad response to impact by a compact, high-energy, aluminum-alloy fragment and to establish a threshold value of fragment energy required to breach the iridium cladding. Test results show that a velocity of 555 m/s (1820 ft/s) with an 18-g bullet is at or near the threshold value of fragment velocity that will cause a clad breach. Results also show that an exothermic Ir/Al reaction occurs if aluminum and hot iridium are in contact, a contact that is possible and most damaging to the clad within a narrow velocity range. The observed reactions between the iridium and the aluminum were studied in the laboratory and are reported in the Appendix.

  12. General-Purpose Heat Source Safety Verification Test program: Edge-on flyer plate tests

    International Nuclear Information System (INIS)

    George, T.G.

    1987-03-01

    The radioisotope thermoelectric generator (RTG) that will supply power for the Galileo and Ulysses space missions contains 18 General-Purpose Heat Source (GPHS) modules. The GPHS modules provide power by transmitting the heat of 238 Pu α-decay to an array of thermoelectric elements. Each module contains four 238 PuO 2 -fueled clads and generates 250 W(t). Because the possibility of a launch vehicle explosion always exists, and because such an explosion could generate a field of high-energy fragments, the fueled clads within each GPHS module must survive fragment impact. The edge-on flyer plate tests were included in the Safety Verification Test series to provide information on the module/clad response to the impact of high-energy plate fragments. The test results indicate that the edge-on impact of a 3.2-mm-thick, aluminum-alloy (2219-T87) plate traveling at 915 m/s causes the complete release of fuel from capsules contained within a bare GPHS module, and that the threshold velocity sufficient to cause the breach of a bare, simulant-fueled clad impacted by a 3.5-mm-thick, aluminum-alloy (5052-T0) plate is approximately 140 m/s

  13. Explosion overpressure test series: General-Purpose Heat Source development: Safety Verification Test program

    International Nuclear Information System (INIS)

    Cull, T.A.; George, T.G.; Pavone, D.

    1986-09-01

    The General-Purpose Heat Source (GPHS) is a modular, radioisotope heat source that will be used in radioisotope thermoelectric generators (RTGs) to supply electric power for space missions. The first two uses will be the NASA Galileo and the ESA Ulysses missions. The RTG for these missions will contain 18 GPHS modules, each of which contains four 238 PuO 2 -fueled clads and generates 250 W/sub (t)/. A series of Safety Verification Tests (SVTs) was conducted to assess the ability of the GPHS modules to contain the plutonia in accident environments. Because a launch pad or postlaunch explosion of the Space Transportation System vehicle (space shuttle) is a conceivable accident, the SVT plan included a series of tests that simulated the overpressure exposure the RTG and GPHS modules could experience in such an event. Results of these tests, in which we used depleted UO 2 as a fuel simulant, suggest that exposure to overpressures as high as 15.2 MPa (2200 psi), without subsequent impact, does not result in a release of fuel

  14. Verification of implementation of the radiological safety standards through the regulatory inspections

    International Nuclear Information System (INIS)

    Perez Gonzalez, Francisco; Fornet Rodriguez, Ofelia M.

    2008-01-01

    Full text: As an element of the updating process of the legal framework on radiological safety in Cuba, a new rule was put into force; the Radiological Basic Safety Standards (RBSS) in January 2002. Five years after the application of these new safety requirements, it was considered appropriate to assess the effectiveness of its implementation. Therefore, in this work the authors analysed the outcomes of the regulatory inspections conducted in this period upon medical and industrial practices in a sample of facilities representative of those with the highest radiological risks in the territory under supervision of a Territorial Delegation of the Nuclear Regulatory Authority. For better understanding of this presentation, a summary explanation of the structure of the rule is given in its introduction. The work was to identify for each deficiency, or finding, or counter-measure; out of the relevant inspections; the corresponding requirement/Article of the RBSS that shows difficulties in implementation. For each installation an analysis is made with regard to the relevant articles difficult to implement. Finally, the appraisal is shown separately for the medical practice, and for the industrial practice, and also in general for the whole sample of installations under review. The study showed that the implementation of the Standards has been satisfactory and uniform in the practices under review. So far it seems that there have not been major difficulties with the implementation of the Titles; III On Intervention, IV Dose Limits, as well as with the Especial, Final, and Transitory Dispositions. On the other hand, it is shown there is a need for continued work only with regard to the implementation of the requirements in Section IV Verification of Safety and in Section V On the responsibilities with regard to occupational exposure in Chapter III Title I, and correspondingly in Chapter II Occupational Exposure in Title II. It is recommended to conduct this kind of

  15. A study on the dependency evaluation for multiple human actions in human reliability analysis of probabilistic safety assessment

    International Nuclear Information System (INIS)

    Kang, D. I.; Yang, J. E.; Jung, W. D.; Sung, T. Y.; Park, J. H.; Lee, Y. H.; Hwang, M. J.; Kim, K. Y.; Jin, Y. H.; Kim, S. C.

    1997-02-01

    This report describes the study results on the method of the dependency evaluation and the modeling, and the limited value of human error probability (HEP) for multiple human actions in accident sequences of probabilistic safety assessment (PSA). THERP and Parry's method, which have been generally used in dependency evaluation of human reliability analysis (HRA), are introduced and their limitations are discussed. New dependency evaluation method in HRA is established to make up for the weak points of THERP and Parry's methods. The limited value of HEP is also established based on the review of several HRA related documents. This report describes the definition, the type, the evaluation method, and the evaluation example of dependency to help the reader's understanding. It is expected that this study results will give a guidance to HRA analysts in dependency evaluation of multiple human actions and enable PSA analysts to understand HRA in detail. (author). 23 refs., 3 tabs., 2 figs

  16. Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (1)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Sinsuke; Muramatsu, Ken

    2007-02-01

    Japan Atomic Energy Agency (JAEA) entrusted with research on the state-of-the-art of probabilistic safety assessment (PSA) for non-reactor nuclear facilities (NRNF) to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of 'research on the analysis methods for accident consequence in NFRF' was organized in the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. (author)

  17. Review and evaluation of the Millstone Unit 3 probabilistic safety study. Containment failure modes, radiological source - terms and offsite consequences

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Pratt, W.; Ludewig, H.

    1985-09-01

    A technical review and evaluation of the Millstone Unit 3 probabilistic safety study has been performed. It was determined that; (1) long-term damage indices (latent fatalities, person-rem, etc.) are dominated by late failure of the containment, (2) short-term damage indices (early fatalities, etc.) are dominated by bypass sequences for internally initiated events, while severe seismic sequences can also contribute significantly to early damage indices. These overall estimates of severe accident risk are extremely low compared with other societal sources of risk. Furthermore, the risks for Millstone-3 are comparable to risks from other nuclear plants at high population sites. Seismically induced accidents dominate the severe accident risks at Millstone-3. Potential mitigative features were shown not to be cost-effective for internal events. Value-impact analysis for seismic events showed that a manually actuated containment spray system might be cost-effective

  18. Methodology of containment response analysis for the Probabilistic Safety Assessment -PSA of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, Jorge

    1996-01-01

    This work is part of the Probabilistic Safety Assessment actually under development for the CAREM-25 Nuclear Power Station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during, severe accidents. Then Plan Damage States are then defined. Furthermore, Containment Damage States are also defined, and Containment Event Trees are built for each Plant Damage State. Those sequences considered representative from the annual probability (those which exceed or equal a probability of 1E-09 per year, are used to quantify the combinations of Plant Damage States/Containment Damage States, based on the estimation of a Vulnerability Matrix. (author)

  19. Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

    2010-03-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

  20. Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (2)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Sinsuke; Muramatsu, Ken

    2007-03-01

    Japan Atomic Energy Agency (JAEA) entrusted with a research on the state-of-the-art of probabilistic safety assessment (PSA) of non-reactor nuclear facilities (NRNF) such as fuel reprocessing and fuel fabrication facilities to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of 'Research on the analysis methods for accident consequence in NFRF' was organized by the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. This report summarizes the results of research conducted by the committee in FY 2005. (author)

  1. Living probabilistic safety assessment of French 1300 MWe PWR nuclear power plant unit: methodology, results and teaching

    International Nuclear Information System (INIS)

    Dubreuil Chambardel, A.; Villemeur, A.; Berger, J.P.; Moroni, J.M.

    1991-02-01

    Launched in 1986 by Electricite de France, the Probabilistic Safety Assessment of a French 1300 MWe Pressurized Water Reactor (called PSA 1300) was completed in 1989. The first objective was to assess the annual core damage frequency by identifying all the accident scenarii likely to contribute significantly to this frequency. The second objective of the study was to provide an automated computerized tool (software) for updating the assessment - in order to take new data and knowledge into account - and for performing numerous sensitivity studies easily. Its scope and characteristics render this study unique. Indeed, it required an effort amounting to 50 engineer-years. The results and the first lessons are presented in this paper. The PSA 1300 teachings will be extensively used for the design and operation of existing or future French nuclear power reactors

  2. Risk assessment methods in radiotherapy: Probabilistic safety assessment (PSA); Los metodos de analisis de riesgo en radioterapia: Analisis Probabilistico de seguridad (APS)

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez Vera, M. L.; Perez Mulas, A.; Delgado, J. M.; Barrientos Ontero, M.; Somoano, F.; Alvarez Garcia, C.; Rodriguez Marti, M.

    2011-07-01

    The understanding of accidents that have occurred in radiotherapy and the lessons learned from them are very useful to prevent repetition, but there are other risks that have not been detected to date. With a view to identifying and preventing such risks, proactive methods successfully applied in other fields, such as probabilistic safety assessment (PSA), have been developed. (Author)

  3. A Methods and procedures to apply probabilistic safety Assessment (PSA) techniques to the cobalt-therapy process. Cuban experience

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Ferro Fernandez, R.; Lozano Lima, B; De la Fuente Puch, A.; Dumenigo Gonzalez, C.; Troncoso Fleitas, M.; Perez Reyes, Y.

    2003-01-01

    This paper presents the results of the Probabilistic Safety Analysis (PSA) to the Cobalt Therapy Process, which was performed as part of the International Atomic Energy Agency's Coordinated Research Project (CRP) to Investigate Appropriate Methods and Procedures to Apply Probabilistic Safety Assessment (PSA) Techniques to Large Radiation Sources. The primary methodological tools used in the analysis were Failure Modes and Effects Analysis (FMEA), Event Trees and Fault Trees. These tools were used to evaluate occupational, public and medical exposures during cobalt therapy treatment. The emphasis of the study was on the radiological protection of patients. During the course of the PSA, several findings were analysed concerning the cobalt treatment process. In relation with the Undesired Events Probabilities, the lowest exposures probabilities correspond to the public exposures during the treatment process (Z21); around 10-10 per year, being the workers exposures (Z11); around 10-4 per year. Regarding to the patient, the Z33 probabilities prevail (not desired dose to normal tissue) and Z34 (not irradiated portion to target volume). Patient accidental exposures are also classified in terms of the extent to which the error is likely to affect individual treatments, individual patients, or all the patients treated on a specific unit. Sensitivity analyses were realised to determine the influence of certain tasks or critical stages on the results. As a conclusion the study establishes that the PSA techniques may effectively and reasonably determine the risk associated to the cobalt-therapy treatment process, though there are some weaknesses in its methodological application for this kind of study requiring further research. These weaknesses are due to the fact that the traditional PSA has been mainly applied to complex hardware systems designed to operate with a high automation level, whilst the cobalt therapy treatment is a relatively simple hardware system with a

  4. Millstone 3 risk evaluation report. An overall review and evaluation of the Millstone Unit 3 probabilistic safety study

    International Nuclear Information System (INIS)

    Kelly, G.; Barrett, R.; Buslik, A.

    1986-06-01

    In 1981, the US Nuclear Regulatory Commission (NRC) requested Northeast Utilities to perform a design-specific probabilistic safety study (PSS) for Millstone Nuclear Power Station, Unit No. 3 (Millstone 3). In 1983, Northeast Utilities submitted the Millstone 3 Probabilistic Safety Study for review by the NRC staff. The NRC staff prepared the Millstone 3 Risk Evaluation Report, which discusses the findings regarding the PSS. The PSS estimates that the mean annual core damage frequency due to internal and external events is 5 x 10 -5 and 2 x 10 -5 , respectively. The NRC staff's Risk Evaluation Report estimates that the mean annual core damage frequency is about 2 x 10 -4 for internal events and lies between 1 x 10 -5 and 2 x 10 -4 for external events. The NRC staff estimates that station blackout dominates internal and external event core damage frequencies. The staff recommends that Northeast Utilities perform an engineering analysis on upgrading the diesel generator lube oil cooler anchorage system and on adding a manually operated, AC-independent containment spray system. The staff also recommends that Northeast Utilities prepare two emergency procedures (loss of room cooling and relay chatter due to an earthquake) to help reduce uncertainties. (Subsequent to the completion of this document, Northeast Utilities and the NRC staff have continued a dialogue regarding station blackout from events other than earthquakes. Both Northeast Utilities and the staff have performed additional evaluations, which have drawn their results closer together. Final requirements, if any, for the prevention or mitigation of station blackout from events other than earthquakes have not yet been determined.) 26 refs., 16 tabs

  5. US Department of Energy Approach to Probabilistic Evaluation of Long-Term Safety for a Potential Yucca Mountain Repository

    International Nuclear Information System (INIS)

    Dr. R. Dyer; Dr. R. Andrews; Dr. A. Van Luik

    2005-01-01

    Regulatory requirements being addressed in the US geological repository program for spent nuclear fuel and high-level waste disposal specify probabilistically defined mean-value dose limits. These dose limits reflect acceptable levels of risk. The probabilistic approach mandated by regulation calculates a ''risk of a dose,'' a risk of a potential given dose value at a specific time in the future to a hypothetical person. The mean value of the time-dependent performance measure needs to remain below an acceptable level defined by regulation. Because there are uncertain parameters that are important to system performance, the regulation mandates an analysis focused on the mean value of the performance measure, but that also explores the ''full range of defensible and reasonable parameter distributions''...System performance evaluations should not be unduly influenced by...''extreme physical situations and parameter values''. Challenges in this approach lie in defending the scientific basis for the models selected, and the data and distributions sampled. A significant challenge lies in showing that uncertainties are properly identified and evaluated. A single-value parameter has no uncertainty, and where used such values need to be supported by scientific information showing the selected value is appropriate. Uncertainties are inherent in data, but are also introduced by creating parameter distributions from data sets, selecting models from among alternative models, abstracting models for use in probabilistic analysis, and in selecting the range of initiating event probabilities for unlikely events. The goal of the assessment currently in progress is to evaluate the level of risk inherent in moving ahead to the next phase of repository development: construction. During the construction phase, more will be learned to inform a new long-term risk evaluation to support moving to the next phase: accepting waste. Therefore, though there was sufficient confidence of safety

  6. Probabilistic safety assessment (PSA) of the radiotherapy treatment process with an Electron Linear Accelerator (LINAC) for medical uses

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Ferro Fernandez, R.; Rodriguez Marti, M.; Ramirez, M.L.; Perez Mulas, A.; Barrientos Montero, M.; Ortiz Lopez, P.; Somoano, F.; Delgado Rodriguez, J.M.; Papadopulos, S.B.; Pereira, P.P. Jr.; Lopez Morones, R.; Larrinaga Cortinai, E.; Rivero Oliva, J.J.; Alemany, J.

    2008-01-01

    This paper presents the results of the Probabilistic Safety Assessment (PSA) to the radiotherapy treatment process with an Electron Linear Accelerator (LINAC) for Medical Uses, which was conducted in the framework of the Extra budgetary Programme on Nuclear and Radiological Safety in Iberian-America. The PSA tools were used to evaluate occupational, public and medical exposures during treatment. The study focused on the radiological protection of patients. Equipment Failure Modes and Human Errors were evaluated for each system and treatment phase by FMEA. It was aimed at obtaining an exhaustive list of deviations with a reasonable probability of occurrence and which might produce significant adverse outcomes. Separate events trees were constructed for each initiating event group. Each event tree had a different structure since the initiating events were grouped according to mitigation requirements. Fault tree models were constructed for each top event. The fault trees were developed up to the level of components. In addition to hardware faults, the fault trees included human errors associated with the response to accidents, and human errors associated with the treatment. Each accident sequence was quantified. The combination of the initiating event and top events through one fault tree was the method used to analyse the accident sequences. After combining the appropriate models, a Boolean reduction was conducted by computer software to produce sequence cut sets. Several findings were analysed concerning the treatment process and the study proposed safety recommendations to avoid them. (author)

  7. Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z., E-mail: raissaomarques@gmail.com, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

  8. Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

    International Nuclear Information System (INIS)

    Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z.

    2017-01-01

    Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

  9. Safety evaluation for the LMFBR plant using probabilistic risk assessment techniques

    International Nuclear Information System (INIS)

    Kani, Y.; Aizawa, K.

    1987-01-01

    This paper presents an application of probabilistic risk assessment techniques to a typical loop-type liquid metal fast breeder reactor (LMFBR) plant in the detailed design stage. A comprehensive systems analysis has been performed to identify event sequences leading to core damage and provide insights into the importance of accident contributors. While traditional event tree/fault tree modeling was used for the analysis, this study involved a thorough investigation of initiating events and of support system faults. The qualification of accident sequences has been conducted by combining the fault trees based on the event trees and obtaining sequence cut sets with the use of the SETS code. This study also attempted to quantify the potential for operator recovery actions in the course of each accident sequence. (author)

  10. Methodological Development of the Probabilistic Model of the Safety Assessment of Hontomin P.D.T

    International Nuclear Information System (INIS)

    Hurtado, A.; Eguilior, S.; Recreo, F.

    2011-01-01

    In the framework of CO 2 Capture and Geological Storage, Risk Analysis plays an important role, because it is an essential requirement of knowledge to make up local, national and supranational definition and planning of carbon injection strategies. This is because each project is at risk of failure. Even from the early stages, it should take into account the possible causes of this risk and propose corrective methods along the process, i.e., managing risk. Proper risk management reduces the negative consequences arising from the project. The main method of reduction or neutralizing of risk is mainly the identification, measurement and evaluation of it, together with the development of decision rules. This report presents the developed methodology for risk analysis and the results of its application. The risk assessment requires determination of the random variables that will influence the functioning of the system. It is very difficult to set up probability distribution of a random variable in the classical sense (objective probability) when a particular event rarely occurred or even it has a incomplete development. In this situation, we have to determine the subjective probability, especially at an early stage of projects, when we have not enough information about the system. This subjective probability is constructed from assessment of experts judgement to estimate the possibility of certain random events could happen depending on geological features of the area of application. The proposed methodology is based on the application of Bayesian Probabilistic Networks for estimating the probability of risk of leakage. These probabilistic networks can define graphically relations of dependence between the variables and joint probability function through a local factorization of probability functions. (Author) 98 refs.

  11. Methods and practices for verification and validation of programmable systems

    International Nuclear Information System (INIS)

    Heimbuerger, H.; Haapanen, P.; Pulkkinen, U.

    1993-01-01

    The programmable systems deviate by their properties and behaviour from the conventional non-programmable systems in such extent, that their verification and validation for safety critical applications requires new methods and practices. The safety assessment can not be based on conventional probabilistic methods due to the difficulties in the quantification of the reliability of the software and hardware. The reliability estimate of the system must be based on qualitative arguments linked to a conservative claim limit. Due to the uncertainty of the quantitative reliability estimate other means must be used to get more assurance about the system safety. Methods and practices based on research done by VTT for STUK, are discussed in the paper as well as the methods applicable in the reliability analysis of software based safety functions. The most essential concepts and models of quantitative reliability analysis are described. The application of software models in probabilistic safety analysis (PSA) is evaluated. (author). 18 refs

  12. Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

  13. Integrated Safety Management System Phase 1 and 2 Verification for the Environmental Restoration Contractor Volumes 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    CARTER, R.P.

    2000-04-04

    DOE Policy 450.4 mandates that safety be integrated into all aspects of the management and operations of its facilities. The goal of an institutionalized Integrated Safety Management System (ISMS) is to have a single integrated system that includes Environment, Safety, and Health requirements in the work planning and execution processes to ensure the protection of the worker, public, environment, and the federal property over the life cycle of the Environmental Restoration (ER) Project. The purpose of this Environmental Restoration Contractor (ERC) ISMS Phase MI Verification was to determine whether ISMS programs and processes were institutionalized within the ER Project, whether these programs and processes were implemented, and whether the system had promoted the development of a safety conscious work culture.

  14. Screening of external hazards for NPP with bank type reactor. Modeling of safety related systems and equipment for RBMK. Probabilistic assessment of NPP safety on aircraft impact. Progress report

    International Nuclear Information System (INIS)

    Kostarev, V.

    1999-01-01

    This progress report was produced within the frame of IAEA research project on screening the hazards for NPP with bank type reactor. It covers the following tasks; development of the model for the primary loop system of RBMK; developing the models for safety related equipment of RBMK; developing of models for safety related models of EGP-6 type reactor (Bilibinskaya Nuclear Co-generated heat and Power Plant); and probabilistic assessment of NPP safety on aircraft impact

  15. Main of probabilistic safety assessment (PSA) of the radiotherapy treatment process with a linear accelerator for medical purposes (linac)

    International Nuclear Information System (INIS)

    Vilaragut Llanes, Juan Jose; Fernandez, Ruben Ferro; Ortiz Lopez, Pedro

    2009-01-01

    The radiation safety assessments traditionally have been based on analyzing the lessons you learn of new events that are becoming known. Although these methods are very valuable, their main limitation is that only cover known events and leave without consider other possible failures that have occurred or have not been published, This does not mean they can not occur. Other tools to analyze prospectively the safety, among which found Probabilistic Safety Assessment (PSA). This paper summarizes the project of American Forum of agencies radiological and nuclear regulators aimed at applying the methods of APS treatment process with a linear accelerator. We defined as unintended consequences accidental exposures both single patient and multiple patients. FMEA methodology was used to define events initiators of accidents and methods of event trees and trees failure to identify the accident sequences that may occur. A Once quantified the frequency of occurrence of accidental sequences Analyses of importance in determining the most recent events significant from the point of view of safety. We identified 158 of equipment failure modes and 295 errors human if they occurred would have the potential to cause the accidental exposures defined. We studied 118 of initiating events accident and 120 barriers. We studied 434 accident sequences. The accidental exposure of a single patient were 40 times likely that multiple patients. 100% of the total frequency of accidental exposures on a single patient is caused by human errors . 8% of the total frequency of accidental exposures on multiple patients initiating events may occur by equipment failure (Computerized tomography, treatment planning system, throttle linear) and 92% by human error. As part of the and recommendations of the study presents the events that are more contribution on the reduction of risk of accidental exposure. (author)

  16. Review of Risk Reduction Methods using Probabilistic Safety Assessment Insights and Improved Technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eun-Chan; Choi, Byung-Pil [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    As seen in the process of the periodic safety review of domestic nuclear power plants, the risk management objectives such as core damage frequency and large early release frequency are not easy to be met without continuous safety improvements and the integratoin of the improved technologies into the PSA evaluation methodologies. Because external event analyses have a protion of uncertainty factors in the current analysis methodologies, the technical efforts in various perspectives.

  17. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

    2017-03-15

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

  18. Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

    International Nuclear Information System (INIS)

    Groth, Katrina; Wang Chengdong; Mosleh, Ali

    2010-01-01

    This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

  19. Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

    Energy Technology Data Exchange (ETDEWEB)

    Groth, Katrina, E-mail: kgroth@umd.ed [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States); Wang Chengdong; Mosleh, Ali [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States)

    2010-12-15

    This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

  20. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    International Nuclear Information System (INIS)

    Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

    2017-01-01

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective

  1. Two viewpoints for software failures and their relation in probabilistic safety assessment of digital instrumentation and control systems

    International Nuclear Information System (INIS)

    Kim, Man Cheol

    2015-01-01

    As the use of digital systems in nuclear power plants increases, the reliability of the software becomes one of the important issues in probabilistic safety assessment. In this paper, two viewpoints for a software failure during the operation of a digital system or a statistical software test are identified, and the relation between them is provided. In conventional software reliability analysis, a failure is mainly viewed with respect to the system operation. A new viewpoint with respect to the system input is suggested. The failure probability density functions for the two viewpoints are defined, and the relation between the two failure probability density functions is derived. Each failure probability density function can be derived from the other failure probability density function by applying the derived relation between the two failure probability density functions. The usefulness of the derived relation is demonstrated by applying it to the failure data obtained from the software testing of a real system. The two viewpoints and their relation, as identified in this paper, are expected to help us extend our understanding of the reliability of safety-critical software. (author)

  2. The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

    International Nuclear Information System (INIS)

    Hill, D.J.; Ragland, W.A.; Roglans, J.

    1998-01-01

    This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

  3. The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D J; Ragland, W A; Roglans, J

    1998-11-01

    This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10{sup -6} yr{sup -1}, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10{sup -6} yr{sup -1} and the contribution of seismic events is 1.7 10{sup -5} yr{sup -1}. Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

  4. The EBR-II probabilistic risk assessment lessons learned regarding passive safety

    International Nuclear Information System (INIS)

    Hill, D.J.; Ragland, W.A.; Roglans, J.

    1994-01-01

    This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

  5. State of the art on the probabilistic safety assessment (P.S.A.); Etat de l'art sur les etudes probabilistes de surete (E.P.S.)

    Energy Technology Data Exchange (ETDEWEB)

    Devictor, N.; Bassi, A.; Saignes, P.; Bertrand, F

    2008-07-01

    The use of Probabilistic Safety Assessment (PSA) is internationally increasing as a means of assessing and improving the safety of nuclear and non-nuclear facilities. To support the development of a competence on Probabilistic Safety Assessment, a set of states of the art regarding these tools and their use has been made between 2001 and 2005, in particular on the following topics: - Definition of the PSA of level 1, 2 and 3; - Use of PSA in support to design and operation of nuclear plants (risk-informed applications); - Applications to Non Reactor Nuclear Facilities. The report compiled in a single document these states of the art in order to ensure a broader use; this work has been done in the frame of the Project 'Reliability and Safety of Nuclear Facility' of the Nuclear Development and Innovation Division of the Nuclear Energy Division. As some of these states of the art have been made in support to exchanges with international partners and were written in English, a section of this document is written in English. This work is now applied concretely in support to the design of 4. Generation nuclear systems as Sodium-cooled Fast Reactors and especially Gas-cooled Fast Reactor, that have been the subject of communications during the conferences ANS (Annual Meeting 2007), PSA'08, ICCAP'08 and in the journal Science and Technology of Nuclear Installations. (authors)

  6. Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Perdomo, Manuel

    1995-01-01

    Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential 'weak points' at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs

  7. OPAD: An expert system for research reactor operations and fault diagnosis using probabilistic safety assessment tools

    International Nuclear Information System (INIS)

    Verma, A.K.; Varde, P.V.; Sankar, S.; Prakash, P.

    1996-01-01

    A prototype Knowledge Based (KB) operator Adviser (OPAD) system has been developed for 100 MW(th) Heavy Water moderated, cooled and Natural Uranium fueled research reactor. The development objective of this system is to improve reliability of operator action and hence the reactor safety at the time of crises as well as normal operation. The jobs performed by this system include alarm analysis, transient identification, reactor safety status monitoring, qualitative fault diagnosis and procedure generation in reactor operation. In order to address safety objectives at various stages of the Operator Adviser (OPAD) system development the Knowledge has been structured using PSA tools/information in an shell environment. To demonstrate the feasibility of using a combination of KB approach with PSA for operator adviser system, salient features of some of the important modules (viz. FUELEX, LOOPEX and LOCAEX) have been discussed. It has been found that this system can serve as an efficient operator support system

  8. Probabilistic logics and probabilistic networks

    CERN Document Server

    Haenni, Rolf; Wheeler, Gregory; Williamson, Jon; Andrews, Jill

    2014-01-01

    Probabilistic Logic and Probabilistic Networks presents a groundbreaking framework within which various approaches to probabilistic logic naturally fit. Additionally, the text shows how to develop computationally feasible methods to mesh with this framework.

  9. Probabilistic safety criteria for improvement of Nuclear Power Plant design and operation

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Chung, Woo Sick; Park, Moon Kyu [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The procedure of this study is to : research on the status of IAEA(International Atomic Energy Agency) member states about the policy of safety goals, study figures of merit and demerit that inherently exist in the existing methodology for reliability allocation, develop an efficient methodology for allocating reliability from top-level safety goals to intermediate and low-level PSC, write a computer code on the basis of the methodology proposed in the study, and apply the methodology to Surry Unit 1 that is the type of PWR.

  10. Probabilistic analysis of safety in industrial irradiation plants; Analisis probabilistico de seguridad en plantas industriales de irradiacion

    Energy Technology Data Exchange (ETDEWEB)

    Alderete, F.; Elechosa, C. [Autoridad Regulatoria Nuclear, Av. del Libertador 8250 - Buenos Aires (Argentina)]. e-mail: falderet@sede.arn.gov.ar

    2006-07-01

    The Argentinean Nuclear Regulatory Authority is carrying out the Probabilistic Safety Analysis (PSA) of the two industrial irradiation plants existent in the country. The objective of this presentation is to show from the regulatory point of view, the advantages of applying this tool, as well as the appeared difficulties; for it will be made a brief description of the facilities, of the method and of the normative one. Both plants are multipurpose facilities classified as 'industrial irradiator category IV' (panoramic irradiator with source deposited in pool). Basically, the execution of an APS consists of the following stages: 1. Identification of initiating events. 2. Modeling of Accidental Sequences (Event Trees). 3. Analysis of Systems (Fault trees). 4. Quantification of Accidental Sequences. The argentine normative doesn't demand to these facilities the realization of an APS, however the basic standard of Radiological Safety establishes that in the design of this type of facilities in the cases that is justified, should make sure that the annual probability of occurrence of an accidental sequence and the resulting dose in a person gives as result an radiological risk inferior to the risk limit adopted as acceptance criteria. On the other hand the design standard specifies for these irradiators it demands a maximum fault rate of 10{sup -2} for the related components with the systems of radiological safety. In our case, the possible initiating events have been identified that carried out to not wanted situations (about people exposure, radioactive contamination). Then, for each one of the significant initiating events, the corresponding accidental sequences were modeled and the safety systems that intervene in this sequences by means of fault trees were analyzed, for then to determine the fault probabilities of the same ones. At the moment they are completing these fault trees, but the difficulty resides in the impossibility of obtaining real data

  11. The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

    International Nuclear Information System (INIS)

    Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

    2007-01-01

    The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods

  12. On the meaning of probability in the context of probabilistic safety assessment

    International Nuclear Information System (INIS)

    Oestberg, G.

    1988-01-01

    Assessments of reliability and safety in technology require the application not only of frequentistic statistics but also of subjective estimates of probabilities. This is true in particular for decision-making about complex systems made up of both 'hard' and 'soft'elements. Problems occur when objective and subjective considerations have to be integrated and accommodated to the decision-makers' mental models. (author)

  13. Applied probabilistic methods in the field of reactor safety in Germany

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1982-01-01

    Some aspects of applied reliability and risk analysis methods in nuclear safety and the present role of both in Germany, are discussed. First, some comments on the status and applications of reliability analysis are given. Second, some conclusions that can be drawn from previous work on the German Risk Study are summarized. (orig.)

  14. Probabilistic approaches to LCO's and surveillance requirements for standby safety systems

    International Nuclear Information System (INIS)

    Lofgren, E.V.; Varcolik, F.

    1982-11-01

    Results are presented for a comprehensive analysis of risk-based methods for establishing Limiting Conditions for Operation (LCO) and surveillance requirements for on-line test and repair of nuclear power plant safety system components. Limiting Conditions for Operation refers to the legal constraint on safety system component outage times that are imposed by the NRC as part of the reactor operating license. Generally, when a safety system component is removed for repair or test for a period of time there is a period of increased vulnerability concerning the probability that the affected safety system will be available to mitigate an accident. This period of increased vulnerability exists until the component is restored to service. The constraint on the duration of this period, the allowed outage time (AOT), is the aspect of LCOs that is of interest here. In particular, methods are reviewed and developed that relate measures of risk to the AOT. Only by explicitly relating risk to AOT can outage times be constrained by placing limits on risk. Methods developed for relating risk measures to outage times are presented. The review and analysis of risk related methods for establishing LCOs are described

  15. Nuclear power and probabilistic safety assessment (PSA): past through future applications

    Science.gov (United States)

    Stamatelatos, M. G.; Moieni, P.; Everline, C. J.

    1995-03-01

    Nuclear power reactor safety in the United States is about to enter a new era -- an era of risk- based management and risk-based regulation. First, there was the age of `prescribed safety assessment,' during which a series of design-basis accidents in eight categories of severity, or classes, were postulated and analyzed. Toward the end of that era, it was recognized that `Class 9,' or `beyond design basis,' accidents would need special attention because of the potentially severe health and financial consequences of these accidents. The accident at Three Mile Island showed that sequences of low-consequence, high-frequency events and human errors can be much more risk dominant than the Class 9 accidents. A different form of safety assessment, PSA, emerged and began to gain ground against the deterministic safety establishment. Eventually, this led to the current regulatory requirements for individual plant examinations (IPEs). The IPEs can serve as a basis for risk-based regulation and management, a concept that may ultimately transform the U.S. regulatory process from its traditional deterministic foundations to a process predicated upon PSA. Beyond the possibility of a regulatory environment predicated upon PSA lies the possibility of using PSA as the foundation for managing daily nuclear power plant operations.

  16. Cuban experience in verification of the execution of the safety requirements during the transport of radioactive materials

    International Nuclear Information System (INIS)

    Quevedo Garcia, J.R.; Lopez Forteza, Y.

    2001-01-01

    The Cuban Regulatory Authority has paid special attention to the verification of the execution of the safety requirements during the transport of radioactive material in the country. With this purpose, the Authority has followed a consequent policy based on supplementary demands to those collections in the juridical mark settled down in 1987 in the sphere of transport of radioactive substances. In the work the technical approaches are exposed kept in mind when establishing the one referred politics, the current situation is characterized, the results are evaluated obtained in correspondence with the pursued objectives and the essential aspects are exposed to keep in mind for the adopted politics ulterior development. (author)

  17. Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Perez, S.S.

    1987-01-01

    The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

  18. Evaluation of replacement tritium facility (RTF) compliance with DOE safety goals using probabilistic consequence assessment methodology

    International Nuclear Information System (INIS)

    O'Kula, K.R.; East, J.M.; Moore, M.L.

    1993-01-01

    The Savannah River Site (SRS), operated by the Westinghouse Savannah River Company (WSRC) for the US Department of Energy (DOE), is a major center for the processing of nuclear materials for national defense, deep-space exploration, and medical treatment applications in the United States. As an integral part of the DOE's effort to modernize facilities, implement improved handling and processing technology, and reduce operational risk to the general public and onsite workers, transition of tritium processing at SRS from the Consolidated Tritium Facility to the Replacement Tritium Facility (RTF) began in 1993. To ensure that operation of new DOE facilities such as RTF present minimum involuntary and voluntary risks to the neighboring public and workers, indices of risk have been established to serve as target levels or safety goals of performance for assessing nuclear safety. These goals are discussed from a historical perspective in the initial part of thi