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

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.

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)

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

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

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)

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)

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)

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

International Nuclear Information System (INIS)

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

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

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)

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

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

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

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)

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

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

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.)

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)

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

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.

Probabilistic Structural Analysis Program

Science.gov (United States)

Pai, Shantaram S.; Chamis, Christos C.; Murthy, Pappu L. N.; Stefko, George L.; Riha, David S.; Thacker, Ben H.; Nagpal, Vinod K.; Mital, Subodh K.

2010-01-01

NASA/NESSUS 6.2c is a general-purpose, probabilistic analysis program that computes probability of failure and probabilistic sensitivity measures of engineered systems. Because NASA/NESSUS uses highly computationally efficient and accurate analysis techniques, probabilistic solutions can be obtained even for extremely large and complex models. Once the probabilistic response is quantified, the results can be used to support risk-informed decisions regarding reliability for safety-critical and one-of-a-kind systems, as well as for maintaining a level of quality while reducing manufacturing costs for larger-quantity products. NASA/NESSUS has been successfully applied to a diverse range of problems in aerospace, gas turbine engines, biomechanics, pipelines, defense, weaponry, and infrastructure. This program combines state-of-the-art probabilistic algorithms with general-purpose structural analysis and lifting methods to compute the probabilistic response and reliability of engineered structures. Uncertainties in load, material properties, geometry, boundary conditions, and initial conditions can be simulated. The structural analysis methods include non-linear finite-element methods, heat-transfer analysis, polymer/ceramic matrix composite analysis, monolithic (conventional metallic) materials life-prediction methodologies, boundary element methods, and user-written subroutines. Several probabilistic algorithms are available such as the advanced mean value method and the adaptive importance sampling method. NASA/NESSUS 6.2c is structured in a modular format with 15 elements.

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)

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

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.

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

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

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.

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)

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

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)

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)

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

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)

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)

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

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)

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

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

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

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)

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)

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

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

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)

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)

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)

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

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

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

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

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

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)

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.

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

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)

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)

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

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)

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.

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)

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.

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)

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

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.

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)

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....

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

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

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

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.)

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)

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)

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

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)

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

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

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

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.

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

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...

Probabilistic Durability Analysis in Advanced Engineering Design

Directory of Open Access Journals (Sweden)

A. Kudzys

2000-01-01

Full Text Available Expedience of probabilistic durability concepts and approaches in advanced engineering design of building materials, structural members and systems is considered. Target margin values of structural safety and serviceability indices are analyzed and their draft values are presented. Analytical methods of the cumulative coefficient of correlation and the limit transient action effect for calculation of reliability indices are given. Analysis can be used for probabilistic durability assessment of carrying and enclosure metal, reinforced concrete, wood, plastic, masonry both homogeneous and sandwich or composite structures and some kinds of equipments. Analysis models can be applied in other engineering fields.

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

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

International Nuclear Information System (INIS)

2002-01-01

publication is not necessary for all types of facility or PSA applications. In fact, it is anticipated that for many facilities, a 'streamlined' or 'simplified' interpretation of the information presented in this TECDOC will be acceptable. The appropriate level and form of streamlining is dependent upon the specific objectives of the analysis and the magnitude of the hazard that the facility represents. Facility hazard can drive the depth of analysis, as it may well be appropriate to analyse a lower hazard facility to less depth than higher hazard facilities (i.e., the depth of analysis is commensurate with the risk). Thus, the concept of hazard-graded depth of probabilistic safety analysis is considered as appropriate for non-reactor nuclear facilities

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

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)

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

2000-01-01

The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

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)

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

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

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)

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.

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

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

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

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)

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

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

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.)

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)

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

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

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)

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

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

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

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

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

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

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

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

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)

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)

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)

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

Probabilistic risk analysis of Angra-1 reactor

International Nuclear Information System (INIS)

Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

1986-01-01

The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt

Probabilistic Structural Analysis Methods (PSAM) for select space propulsion system structural components

Science.gov (United States)

Cruse, T. A.

1987-01-01

The objective is the development of several modular structural analysis packages capable of predicting the probabilistic response distribution for key structural variables such as maximum stress, natural frequencies, transient response, etc. The structural analysis packages are to include stochastic modeling of loads, material properties, geometry (tolerances), and boundary conditions. The solution is to be in terms of the cumulative probability of exceedance distribution (CDF) and confidence bounds. Two methods of probability modeling are to be included as well as three types of structural models - probabilistic finite-element method (PFEM); probabilistic approximate analysis methods (PAAM); and probabilistic boundary element methods (PBEM). The purpose in doing probabilistic structural analysis is to provide the designer with a more realistic ability to assess the importance of uncertainty in the response of a high performance structure. Probabilistic Structural Analysis Method (PSAM) tools will estimate structural safety and reliability, while providing the engineer with information on the confidence that should be given to the predicted behavior. Perhaps most critically, the PSAM results will directly provide information on the sensitivity of the design response to those variables which are seen to be uncertain.

Probabilistic Structural Analysis Methods for select space propulsion system structural components (PSAM)

Science.gov (United States)

Cruse, T. A.; Burnside, O. H.; Wu, Y.-T.; Polch, E. Z.; Dias, J. B.

1988-01-01

The objective is the development of several modular structural analysis packages capable of predicting the probabilistic response distribution for key structural variables such as maximum stress, natural frequencies, transient response, etc. The structural analysis packages are to include stochastic modeling of loads, material properties, geometry (tolerances), and boundary conditions. The solution is to be in terms of the cumulative probability of exceedance distribution (CDF) and confidence bounds. Two methods of probability modeling are to be included as well as three types of structural models - probabilistic finite-element method (PFEM); probabilistic approximate analysis methods (PAAM); and probabilistic boundary element methods (PBEM). The purpose in doing probabilistic structural analysis is to provide the designer with a more realistic ability to assess the importance of uncertainty in the response of a high performance structure. Probabilistic Structural Analysis Method (PSAM) tools will estimate structural safety and reliability, while providing the engineer with information on the confidence that should be given to the predicted behavior. Perhaps most critically, the PSAM results will directly provide information on the sensitivity of the design response to those variables which are seen to be uncertain.

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

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

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

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

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)

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)

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...

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

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)

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

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

International Nuclear Information System (INIS)

Purba, Julwan Hendry

2014-01-01

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

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)

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

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.

Bayesian Inference for NASA Probabilistic Risk and Reliability Analysis

Science.gov (United States)

Dezfuli, Homayoon; Kelly, Dana; Smith, Curtis; Vedros, Kurt; Galyean, William

2009-01-01

This document, Bayesian Inference for NASA Probabilistic Risk and Reliability Analysis, is intended to provide guidelines for the collection and evaluation of risk and reliability-related data. It is aimed at scientists and engineers familiar with risk and reliability methods and provides a hands-on approach to the investigation and application of a variety of risk and reliability data assessment methods, tools, and techniques. This document provides both: A broad perspective on data analysis collection and evaluation issues. A narrow focus on the methods to implement a comprehensive information repository. The topics addressed herein cover the fundamentals of how data and information are to be used in risk and reliability analysis models and their potential role in decision making. Understanding these topics is essential to attaining a risk informed decision making environment that is being sought by NASA requirements and procedures such as 8000.4 (Agency Risk Management Procedural Requirements), NPR 8705.05 (Probabilistic Risk Assessment Procedures for NASA Programs and Projects), and the System Safety requirements of NPR 8715.3 (NASA General Safety Program Requirements).

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)

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.

Probabilistic precursor analysis - an application of PSA

International Nuclear Information System (INIS)

Hari Prasad, M.; Gopika, V.; Sanyasi Rao, V.V.S.; Vaze, K.K.

2011-01-01

Incidents are inevitably part of the operational life of any complex industrial facility, and it is hard to predict how various contributing factors combine to cause the outcome. However, it should be possible to detect the existence of latent conditions that, together with the triggering failure(s), result in abnormal events. These incidents are called precursors. Precursor study, by definition, focuses on how a particular event might have adversely developed. This paper focuses on the events which can be analyzed to assess their potential to develop into core damage situation and looks into extending Probabilistic Safety Assessment techniques to precursor studies and explains the benefits through a typical case study. A preliminary probabilistic precursor analysis has been carried out for a typical NPP. The major advantages of this approach are the strong potential for augmenting event analysis which is currently carried out purely on deterministic basis. (author)

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.

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

Probabilistic Analysis of a Composite Crew Module

Science.gov (United States)

Mason, Brian H.; Krishnamurthy, Thiagarajan

2011-01-01

An approach for conducting reliability-based analysis (RBA) of a Composite Crew Module (CCM) is presented. The goal is to identify and quantify the benefits of probabilistic design methods for the CCM and future space vehicles. The coarse finite element model from a previous NASA Engineering and Safety Center (NESC) project is used as the baseline deterministic analysis model to evaluate the performance of the CCM using a strength-based failure index. The first step in the probabilistic analysis process is the determination of the uncertainty distributions for key parameters in the model. Analytical data from water landing simulations are used to develop an uncertainty distribution, but such data were unavailable for other load cases. The uncertainty distributions for the other load scale factors and the strength allowables are generated based on assumed coefficients of variation. Probability of first-ply failure is estimated using three methods: the first order reliability method (FORM), Monte Carlo simulation, and conditional sampling. Results for the three methods were consistent. The reliability is shown to be driven by first ply failure in one region of the CCM at the high altitude abort load set. The final predicted probability of failure is on the order of 10-11 due to the conservative nature of the factors of safety on the deterministic loads.

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

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.

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

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

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

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.

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)

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.

Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

International Nuclear Information System (INIS)

2014-01-01

The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

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

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

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)

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

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.

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

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

Probabilistic risk analysis in chemical engineering

International Nuclear Information System (INIS)

Schmalz, F.

1991-01-01

In risk analysis in the chemical industry, recognising potential risks is considered more important than assessing their quantitative extent. Even in assessing risks, emphasis is not on the probability involved but on the possible extent. Qualitative assessment has proved valuable here. Probabilistic methods are used in individual cases where the wide implications make it essential to be able to assess the reliability of safety precautions. In this case, assessment therefore centres on the reliability of technical systems and not on the extent of a chemical risk. 7 figs

Probabilistic Design and Analysis Framework

Science.gov (United States)

Strack, William C.; Nagpal, Vinod K.

2010-01-01

PRODAF is a software package designed to aid analysts and designers in conducting probabilistic analysis of components and systems. PRODAF can integrate multiple analysis programs to ease the tedious process of conducting a complex analysis process that requires the use of multiple software packages. The work uses a commercial finite element analysis (FEA) program with modules from NESSUS to conduct a probabilistic analysis of a hypothetical turbine blade, disk, and shaft model. PRODAF applies the response surface method, at the component level, and extrapolates the component-level responses to the system level. Hypothetical components of a gas turbine engine are first deterministically modeled using FEA. Variations in selected geometrical dimensions and loading conditions are analyzed to determine the effects of the stress state within each component. Geometric variations include the cord length and height for the blade, inner radius, outer radius, and thickness, which are varied for the disk. Probabilistic analysis is carried out using developing software packages like System Uncertainty Analysis (SUA) and PRODAF. PRODAF was used with a commercial deterministic FEA program in conjunction with modules from the probabilistic analysis program, NESTEM, to perturb loads and geometries to provide a reliability and sensitivity analysis. PRODAF simplified the handling of data among the various programs involved, and will work with many commercial and opensource deterministic programs, probabilistic programs, or modules.

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)

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

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

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

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

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

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)

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

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.

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)

Integrated framework for dynamic safety analysis

International Nuclear Information System (INIS)

Kim, Tae Wan; Karanki, Durga R.

2012-01-01

In the conventional PSA (Probabilistic Safety Assessment), detailed plant simulations by independent thermal hydraulic (TH) codes are used in the development of accident sequence models. Typical accidents in a NPP involve complex interactions among process, safety systems, and operator actions. As independent TH codes do not have the models of operator actions and full safety systems, they cannot literally simulate the integrated and dynamic interactions of process, safety systems, and operator responses. Offline simulation with pre decided states and time delays may not model the accident sequences properly. Moreover, when stochastic variability in responses of accident models is considered, defining all the combinations for simulations will be cumbersome task. To overcome some of these limitations of conventional safety analysis approach, TH models are coupled with the stochastic models in the dynamic event tree (DET) framework, which provides flexibility to model the integrated response due to better communication as all the accident elements are in the same model. The advantages of this framework also include: Realistic modeling in dynamic scenarios, comprehensive results, integrated approach (both deterministic and probabilistic models), and support for HRA (Human Reliability Analysis)

Method of extracting significant trouble information of nuclear power plants using probabilistic analysis technique

International Nuclear Information System (INIS)

Shimada, Yoshio; Miyazaki, Takamasa

2005-01-01

In order to analyze and evaluate large amounts of trouble information of overseas nuclear power plants, it is necessary to select information that is significant in terms of both safety and reliability. In this research, a method of efficiently and simply classifying degrees of importance of components in terms of safety and reliability while paying attention to root-cause components appearing in the information was developed. Regarding safety, the reactor core damage frequency (CDF), which is used in the probabilistic analysis of a reactor, was used. Regarding reliability, the automatic plant trip probability (APTP), which is used in the probabilistic analysis of automatic reactor trips, was used. These two aspects were reflected in the development of criteria for classifying degrees of importance of components. By applying these criteria, a simple method of extracting significant trouble information of overseas nuclear power plants was developed. (author)

Probabilistic Structural Analysis Theory Development

Science.gov (United States)

Burnside, O. H.

1985-01-01

The objective of the Probabilistic Structural Analysis Methods (PSAM) project is to develop analysis techniques and computer programs for predicting the probabilistic response of critical structural components for current and future space propulsion systems. This technology will play a central role in establishing system performance and durability. The first year's technical activity is concentrating on probabilistic finite element formulation strategy and code development. Work is also in progress to survey critical materials and space shuttle mian engine components. The probabilistic finite element computer program NESSUS (Numerical Evaluation of Stochastic Structures Under Stress) is being developed. The final probabilistic code will have, in the general case, the capability of performing nonlinear dynamic of stochastic structures. It is the goal of the approximate methods effort to increase problem solving efficiency relative to finite element methods by using energy methods to generate trial solutions which satisfy the structural boundary conditions. These approximate methods will be less computer intensive relative to the finite element approach.

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)

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

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)

Safety analysis in support of regulatory decision marking

International Nuclear Information System (INIS)

Pomier Baez, L.; Troncoso Fleitas, M.; Valhuerdi Debesa, C.; Valle Cepero, R.; Hernandez, J.L.

1996-01-01

Features of different safety analysis techniques by means of calculation thermohydraulic a probabilistic and severe accidents used in the safety assessment, as well as the development of these techniques in Cuba and their use in support of regulatory decision making are presented

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.

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.

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

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

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)

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

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

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.)

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

Advanced methods for a probabilistic safety analysis of fires. Development of advanced methods for performing as far as possible realistic plant specific fire risk analysis (fire PSA)

International Nuclear Information System (INIS)

Hofer, E.; Roewekamp, M.; Tuerschmann, M.

2003-07-01

In the frame of the research project RS 1112 'Development of Methods for a Recent Probabilistic Safety Analysis, Particularly Level 2' funded by the German Federal Ministry of Economics and Technology (BMWi), advanced methods, in particular for performing as far as possible realistic plant specific fire risk analyses (fire PSA), should be developed. The present Technical Report gives an overview on the methodologies developed in this context for assessing the fire hazard. In the context of developing advanced methodologies for fire PSA, a probabilistic dynamics analysis with a fire simulation code including an uncertainty and sensitivity study has been performed for an exemplary scenario of a cable fire induced by an electric cabinet inside the containment of a modern Konvoi type German nuclear power plant taking into consideration the effects of fire detection and fire extinguishing means. With the present study, it was possible for the first time to determine the probabilities of specified fire effects from a class of fire events by means of probabilistic dynamics supplemented by uncertainty and sensitivity analyses. The analysis applies a deterministic dynamics model, consisting of a dynamic fire simulation code and a model of countermeasures, considering effects of the stochastics (so-called aleatory uncertainties) as well as uncertainties in the state of knowledge (so-called epistemic uncertainties). By this means, probability assessments including uncertainties are provided to be used within the PSA. (orig.) [de

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)

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

Use of probabilistic studies in the analysis of modifications of French nuclear power plants

International Nuclear Information System (INIS)

Gros, G.; Milhem, J.M.

1985-11-01

The 900 MWe water pressurized reactors have been designed on deterministic basis. It appeared that some safety systems had a probability of failure non negligible and that their total failure could involve, in a short-term, severe consequences. This situation led Electricite de France to propose complementary measures (control-procedures, and associated modifications). To judge the efficiency of such measures, the safety authorities thought it was advisable to rest on probabilistic studies which have been developed by the Department of Safety Analysis of the C.E.A. The contribution of such studies, in the choice of modifications by information on the weak points and in the judgement on the efficiency of these modifications by probabilistic estimation of meltdown, is illustrated with the example of electric power supplies [fr

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)

Probabilistic Modeling of the Fatigue Crack Growth Rate for Ni-base Alloy X-750

International Nuclear Information System (INIS)

Yoon, J.Y.; Nam, H.O.; Hwang, I.S.; Lee, T.H.

2012-01-01

Extending the operating life of existing nuclear power plants (NPP's) beyond 60 years. Many aging problems of passive components such as PWSCC, IASCC, FAC and Corrosion Fatigue; Safety analysis: Deterministic analysis + Probabilistic analysis; Many uncertainties of parameters or relationship in general probabilistic analysis such as probabilistic safety assessment (PSA); Bayesian inference: Decreasing uncertainties by updating unknown parameter; Ensuring the reliability of passive components (e.g. pipes) as well as active components (e.g. valve, pump) in NPP's; Developing probabilistic model for failures; Updating the fatigue crack growth rate (FCGR)

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)

Probabilistic risk analysis and its role in regulatory activity in a developing country

International Nuclear Information System (INIS)

Arredondo-Sanchez, C.

1985-01-01

The author discusses the criterion adopted for regulatory activity in a developing country with a nuclear power plant. He describes the problems that have to be overcome as a result of changes in the regulations during construction of the plant. There is discussion of the action taken by the regulatory body when introducing the method of probabilistic risk analysis. The part played by this form of analysis in quantifying the safety objectives proposed in the USA together with its limitations and the problems involved in this methodology are examined. Lastly, the author gives an opinion on the use that probabilistic risk analysis should be put to in developing countries such as Mexico. (author)

Rad waste disposal safety analysis / Integrated safety assessment of a waste repository

International Nuclear Information System (INIS)

Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

2012-04-01

We developed CYPRUS+and adopted PID and RES method for the development of scenario. Safety performance assessment program was developed using GoldSim for the safety assessment of disposal system for the disposal of spnet fuels and wastes resulting from the pyrpoprocessing. Biosphere model was developed and verified in cooperation with JAEA. The capability to evaluate post-closure performance and safety was added to the previously developed program. And, nuclide migration and release to the biosphere considering site characteristics was evaluated by using deterministic and probabilistic approach. Operational safety assessment for drop, fire, and earthquake was also statistically evaluated considering well-established input parameter distribution. Conservative assessment showed that dose rate is below the limit value of low- and intermediate-level repository. Gas generation mechanism within engineered barrier was defined and its influence on safety was evaluated. We made probabilistic safety assessment by obtaining the probability distribution functions of important input variables and also made a sensitivity analysis. The maximum annual dose rate was shown to be below the safety limit value of 10 mSv/yr. The structure and element of safety case was developed to increase reliability of safety assessment methodology for a deep geological repository. Finally, milestone for safety case development and implementation strategy for each safety case element was also proposed

Support analysis for safety analysis development for CANDU nuclear power plant

International Nuclear Information System (INIS)

Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

2004-01-01

Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

Why do probabilistic finite element analysis ?

CERN Document Server

Thacker, Ben H

2008-01-01

The intention of this book is to provide an introduction to performing probabilistic finite element analysis. As a short guideline, the objective is to inform the reader of the use, benefits and issues associated with performing probabilistic finite element analysis without excessive theory or mathematical detail.

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.

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

Probabilistic methods used in NUSS

International Nuclear Information System (INIS)

Fischer, J.; Giuliani, P.

1985-01-01

Probabilistic considerations are used implicitly or explicitly in all technical areas. In the NUSS codes and guides the two areas of design and siting are those where more use is made of these concepts. A brief review of the relevant documents in these two areas is made in this paper. It covers the documents where either probabilistic considerations are implied or where probabilistic approaches are recommended in the evaluation of situations and of events. In the siting guides the review mainly covers the area of seismic hydrological and external man-made events analysis, as well as some aspects of meteorological extreme events analysis. Probabilistic methods are recommended in the design guides but they are not made a requirement. There are several reasons for this, mainly lack of reliable data and the absence of quantitative safety limits or goals against which to judge the design analysis. As far as practical, engineering judgement should be backed up by quantitative probabilistic analysis. Examples are given and the concept of design basis as used in NUSS design guides is explained. (author)

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

Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities (4). Investigation of safety evaluation method for fire and explosion incidents

International Nuclear Information System (INIS)

Abe, Hitoshi; Tashiro, Shinsuke; Ueda, Yoshinori

2010-01-01

A special committee on 'Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)' was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the useful information related to the establishment of quantitative performance objectives and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the analysis method of consequences for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution, and fire (including rapid decomposition of TBP complexes), resulting in the release of radio active materials into the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this technical report, the research results about basic experimental data and the method for safety evaluation of fire and explosion incidents were summarized. (author)

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.

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)

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.)

Application of the probabilistic method at the E.D.F

International Nuclear Information System (INIS)

Gachot, Bernard

1976-01-01

Having first evoked the problems arising from the definition of a so-called 'acceptable risk', the probabilistic study programme on safety carried out at the E.D.F. is described. The different aspects of the probabilistic estimation of a hazard are presented as well as the different steps i.e. collecting the information, carrying out a quantitative and qualitative analysis, which characterize the probabilistic study of safety problems. The problem of data determination is considered on reliability of the equipment, noting as a conclusion, that in spite of the lack of accuracy of the present data, the probabilistic methods already appear as a highly valuable tool favouring an homogenous and coherent approach of nuclear plant safety [fr

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.

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.

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)

Probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Shinaishin, M.A.

1988-06-01

The objective of this work is to provide the tools necessary for clear identification of: the purpose of a Probabilistic Risk Study, the bounds and depth of the study, the proper modeling techniques to be used, the failure modes contributing to the analysis, the classical and baysian approaches for manipulating data necessary for quantification, ways for treating uncertainties, and available computer codes that may be used in performing such probabilistic analysis. In addition, it provides the means for measuring the importance of a safety feature to maintaining a level of risk at a Nuclear Power Plant and the worth of optimizing a safety system in risk reduction. In applying these techniques so that they accommodate our national resources and needs it was felt that emphasis should be put on the system reliability analysis level of PRA. Objectives of such studies could include: comparing systems' designs of the various vendors in the bedding stage, and performing grid reliability and human performance analysis using national specific data. (author)

Probabilistic risk assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Shinaishin, M A

1988-06-15

The objective of this work is to provide the tools necessary for clear identification of: the purpose of a Probabilistic Risk Study, the bounds and depth of the study, the proper modeling techniques to be used, the failure modes contributing to the analysis, the classical and baysian approaches for manipulating data necessary for quantification, ways for treating uncertainties, and available computer codes that may be used in performing such probabilistic analysis. In addition, it provides the means for measuring the importance of a safety feature to maintaining a level of risk at a Nuclear Power Plant and the worth of optimizing a safety system in risk reduction. In applying these techniques so that they accommodate our national resources and needs it was felt that emphasis should be put on the system reliability analysis level of PRA. Objectives of such studies could include: comparing systems' designs of the various vendors in the bedding stage, and performing grid reliability and human performance analysis using national specific data. (author)

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

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

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)

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)

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)

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)

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.

Failure analysis of the cement mantle in total hip arthroplasty with an efficient probabilistic method.

Science.gov (United States)

Kaymaz, Irfan; Bayrak, Ozgu; Karsan, Orhan; Celik, Ayhan; Alsaran, Akgun

2014-04-01

Accurate prediction of long-term behaviour of cemented hip implants is very important not only for patient comfort but also for elimination of any revision operation due to failure of implants. Therefore, a more realistic computer model was generated and then used for both deterministic and probabilistic analyses of the hip implant in this study. The deterministic failure analysis was carried out for the most common failure states of the cement mantle. On the other hand, most of the design parameters of the cemented hip are inherently uncertain quantities. Therefore, the probabilistic failure analysis was also carried out considering the fatigue failure of the cement mantle since it is the most critical failure state. However, the probabilistic analysis generally requires large amount of time; thus, a response surface method proposed in this study was used to reduce the computation time for the analysis of the cemented hip implant. The results demonstrate that using an efficient probabilistic approach can significantly reduce the computation time for the failure probability of the cement from several hours to minutes. The results also show that even the deterministic failure analyses do not indicate any failure of the cement mantle with high safety factors, the probabilistic analysis predicts the failure probability of the cement mantle as 8%, which must be considered during the evaluation of the success of the cemented hip implants.

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

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.

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

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

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)

A review of the South Texas Project probabilistic safety analysis for accident frequency estimates and containment binning

International Nuclear Information System (INIS)

Wheeler, T.A.; Lambright, J.A.; Sype, T.T.; Darby, J.L.; Walsh, B.

1991-08-01

The objective of this review is to evaluate the South Texas Project (STP) Probabilistic Safety Analysis (PSA) for the USNRC. The PSA was reviewed for thoroughness of analysis, accuracy in plant modeling, legitimacy of assumptions, and overall quality of the work. The review is limited to the internal event analysis and the fire sequence analysis. This review is not a quantitative evaluation of the adequacy of the PSA. The adequacy of the PSA depends on the intended uses and must be addressed on a case-by-case basis by the licensee and the NRC. This review identifies strengths, weakness, and areas where additional clarification would assist the NRC in evaluating the PSA for specific regulatory purposes. The licensee, Houston Lighting and Power (HL ampersand P), reviewed a draft version of this report prior to its final release to the USNRC. The responses provided by HL ampersand P are provided in detail in appendices to this report, and they are summarized in the main body of the report. All issues raised during the review were adequately addressed by HL ampersand P in the responses. 27 refs., 4 tabs

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

Sensitivity analysis in multi-parameter probabilistic systems

International Nuclear Information System (INIS)

Walker, J.R.

1987-01-01

Probabilistic methods involving the use of multi-parameter Monte Carlo analysis can be applied to a wide range of engineering systems. The output from the Monte Carlo analysis is a probabilistic estimate of the system consequence, which can vary spatially and temporally. Sensitivity analysis aims to examine how the output consequence is influenced by the input parameter values. Sensitivity analysis provides the necessary information so that the engineering properties of the system can be optimized. This report details a package of sensitivity analysis techniques that together form an integrated methodology for the sensitivity analysis of probabilistic systems. The techniques have known confidence limits and can be applied to a wide range of engineering problems. The sensitivity analysis methodology is illustrated by performing the sensitivity analysis of the MCROC rock microcracking model

Probabilistic structural analysis of aerospace components using NESSUS

Science.gov (United States)

Shiao, Michael C.; Nagpal, Vinod K.; Chamis, Christos C.

1988-01-01

Probabilistic structural analysis of a Space Shuttle main engine turbopump blade is conducted using the computer code NESSUS (numerical evaluation of stochastic structures under stress). The goal of the analysis is to derive probabilistic characteristics of blade response given probabilistic descriptions of uncertainties in blade geometry, material properties, and temperature and pressure distributions. Probability densities are derived for critical blade responses. Risk assessment and failure life analysis is conducted assuming different failure models.

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.

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.

Analysis of area events as part of probabilistic safety assessment for Romanian TRIGA SSR 14 MW reactor

International Nuclear Information System (INIS)

Mladin, D.; Stefan, I.

2005-01-01

The international experience has shown that the external events could be an important contributor to plant/ reactor risk. For this reason such events have to be included in the PSA studies. In the context of PSA for nuclear facilities, external events are defined as events originating from outside the plant, but with the potential to create an initiating event at the plant. To support plant safety assessment, PSA can be used to find methods for identification of vulnerable features of the plant and to suggest modifications in order to mitigate the impact of external events or the producing of initiating events. For that purpose, probabilistic assessment of area events concerning fire and flooding risk and impact is necessary. Due to the relatively large power level amongst research reactors, the approach to safety analysis of Romanian 14 MW TRIGA benefits from an ongoing PSA project. In this context, treatment of external events should be considered. The specific tasks proposed for the complete evaluation of area event analysis are: identify the rooms important for facility safety, determine a relative area event risk index for these rooms and a relative area event impact index if the event occurs, evaluate the rooms specific area event frequency, determine the rooms contribution to reactor hazard state frequencies, analyze power supply and room dependencies of safety components (as pumps, motor operated valves). The fire risk analysis methodology is based on Berry's method [1]. This approach provides a systematic procedure to carry out a relative index of different rooms. The factors, which affect the fire probability, are: personal presence in the room, number and type of ignition sources, type and area of combustibles, fuel available in the room, fuel location, and ventilation. The flooding risk analysis is based on the amount of piping in the room. For accuracy of the information regarding piping a facility walk-about is necessary. In case of flooding risk

Probabilistic Fatigue Analysis of Jacket Support Structures for Offshore Wind Turbines Exemplified on Tubular Joints

OpenAIRE

Kelma, Sebastian; Schaumann, Peter

2015-01-01

The design of offshore wind turbines is usually based on the semi-probabilistic safety concept. Using probabilistic methods, the aim is to find an advanced structural design of OWTs in order to improve safety and reduce costs. The probabilistic design is exemplified on tubular joints of a jacket substructure. Loads and resistance are considered by their respective probability distributions. Time series of loads are generated by fully-coupled numerical simulation of the offshore wind turbine. ...

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)

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)

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

Ignalina Safety Analysis Group

International Nuclear Information System (INIS)

Ushpuras, E.

1995-01-01

The article describes the fields of activities of Ignalina NPP Safety Analysis Group (ISAG) in the Lithuanian Energy Institute and overview the main achievements gained since the group establishment in 1992. The group is working under the following guidelines: in-depth analysis of the fundamental physical processes of RBMK-1500 reactors; collection, systematization and verification of the design and operational data; simulation and analysis of potential accident consequences; analysis of thermohydraulic and neutronic characteristics of the plant; provision of technical and scientific consultations to VATESI, Governmental authorities, and also international institutions, participating in various projects aiming at Ignalina NPP safety enhancement. The ISAG is performing broad scientific co-operation programs with both Eastern and Western scientific groups, supplying engineering assistance for Ignalina NPP. ISAG is also participating in the joint Lithuanian - Swedish - Russian project - Barselina, the first Probabilistic Safety Assessment (PSA) study of Ignalina NPP. The work is underway together with Maryland University (USA) for assessment of the accident confinement system for a range of breaks in the primary circuit. At present the ISAG personnel is also involved in the project under the grant from the Nuclear Safety Account, administered by the European Bank for reconstruction and development for the preparation and review of an in-depth safety assessment of the Ignalina plant

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

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)

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)

Presurized water reactor safety approach and analysis. From conception to experience feedback

International Nuclear Information System (INIS)

Libmann, J.

1987-04-01

This report deals in ten chapters, with the following subjects: 1. Safety approach methods; 2. Study of accidents; 3. Safety analysis; 4. Study of internal aggressions or those involved by the site; 5. Consideration of complementary situations; 6. Three Mile Island accident; 7. Safety during operation and experience feedback; 8. An example of analysis: steam generator closure plug; 9. Probabilistic safety evaluation; 10. Chernobyl accident. 30 refs [fr

The PSA of safety-critical digital I and C system: the determination of important factors and sensitivity analysis

International Nuclear Information System (INIS)

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

2002-01-01

This report is prepared to suggest a practical Probabilistic Safety Assessment (PSA) methodology of safety-critical digital instrumentation and control (I and C) systems. Even though conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it because the result of probabilistic safety assessment plays very important role in proving the safety of a designed system. Microprocessors and software technologies make the digital system very complex and hard to analyze the safety of their applications. The aim of this is: (1) To summarize the factors which should be represented by the model for probabilistic safety assessment and to propose a standpoint of evaluation for digital systems. (2) To quantitatively presents the results of a mathematical case study which examines the analysis framework of the safety of digital systems in the context of the PSA. (3) To show the results of a sensitivity study for some critical factors

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

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)

Probabilistic structural analysis methods for select space propulsion system components

Science.gov (United States)

Millwater, H. R.; Cruse, T. A.

1989-01-01

The Probabilistic Structural Analysis Methods (PSAM) project developed at the Southwest Research Institute integrates state-of-the-art structural analysis techniques with probability theory for the design and analysis of complex large-scale engineering structures. An advanced efficient software system (NESSUS) capable of performing complex probabilistic analysis has been developed. NESSUS contains a number of software components to perform probabilistic analysis of structures. These components include: an expert system, a probabilistic finite element code, a probabilistic boundary element code and a fast probability integrator. The NESSUS software system is shown. An expert system is included to capture and utilize PSAM knowledge and experience. NESSUS/EXPERT is an interactive menu-driven expert system that provides information to assist in the use of the probabilistic finite element code NESSUS/FEM and the fast probability integrator (FPI). The expert system menu structure is summarized. The NESSUS system contains a state-of-the-art nonlinear probabilistic finite element code, NESSUS/FEM, to determine the structural response and sensitivities. A broad range of analysis capabilities and an extensive element library is present.

Status of Ignalina's safety analysis reports

International Nuclear Information System (INIS)

Uspuras, E.

1999-01-01

Ignalina NPP is unique among RBMK type reactors in the scope and comprehensiveness of international studies which have been performed to verify its design parameters and analyze risk levels. International assistance took several forms, a very valuable mod of assistance utilized the knowledge of international experts in extensive international studies whose purpose was: collection, systematization and verification of plant design data; analysis of risk levels; recommendations leading to improvements in the safety lave; transfer of state of the art analytical methodology to Lithuanian specialists. The major large scale international studies include: probabilistic risk analysis; extensive international study meant to provide comprehensive overview of plant status with special emphasis on safety aspects; an extensive review of the Safety Analysis Report by an independent group of international experts. In spite of the safety improvements and analyses which have been performed at the Ignalina NPP, much remains to be done in the nearest future

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

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)

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.

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.

Optimization (Alara) and probabilistic exposures: the application of optimization criteria to the control of risks due to exposures of a probabilistic nature

International Nuclear Information System (INIS)

Gonzalez, A.J.

1989-01-01

The paper described the application of the principles of optimization recommended by the International Commission on Radiological Protection (ICRP) to the restrain of radiation risks due to exposures that may or may not be incurred and to which a probability of occurrence can be assigned. After describing the concept of probabilistic exposures, it proposes a basis for a converging policy of control for both certain and probabilistic exposures, namely the dose-risk relationship adopted for radiation protection purposes. On that basis some coherent approaches for dealing with probabilistic exposures, such as the limitation of individual risks, are discussed. The optimization of safety for reducing all risks from probabilistic exposures to as-low-as-reasonably-achievable (ALARA) levels is reviewed in full. The principles of optimization of protection are used as a basic framework and the relevant factors to be taken into account when moving to probabilistic exposures are presented. The paper also reviews the decision-aiding techniques suitable for performing optimization with particular emphasis to the multi-attribute utility-analysis technique. Finally, there is a discussion on some practical application of decision-aiding multi-attribute utility analysis to probabilistic exposures including the use of probabilistic utilities. In its final outlook, the paper emphasizes the need for standardization and solutions to generic problems, if optimization of safety is to be successful

Probabilistic evaluations for CANTUP computer code analysis improvement

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2004-01-01

Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

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)

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

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

International Nuclear Information System (INIS)

Lin, Y.C.; Xie, Y.J.; Wang, X.H.

2004-01-01

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software

Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models

Energy Technology Data Exchange (ETDEWEB)

Cetiner, Mustafa Sacit; none,; Flanagan, George F. [ORNL; Poore III, Willis P. [ORNL; Muhlheim, Michael David [ORNL

2014-07-30

An integrated simulation platform that couples probabilistic analysis-based tools with model-based simulation tools can provide valuable insights for reactive and proactive responses to plant operating conditions. The objective of this work is to demonstrate the benefits of a partial implementation of the Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Framework Specification through the coupling of advanced PRA capabilities and accurate multi-physics plant models. Coupling a probabilistic model with a multi-physics model will aid in design, operations, and safety by providing a more accurate understanding of plant behavior. This represents the first attempt at actually integrating these two types of analyses for a control system used for operations, on a faster than real-time basis. This report documents the development of the basic communication capability to exchange data with the probabilistic model using Reliability Workbench (RWB) and the multi-physics model using Dymola. The communication pathways from injecting a fault (i.e., failing a component) to the probabilistic and multi-physics models were successfully completed. This first version was tested with prototypic models represented in both RWB and Modelica. First, a simple event tree/fault tree (ET/FT) model was created to develop the software code to implement the communication capabilities between the dynamic-link library (dll) and RWB. A program, written in C#, successfully communicates faults to the probabilistic model through the dll. A systems model of the Advanced Liquid-Metal Reactor–Power Reactor Inherently Safe Module (ALMR-PRISM) design developed under another DOE project was upgraded using Dymola to include proper interfaces to allow data exchange with the control application (ConApp). A program, written in C+, successfully communicates faults to the multi-physics model. The results of the example simulation were successfully plotted.

Global/local methods for probabilistic structural analysis

Science.gov (United States)

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

1993-04-01

A probabilistic global/local method is proposed to reduce the computational requirements of probabilistic structural analysis. A coarser global model is used for most of the computations with a local more refined model used only at key probabilistic conditions. The global model is used to establish the cumulative distribution function (cdf) and the Most Probable Point (MPP). The local model then uses the predicted MPP to adjust the cdf value. The global/local method is used within the advanced mean value probabilistic algorithm. The local model can be more refined with respect to the g1obal model in terms of finer mesh, smaller time step, tighter tolerances, etc. and can be used with linear or nonlinear models. The basis for this approach is described in terms of the correlation between the global and local models which can be estimated from the global and local MPPs. A numerical example is presented using the NESSUS probabilistic structural analysis program with the finite element method used for the structural modeling. The results clearly indicate a significant computer savings with minimal loss in accuracy.

A Probabilistic Approach for Robustness Evaluation of Timber Structures

DEFF Research Database (Denmark)

Kirkegaard, Poul Henning; Sørensen, John Dalsgaard

of Structures and a probabilistic modelling of the timber material proposed in the Probabilistic Model Code (PMC) of the Joint Committee on Structural Safety (JCSS). Due to the framework in the Danish Code the timber structure has to be evaluated with respect to the following criteria where at least one shall...... to criteria a) and b) the timber frame structure has one column with a reliability index a bit lower than an assumed target level. By removal three columns one by one no significant extensive failure of the entire structure or significant parts of it are obatined. Therefore the structure can be considered......A probabilistic based robustness analysis has been performed for a glulam frame structure supporting the roof over the main court in a Norwegian sports centre. The robustness analysis is based on the framework for robustness analysis introduced in the Danish Code of Practice for the Safety...

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

PRA (probabilistic risk analysis) in the nuclear sector. Quantifying human error and human malice

International Nuclear Information System (INIS)

Heyes, A.G.

1995-01-01

Regardless of the regulatory style chosen ('command and control' or 'functional') a vital prerequisite for coherent safety regulations in the nuclear power industry is the ability to assess accident risk. In this paper we present a critical analysis of current techniques of probabilistic risk analysis applied in the industry, with particular regard to the problems of quantifying risks arising from, or exacerbated by, human risk and/or human error. (Author)

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)

Extended probabilistic system assessment calculations within the SKI project-90

International Nuclear Information System (INIS)

Pereira, A.

1993-03-01

The probabilistic system assessment calculation reported in the SKI Project-90 final documents were restricted to the following nuclides: 14 C, 129 I, 135 Cs, 237 Np and 240 Pu. In this report we have extended those calculations to another five nuclides: 79 Se, 243 Am, 240 Pu, 93 Zr and 99 Tc. The execution of probabilistic assessment calculations integrated in the context of SKIs first safety analysis exercise of an hypothetic final repository for high-level nuclear waste in Sweden, was a learning experience of relevance for the conduction of probabilistic safety assessment in future exercises. Some major conclusions and viewpoints of future need related with probabilistic assessment were withdrawn from this work and are presented in our report

Annual activity report of Ignalina NPP Safety Analysis Group for 1996 year

International Nuclear Information System (INIS)

Ushpuras, E.; Augutis, J.; Bubelis, E.

1997-03-01

The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for 1996 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system. Ignalina Safety Analysis Report was prepared on the basis of these results. 37 refs., 9 tabs., 96 figs

P-CARES 2.0.0, Probabilistic Computer Analysis for Rapid Evaluation of Structures

International Nuclear Information System (INIS)

2008-01-01

1 - Description of program or function: P-CARES 2.0.0 (Probabilistic Computer Analysis for Rapid Evaluation of Structures) was developed for NRC staff use to determine the validity and accuracy of the analysis methods used by various utilities for structural safety evaluations of nuclear power plants. P-CARES provides the capability to effectively evaluate the probabilistic seismic response using simplified soil and structural models and to quickly check the validity and/or accuracy of the SSI data received from applicants and licensees. The code is organized in a modular format with the basic modules of the system performing static, seismic, and nonlinear analysis. 2 - Methods: P-CARES is an update of the CARES program developed at Brookhaven National Laboratory during the 1980's. A major improvement is the enhanced analysis capability in which a probabilistic algorithm has been implemented to perform the probabilistic site response and soil-structure interaction (SSI) analyses. This is accomplished using several sampling techniques such as the Latin Hypercube sampling (LHC), engineering LHC, the Fekete Point Set method, and also the traditional Monte Carlo simulation. This new feature enhances the site response and SSI analysis such that the effect of uncertainty in local site soil properties can now be quantified. Another major addition to P-CARES is a graphical user interface (GUI) which significantly improves the performance of P-Cares in terms of the inter-relations among different functions of the program, and facilitates the input/output processing and execution management. It also provides many user friendly features that would allow an analyst to quickly develop insights from the analysis results. 3 - Restrictions on the complexity of the problem: None noted

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

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.

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

Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Mi; Kim, Ji Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of); Seok, Ho [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

2016-10-15

An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics.

Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram

International Nuclear Information System (INIS)

Choi, Sun Mi; Kim, Ji Hwan; Seok, Ho

2016-01-01

An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics

Probabilistic design of fibre concrete structures

Science.gov (United States)

Pukl, R.; Novák, D.; Sajdlová, T.; Lehký, D.; Červenka, J.; Červenka, V.

2017-09-01

Advanced computer simulation is recently well-established methodology for evaluation of resistance of concrete engineering structures. The nonlinear finite element analysis enables to realistically predict structural damage, peak load, failure, post-peak response, development of cracks in concrete, yielding of reinforcement, concrete crushing or shear failure. The nonlinear material models can cover various types of concrete and reinforced concrete: ordinary concrete, plain or reinforced, without or with prestressing, fibre concrete, (ultra) high performance concrete, lightweight concrete, etc. Advanced material models taking into account fibre concrete properties such as shape of tensile softening branch, high toughness and ductility are described in the paper. Since the variability of the fibre concrete material properties is rather high, the probabilistic analysis seems to be the most appropriate format for structural design and evaluation of structural performance, reliability and safety. The presented combination of the nonlinear analysis with advanced probabilistic methods allows evaluation of structural safety characterized by failure probability or by reliability index respectively. Authors offer a methodology and computer tools for realistic safety assessment of concrete structures; the utilized approach is based on randomization of the nonlinear finite element analysis of the structural model. Uncertainty of the material properties or their randomness obtained from material tests are accounted in the random distribution. Furthermore, degradation of the reinforced concrete materials such as carbonation of concrete, corrosion of reinforcement, etc. can be accounted in order to analyze life-cycle structural performance and to enable prediction of the structural reliability and safety in time development. The results can serve as a rational basis for design of fibre concrete engineering structures based on advanced nonlinear computer analysis. The presented

Uncertainty analysis for Ulysses safety evaluation report

International Nuclear Information System (INIS)

Frank, M.V.

1991-01-01

As part of the effort to review the Ulysses Final Safety Analysis Report and to understand the risk of plutonium release from the Ulysses spacecraft General Purpose Heat Source---Radioisotope Thermal Generator (GPHS-RTG), the Interagency Nuclear Safety Review Panel (INSRP) and the author performed an integrated, quantitative analysis of the uncertainties of the calculated risk of plutonium release from Ulysses. Using state-of-art probabilistic risk assessment technology, the uncertainty analysis accounted for both variability and uncertainty of the key parameters of the risk analysis. The results show that INSRP had high confidence that risk of fatal cancers from potential plutonium release associated with calculated launch and deployment accident scenarios is low

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)

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

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)

Seismic fragility analysis of a nuclear building based on probabilistic seismic hazard assessment and soil-structure interaction analysis

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, R.; Ni, S.; Chen, R.; Han, X.M. [CANDU Energy Inc, Mississauga, Ontario (Canada); Mullin, D. [New Brunswick Power, Point Lepreau, New Brunswick (Canada)

2016-09-15

Seismic fragility analyses are conducted as part of seismic probabilistic safety assessment (SPSA) for nuclear facilities. Probabilistic seismic hazard assessment (PSHA) has been undertaken for a nuclear power plant in eastern Canada. Uniform Hazard Spectra (UHS), obtained from the PSHA, is characterized by high frequency content which differs from the original plant design basis earthquake spectral shape. Seismic fragility calculations for the service building of a CANDU 6 nuclear power plant suggests that the high frequency effects of the UHS can be mitigated through site response analysis with site specific geological conditions and state-of-the-art soil-structure interaction analysis. In this paper, it is shown that by performing a detailed seismic analysis using the latest technology, the conservatism embedded in the original seismic design can be quantified and the seismic capacity of the building in terms of High Confidence of Low Probability of Failure (HCLPF) can be improved. (author)

Recent developments of the NESSUS probabilistic structural analysis computer program

Science.gov (United States)

Millwater, H.; Wu, Y.-T.; Torng, T.; Thacker, B.; Riha, D.; Leung, C. P.

1992-01-01

The NESSUS probabilistic structural analysis computer program combines state-of-the-art probabilistic algorithms with general purpose structural analysis methods to compute the probabilistic response and the reliability of engineering structures. Uncertainty in loading, material properties, geometry, boundary conditions and initial conditions can be simulated. The structural analysis methods include nonlinear finite element and boundary element methods. Several probabilistic algorithms are available such as the advanced mean value method and the adaptive importance sampling method. The scope of the code has recently been expanded to include probabilistic life and fatigue prediction of structures in terms of component and system reliability and risk analysis of structures considering cost of failure. The code is currently being extended to structural reliability considering progressive crack propagation. Several examples are presented to demonstrate the new capabilities.

Probabilistic reasoning in data analysis.

Science.gov (United States)

Sirovich, Lawrence

2011-09-20

This Teaching Resource provides lecture notes, slides, and a student assignment for a lecture on probabilistic reasoning in the analysis of biological data. General probabilistic frameworks are introduced, and a number of standard probability distributions are described using simple intuitive ideas. Particular attention is focused on random arrivals that are independent of prior history (Markovian events), with an emphasis on waiting times, Poisson processes, and Poisson probability distributions. The use of these various probability distributions is applied to biomedical problems, including several classic experimental studies.

Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events

Directory of Open Access Journals (Sweden)

Robertas Alzbutas

2015-01-01

Full Text Available The analytical/deterministic modelling and simulation/probabilistic methods are used separately as a rule in order to analyse the physical processes and random or uncertain events. However, in the currently used probabilistic safety assessment this is an issue. The lack of treatment of dynamic interactions between the physical processes on one hand and random events on the other hand causes the limited assessment. In general, there are a lot of mathematical modelling theories, which can be used separately or integrated in order to extend possibilities of modelling and analysis. The Theory of Probabilistic Dynamics (TPD and its augmented version based on the concept of stimulus and delay are introduced for the dynamic reliability modelling and the simulation of accidents in hybrid (continuous-discrete systems considering uncertain events. An approach of non-Markovian simulation and uncertainty analysis is discussed in order to adapt the Stimulus-Driven TPD for practical applications. The developed approach and related methods are used as a basis for a test case simulation in view of various methods applications for severe accident scenario simulation and uncertainty analysis. For this and for wider analysis of accident sequences the initial test case specification is then extended and discussed. Finally, it is concluded that enhancing the modelling of stimulated dynamics with uncertainty and sensitivity analysis allows the detailed simulation of complex system characteristics and representation of their uncertainty. The developed approach of accident modelling and analysis can be efficiently used to estimate the reliability of hybrid systems and at the same time to analyze and possibly decrease the uncertainty of this estimate.

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

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...

HERMES probabilistic risk assessment. Pilot study

International Nuclear Information System (INIS)

Parisot, F.; Munoz, J.

1993-01-01

The study was performed in 1989 of the contribution of probabilistic analysis for the optimal construction of system safety status in aeronautical and European nuclear industries, shows the growing trends towards incorporation of quantitative safety assessment and lead to an agreement to undertake a prototype proof study on Hermes. The main steps of the study and results are presented in the paper

Deterministic and Probabilistic Analysis of NPP Communication Bridge Resistance Due to Extreme Loads

Directory of Open Access Journals (Sweden)

Králik Juraj

2014-12-01

Full Text Available This paper presents the experiences from the deterministic and probability analysis of the reliability of communication bridge structure resistance due to extreme loads - wind and earthquake. On the example of the steel bridge between two NPP buildings is considered the efficiency of the bracing systems. The advantages and disadvantages of the deterministic and probabilistic analysis of the structure resistance are discussed. The advantages of the utilization the LHS method to analyze the safety and reliability of the structures is presented

Human factors and fuzzy set theory for safety analysis

International Nuclear Information System (INIS)

Nishiwaki, Y.

1987-01-01

Human reliability and performance is affected by many factors: medical, physiological and psychological, etc. The uncertainty involved in human factors may not necessarily be probabilistic, but fuzzy. Therefore, it is important to develop a theory by which both the non-probabilistic uncertainties, or fuzziness, of human factors and the probabilistic properties of machines can be treated consistently. In reality, randomness and fuzziness are sometimes mixed. From the mathematical point of view, probabilistic measures may be considered a special case of fuzzy measures. Therefore, fuzzy set theory seems to be an effective tool for analysing man-machine systems. The concept 'failure possibility' based on fuzzy sets is suggested as an approach to safety analysis and fault diagnosis of a large complex system. Fuzzy measures and fuzzy integrals are introduced and their possible applications are also discussed. (author)

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

Deliverable D74.2. Probabilistic analysis methods for support structures

DEFF Research Database (Denmark)

Gintautas, Tomas

2018-01-01

Relevant Description: Report describing the probabilistic analysis for offshore substructures and results attained. This includes comparison with experimental data and with conventional design. Specific targets: 1) Estimate current reliability level of support structures 2) Development of basis...... for probabilistic calculations and evaluation of reliability for offshore support structures (substructures) 3) Development of a probabilistic model for stiffness and strength of soil parameters and for modeling geotechnical load bearing capacity 4) Comparison between probabilistic analysis and deterministic...

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)

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.)

Probabilistic Analysis Methods for Hybrid Ventilation

DEFF Research Database (Denmark)

Brohus, Henrik; Frier, Christian; Heiselberg, Per

This paper discusses a general approach for the application of probabilistic analysis methods in the design of ventilation systems. The aims and scope of probabilistic versus deterministic methods are addressed with special emphasis on hybrid ventilation systems. A preliminary application...... of stochastic differential equations is presented comprising a general heat balance for an arbitrary number of loads and zones in a building to determine the thermal behaviour under random conditions....

Probabilistic risk assessment: Number 219

International Nuclear Information System (INIS)

Bari, R.A.

1985-01-01

This report describes a methodology for analyzing the safety of nuclear power plants. A historical overview of plants in the US is provided, and past, present, and future nuclear safety and risk assessment are discussed. A primer on nuclear power plants is provided with a discussion of pressurized water reactors (PWR) and boiling water reactors (BWR) and their operation and containment. Probabilistic Risk Assessment (PRA), utilizing both event-tree and fault-tree analysis, is discussed as a tool in reactor safety, decision making, and communications. (FI)

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

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

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

Stockholm Safety Conference. Analysis of the sessions on radiological protection, licensing and risk assessment

International Nuclear Information System (INIS)

Gea, A.

1981-01-01

A summary of the sessions on radiological protection, licensing and risk assessment in the safety conference of Stockholm is presented. It is considered the new point of view of the nuclear safety, probabilistic analysis, components failures probability and accident analysis. They are included conclusions applicable in many cases to development countries. (author)

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)

Unsteady Probabilistic Analysis of a Gas Turbine System

Science.gov (United States)

Brown, Marilyn

2003-01-01

In this work, we have considered an annular cascade configuration subjected to unsteady inflow conditions. The unsteady response calculation has been implemented into the time marching CFD code, MSUTURBO. The computed steady state results for the pressure distribution demonstrated good agreement with experimental data. We have computed results for the amplitudes of the unsteady pressure over the blade surfaces. With the increase in gas turbine engine structural complexity and performance over the past 50 years, structural engineers have created an array of safety nets to ensure against component failures in turbine engines. In order to reduce what is now considered to be excessive conservatism and yet maintain the same adequate margins of safety, there is a pressing need to explore methods of incorporating probabilistic design procedures into engine development. Probabilistic methods combine and prioritize the statistical distributions of each design variable, generate an interactive distribution and offer the designer a quantified relationship between robustness, endurance and performance. The designer can therefore iterate between weight reduction, life increase, engine size reduction, speed increase etc.

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.

Scalable group level probabilistic sparse factor analysis

DEFF Research Database (Denmark)

Hinrich, Jesper Løve; Nielsen, Søren Føns Vind; Riis, Nicolai Andre Brogaard

2017-01-01

Many data-driven approaches exist to extract neural representations of functional magnetic resonance imaging (fMRI) data, but most of them lack a proper probabilistic formulation. We propose a scalable group level probabilistic sparse factor analysis (psFA) allowing spatially sparse maps, component...... pruning using automatic relevance determination (ARD) and subject specific heteroscedastic spatial noise modeling. For task-based and resting state fMRI, we show that the sparsity constraint gives rise to components similar to those obtained by group independent component analysis. The noise modeling...... shows that noise is reduced in areas typically associated with activation by the experimental design. The psFA model identifies sparse components and the probabilistic setting provides a natural way to handle parameter uncertainties. The variational Bayesian framework easily extends to more complex...

Development of a simple method for classifying the degree of importance of components in nuclear power plants using probabilistic analysis technique

International Nuclear Information System (INIS)

Shimada, Yoshio; Miyazaki, Takamasa

2006-01-01

In order to analyze large amounts of trouble information of overseas nuclear power plants, it is necessary to select information that is significant in terms of both safety and reliability. In this research, a method of efficiently and simply classifying degrees of importance of components in terms of safety and reliability while paying attention to root-cause components appearing in the information was developed. Regarding safety, the reactor core damage frequency (CDF), which is used in the probabilistic analysis of a reactor, was used. Regarding reliability, the automatic plant trip probability (APTP), which is used in the probabilistic analysis of automatic reactor trips, was used. These two aspects were reflected in the development of criteria for classifying degrees of importance of components. By applying these criteria, a method of quantitatively and simply judging the significance of trouble information of overseas nuclear power plants was developed. (author)

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)

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

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.)

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

Probabilistic Flood Defence Assessment Tools

Directory of Open Access Journals (Sweden)

Slomp Robert

2016-01-01

Full Text Available The WTI2017 project is responsible for the development of flood defence assessment tools for the 3600 km of Dutch primary flood defences, dikes/levees, dunes and hydraulic structures. These tools are necessary, as per January 1st 2017, the new flood risk management policy for the Netherlands will be implemented. Then, the seven decades old design practice (maximum water level methodology of 1958 and two decades old safety standards (and maximum hydraulic load methodology of 1996 will formally be replaced by a more risked based approach for the national policy in flood risk management. The formal flood defence assessment is an important part of this new policy, especially for flood defence managers, since national and regional funding for reinforcement is based on this assessment. This new flood defence policy is based on a maximum allowable probability of flooding. For this, a maximum acceptable individual risk was determined at 1/100 000 per year, this is the probability of life loss of for every protected area in the Netherlands. Safety standards of flood defences were then determined based on this acceptable individual risk. The results were adjusted based on information from cost -benefit analysis, societal risk and large scale societal disruption due to the failure of critical infrastructure e.g. power stations. The resulting riskbased flood defence safety standards range from a 300 to a 100 000 year return period for failure. Two policy studies, WV21 (Safety from floods in the 21st century and VNK-2 (the National Flood Risk in 2010 provided the essential information to determine the new risk based safety standards for flood defences. The WTI2017 project will provide the safety assessment tools based on these new standards and is thus an essential element for the implementation of this policy change. A major issue to be tackled was the development of user-friendly tools, as the new assessment is to be carried out by personnel of the

Probabilistic Tsunami Hazard Analysis

Science.gov (United States)

Thio, H. K.; Ichinose, G. A.; Somerville, P. G.; Polet, J.

2006-12-01

The recent tsunami disaster caused by the 2004 Sumatra-Andaman earthquake has focused our attention to the hazard posed by large earthquakes that occur under water, in particular subduction zone earthquakes, and the tsunamis that they generate. Even though these kinds of events are rare, the very large loss of life and material destruction caused by this earthquake warrant a significant effort towards the mitigation of the tsunami hazard. For ground motion hazard, Probabilistic Seismic Hazard Analysis (PSHA) has become a standard practice in the evaluation and mitigation of seismic hazard to populations in particular with respect to structures, infrastructure and lifelines. Its ability to condense the complexities and variability of seismic activity into a manageable set of parameters greatly facilitates the design of effective seismic resistant buildings but also the planning of infrastructure projects. Probabilistic Tsunami Hazard Analysis (PTHA) achieves the same goal for hazards posed by tsunami. There are great advantages of implementing such a method to evaluate the total risk (seismic and tsunami) to coastal communities. The method that we have developed is based on the traditional PSHA and therefore completely consistent with standard seismic practice. Because of the strong dependence of tsunami wave heights on bathymetry, we use a full waveform tsunami waveform computation in lieu of attenuation relations that are common in PSHA. By pre-computing and storing the tsunami waveforms at points along the coast generated for sets of subfaults that comprise larger earthquake faults, we can efficiently synthesize tsunami waveforms for any slip distribution on those faults by summing the individual subfault tsunami waveforms (weighted by their slip). This efficiency make it feasible to use Green's function summation in lieu of attenuation relations to provide very accurate estimates of tsunami height for probabilistic calculations, where one typically computes

Process for computing geometric perturbations for probabilistic analysis

Science.gov (United States)

Fitch, Simeon H. K. [Charlottesville, VA; Riha, David S [San Antonio, TX; Thacker, Ben H [San Antonio, TX

2012-04-10

A method for computing geometric perturbations for probabilistic analysis. The probabilistic analysis is based on finite element modeling, in which uncertainties in the modeled system are represented by changes in the nominal geometry of the model, referred to as "perturbations". These changes are accomplished using displacement vectors, which are computed for each node of a region of interest and are based on mean-value coordinate calculations.

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

Probabilistic analysis of extreme wind events

Energy Technology Data Exchange (ETDEWEB)

Chaviaropoulos, P.K. [Center for Renewable Energy Sources (CRES), Pikermi Attikis (Greece)

1997-12-31

A vital task in wind engineering and meterology is to understand, measure, analyse and forecast extreme wind conditions, due to their significant effects on human activities and installations like buildings, bridges or wind turbines. The latest version of the IEC standard (1996) pays particular attention to the extreme wind events that have to be taken into account when designing or certifying a wind generator. Actually, the extreme wind events within a 50 year period are those which determine the ``static`` design of most of the wind turbine components. The extremes which are important for the safety of wind generators are those associated with the so-called ``survival wind speed``, the extreme operating gusts and the extreme wind direction changes. A probabilistic approach for the analysis of these events is proposed in this paper. Emphasis is put on establishing the relation between extreme values and physically meaningful ``site calibration`` parameters, like probability distribution of the annual wind speed, turbulence intensity and power spectra properties. (Author)

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

Development and improvement of safety analysis code for geological disposal

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

A Probabilistic Analysis of the Sacco and Vanzetti Evidence

CERN Document Server

Kadane, Joseph B

2011-01-01

A Probabilistic Analysis of the Sacco and Vanzetti Evidence is a Bayesian analysis of the trial and post-trial evidence in the Sacco and Vanzetti case, based on subjectively determined probabilities and assumed relationships among evidential events. It applies the ideas of charting evidence and probabilistic assessment to this case, which is perhaps the ranking cause celebre in all of American legal history. Modern computation methods applied to inference networks are used to show how the inferential force of evidence in a complicated case can be graded. The authors employ probabilistic assess

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

Qualitative safety analysis in accelerator based systems

International Nuclear Information System (INIS)

Sarkar, P.K.; Chowdhury, Lekha M.

2006-01-01

In recent developments connected to high energy and high current accelerators, the accelerator driven systems (ADS) and the Radioactive Ion Beam (RIB) facilities come in the forefront of application. For medical and industrial applications high current accelerators often need to be located in populated areas. These facilities pose significant radiological hazard during their operation and accidental situations. We have done a qualitative evaluation of radiological safety analysis using the probabilistic safety analysis (PSA) methods for accelerator-based systems. The major contribution to hazard comes from a target rupture scenario in both ADS and RIB facilities. Other significant contributors to hazard in the facilities are also discussed using fault tree and event tree methodologies. (author)

GUI program to compute probabilistic seismic hazard analysis

International Nuclear Information System (INIS)

Shin, Jin Soo; Chi, H. C.; Cho, J. C.; Park, J. H.; Kim, K. G.; Im, I. S.

2006-12-01

The development of program to compute probabilistic seismic hazard is completed based on Graphic User Interface(GUI). The main program consists of three part - the data input processes, probabilistic seismic hazard analysis and result output processes. The probabilistic seismic hazard analysis needs various input data which represent attenuation formulae, seismic zoning map, and earthquake event catalog. The input procedure of previous programs based on text interface take a much time to prepare the data. The data cannot be checked directly on screen to prevent input erroneously in existing methods. The new program simplifies the input process and enable to check the data graphically in order to minimize the artificial error within limits of the possibility

Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

International Nuclear Information System (INIS)

Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z.

2015-01-01

The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

Energy Technology Data Exchange (ETDEWEB)

Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z., E-mail: jonatasfmata@yahoo.com.br, E-mail: vasconv@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

Annual activity report of Ignalina NPP Safety Analysis Group for 1995 year

International Nuclear Information System (INIS)

Ushpuras, E.; Augutis, J.; Bubelis, E.

1995-01-01

The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for 1995 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system. 18 refs., 9 tabs., 110 figs

Probabilistic assessment of faults

International Nuclear Information System (INIS)

Foden, R.W.

1987-01-01

Probabilistic safety analysis (PSA) is the process by which the probability (or frequency of occurrence) of reactor fault conditions which could lead to unacceptable consequences is assessed. The basic objective of a PSA is to allow a judgement to be made as to whether or not the principal probabilistic requirement is satisfied. It also gives insights into the reliability of the plant which can be used to identify possible improvements. This is explained in the article. The scope of a PSA and the PSA performed by the National Nuclear Corporation (NNC) for the Heysham II and Torness AGRs and Sizewell-B PWR are discussed. The NNC methods for hazards, common cause failure and operator error are mentioned. (UK)

Specifying design conservatism: Worst case versus probabilistic analysis

Science.gov (United States)

Miles, Ralph F., Jr.

1993-01-01

Design conservatism is the difference between specified and required performance, and is introduced when uncertainty is present. The classical approach of worst-case analysis for specifying design conservatism is presented, along with the modern approach of probabilistic analysis. The appropriate degree of design conservatism is a tradeoff between the required resources and the probability and consequences of a failure. A probabilistic analysis properly models this tradeoff, while a worst-case analysis reveals nothing about the probability of failure, and can significantly overstate the consequences of failure. Two aerospace examples will be presented that illustrate problems that can arise with a worst-case analysis.

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

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

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

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

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.

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.

Use of probabilistic safety analysis for design of emergency mitigation systems in hydrogen producer plant with sulfur-iodine technology, Section II: sulfuric acid decomposition

International Nuclear Information System (INIS)

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

2009-10-01

Over the last decades, the need to reduce emissions of greenhouse gases has prompted the development of technologies for the production of clean fuels through the use of primary energy resources of zero emissions, as the heat of nuclear reactors of high temperature. Within these technologies, one of the most promising is the hydrogen production by sulfur-iodine cycle coupled to a high temperature reactor initially proposed by General Atomics. By their nature and because it will be large-scale plants, the development of these technologies from its present phase to its procurement and construction, will have to incorporate emergency mitigation systems in all its parts and interconnections to prevent undesired events that could put threaten the plant integrity and the nearby area. For the particular case of sulfur-iodine thermochemical cycle, most analysis have focused on hydrogen explosions and failures in the primary cooling systems. While these events are the most catastrophic, is that there are also many other events that even taking less direct consequences, could jeopardize the plant operation, the people safety of nearby communities and carry the same economic consequences. In this study we analyzed one of these events, which is the formation of a toxic cloud prompted by uncontrolled leakage of concentrated sulfuric acid in the second section of sulfur-iodine process of General Atomics. In this section, the sulfuric acid concentration is near to 90% in conditions of high temperature and positive pressure. Under these conditions the sulfuric acid and sulfur oxides from the reactor will form a toxic cloud that the have contact with the plant personnel could cause fatalities, or to reach a town would cause suffocation, respiratory problems and eye irritation. The methodology used for this study is the supported design in probabilistic safety analysis. Mitigation systems were postulated based on the isolation of a possible leak, the neutralization of a pond of

Annual activity report of Ignalina NPP Safety Analysis Group for the year 1997

International Nuclear Information System (INIS)

Ushpuras, E.; Augutis, J.; Bubelis, E.; Kaliatka, A

1998-01-01

The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for the year 1997 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system

Probabilistic assessment of dry transport with burnup credit

International Nuclear Information System (INIS)

Lake, W.H.

2003-01-01

The general concept of probabilistic analysis and its application to the use of burnup credit in spent fuel transport is explored. Discussion of the probabilistic analysis method is presented. The concepts of risk and its perception are introduced, and models are suggested for performing probability and risk estimates. The general probabilistic models are used for evaluating the application of burnup credit for dry spent nuclear fuel transport. Two basic cases are considered. The first addresses the question of the relative likelihood of exceeding an established criticality safety limit with and without burnup credit. The second examines the effect of using burnup credit on the overall risk for dry spent fuel transport. Using reasoned arguments and related failure probability and consequence data analysis is performed to estimate the risks of using burnup credit for dry transport of spent nuclear fuel. (author)

Development of a probabilistic safety assessment framework for an interim dry storage facility subjected to an aircraft crash using best-estimate structural analysis

International Nuclear Information System (INIS)

Almomani, Belal; Jang, Dong Chan; Lee, Sang Hoon; Kang, Hyun Gook

2017-01-01

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research

Development of a Probabilistic Safety Assessment Framework for an Interim Dry Storage Facility Subjected to an Aircraft Crash Using Best-Estimate Structural Analysis

Directory of Open Access Journals (Sweden)

Belal Almomani

2017-03-01

Full Text Available Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

Development of a probabilistic safety assessment framework for an interim dry storage facility subjected to an aircraft crash using best-estimate structural analysis

Energy Technology Data Exchange (ETDEWEB)

Almomani, Belal; Jang, Dong Chan [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of); Kang, Hyun Gook [Dept. of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy (United States)

2017-03-15

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

2009-03-01

The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

Aluminum 7075-T6 fatigue data generation and probabilistic life prediction formulation

OpenAIRE

Kemna, John G.

1998-01-01

Approved for public release; distribution is unlimited. The life extension of aging fleet aircraft requires an assessment of the safe-life remaining after refurbishment. Risk can be estimated by conventional deterministic fatigue analysis coupled with a subjective factor of safety. Alternatively, risk can be quantitatively and objectively predicted by probabilistic analysis. In this investigation, a general probabilistic life formulation is specialized for constant amplitude, fully reverse...

Probabilistic approaches for geotechnical site characterization and slope stability analysis

CERN Document Server

Cao, Zijun; Li, Dianqing

2017-01-01

This is the first book to revisit geotechnical site characterization from a probabilistic point of view and provide rational tools to probabilistically characterize geotechnical properties and underground stratigraphy using limited information obtained from a specific site. This book not only provides new probabilistic approaches for geotechnical site characterization and slope stability analysis, but also tackles the difficulties in practical implementation of these approaches. In addition, this book also develops efficient Monte Carlo simulation approaches for slope stability analysis and implements these approaches in a commonly available spreadsheet environment. These approaches and the software package are readily available to geotechnical practitioners and alleviate them from reliability computational algorithms. The readers will find useful information for a non-specialist to determine project-specific statistics of geotechnical properties and to perform probabilistic analysis of slope stability.

Development of a procedure for qualitative and quantitative evaluation of human factors as a part of probabilistic safety assessments of nuclear power plants. Part A

International Nuclear Information System (INIS)

Richei, A.

1998-01-01

The objective of this project is the development of a procedure for the qualitative and quantitative evaluation of human factors in the probabilistic safety assessment for nuclear power plants. The Human Error Rate Assessment and Optimizing System (HEROS) is introduced. The evaluation of a task with HEROS is realized in the three evaluation levels, i.e. 'Management Structure', 'Working Environment' and 'Man-Machine-Interface'. The developed expert system uses the fuzzy set theory for an assessment. For the evaluation of cognitive tasks evaluation criteria are derived also. The validation of the procedure is based on three examples, reflecting the common practice of probabilistic safety assessments and including problems, which cannot, respectively - only insufficiently - be evaluated with the established human risk analysis procedures. HERO applications give plausible and comprehensible results. (orig.) [de

Systems analysis approach to probabilistic modeling of fault trees

International Nuclear Information System (INIS)

Bartholomew, R.J.; Qualls, C.R.

1985-01-01

A method of probabilistic modeling of fault tree logic combined with stochastic process theory (Markov modeling) has been developed. Systems are then quantitatively analyzed probabilistically in terms of their failure mechanisms including common cause/common mode effects and time dependent failure and/or repair rate effects that include synergistic and propagational mechanisms. The modeling procedure results in a state vector set of first order, linear, inhomogeneous, differential equations describing the time dependent probabilities of failure described by the fault tree. The solutions of this Failure Mode State Variable (FMSV) model are cumulative probability distribution functions of the system. A method of appropriate synthesis of subsystems to form larger systems is developed and applied to practical nuclear power safety systems

Probabilistic Output Analysis by Program Manipulation

DEFF Research Database (Denmark)

Rosendahl, Mads; Kirkeby, Maja Hanne

2015-01-01

The aim of a probabilistic output analysis is to derive a probability distribution of possible output values for a program from a probability distribution of its input. We present a method for performing static output analysis, based on program transformation techniques. It generates a probability...

Incorporating psychological influences in probabilistic cost analysis

Energy Technology Data Exchange (ETDEWEB)

Kujawski, Edouard; Alvaro, Mariana; Edwards, William

2004-01-08

Today's typical probabilistic cost analysis assumes an ''ideal'' project that is devoid of the human and organizational considerations that heavily influence the success and cost of real-world projects. In the real world ''Money Allocated Is Money Spent'' (MAIMS principle); cost underruns are rarely available to protect against cost overruns while task overruns are passed on to the total project cost. Realistic cost estimates therefore require a modified probabilistic cost analysis that simultaneously models the cost management strategy including budget allocation. Psychological influences such as overconfidence in assessing uncertainties and dependencies among cost elements and risks are other important considerations that are generally not addressed. It should then be no surprise that actual project costs often exceed the initial estimates and are delivered late and/or with a reduced scope. This paper presents a practical probabilistic cost analysis model that incorporates recent findings in human behavior and judgment under uncertainty, dependencies among cost elements, the MAIMS principle, and project management practices. Uncertain cost elements are elicited from experts using the direct fractile assessment method and fitted with three-parameter Weibull distributions. The full correlation matrix is specified in terms of two parameters that characterize correlations among cost elements in the same and in different subsystems. The analysis is readily implemented using standard Monte Carlo simulation tools such as {at}Risk and Crystal Ball{reg_sign}. The analysis of a representative design and engineering project substantiates that today's typical probabilistic cost analysis is likely to severely underestimate project cost for probability of success values of importance to contractors and procuring activities. The proposed approach provides a framework for developing a viable cost management strategy for

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)

Site-specific Probabilistic Analysis of DCGLs Using RESRAD Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Jeongju; Yoon, Suk Bon; Sohn, Wook [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In general, DCGLs can be conservative (screening DCGL) if they do not take into account site specific factors. Use of such conservative DCGLs can lead to additional remediation that would not be required if the effort was made to develop site-specific DCGLs. Therefore, the objective of this work is to provide an example on the use of the RESRAD 6.0 probabilistic (site-specific) dose analysis to compare with the screening DCGL. Site release regulations state that a site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group of less than the site release criteria, for example 0.25 mSv per year in U.S. Utilities use computer dose modeling codes to establish an acceptable level of contamination, the derived concentration guideline level (DCGL) that will meet this regulatory limit. Since the DCGL value is the principal measure of residual radioactivity, it is critical to understand the technical basis of these dose modeling codes. The objective this work was to provide example on nuclear power plant decommissioning dose analysis in a probabilistic analysis framework. The focus was on the demonstration of regulatory compliance for surface soil contamination using the RESRAD 6.0 code. Both the screening and site-specific probabilistic dose analysis methodologies were examined. Example analyses performed with the screening probabilistic dose analysis confirmed the conservatism of the NRC screening values and indicated the effectiveness of probabilistic dose analysis in reducing the conservatism in DCGL derivation.

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

Reactor safety

International Nuclear Information System (INIS)

Butz, H.P.; Heuser, F.W.; May, H.

1985-01-01

The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

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)

A Sensitivity Study for an Evaluation of Input Parameters Effect on a Preliminary Probabilistic Tsunami Hazard Analysis

Energy Technology Data Exchange (ETDEWEB)

Rhee, Hyun-Me; Kim, Min Kyu; Choi, In-Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Sheen, Dong-Hoon [Chonnam National University, Gwangju (Korea, Republic of)

2014-10-15

The tsunami hazard analysis has been based on the seismic hazard analysis. The seismic hazard analysis has been performed by using the deterministic method and the probabilistic method. To consider the uncertainties in hazard analysis, the probabilistic method has been regarded as attractive approach. The various parameters and their weight are considered by using the logic tree approach in the probabilistic method. The uncertainties of parameters should be suggested by analyzing the sensitivity because the various parameters are used in the hazard analysis. To apply the probabilistic tsunami hazard analysis, the preliminary study for the Ulchin NPP site had been performed. The information on the fault sources which was published by the Atomic Energy Society of Japan (AESJ) had been used in the preliminary study. The tsunami propagation was simulated by using the TSUNAMI{sub 1}.0 which was developed by Japan Nuclear Energy Safety Organization (JNES). The wave parameters have been estimated from the result of tsunami simulation. In this study, the sensitivity analysis for the fault sources which were selected in the previous studies has been performed. To analyze the effect of the parameters, the sensitivity analysis for the E3 fault source which was published by AESJ was performed. The effect of the recurrence interval, the potential maximum magnitude, and the beta were suggested by the sensitivity analysis results. Level of annual exceedance probability has been affected by the recurrence interval.. Wave heights have been influenced by the potential maximum magnitude and the beta. In the future, the sensitivity analysis for the all fault sources in the western part of Japan which were published AESJ would be performed.

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

GUI program to compute probabilistic seismic hazard analysis

International Nuclear Information System (INIS)

Shin, Jin Soo; Chi, H. C.; Cho, J. C.; Park, J. H.; Kim, K. G.; Im, I. S.

2005-12-01

The first stage of development of program to compute probabilistic seismic hazard is completed based on Graphic User Interface (GUI). The main program consists of three part - the data input processes, probabilistic seismic hazard analysis and result output processes. The first part has developed and others are developing now in this term. The probabilistic seismic hazard analysis needs various input data which represent attenuation formulae, seismic zoning map, and earthquake event catalog. The input procedure of previous programs based on text interface take a much time to prepare the data. The data cannot be checked directly on screen to prevent input erroneously in existing methods. The new program simplifies the input process and enable to check the data graphically in order to minimize the artificial error within the limits of the possibility

Probabilistic modeling of timber structures

DEFF Research Database (Denmark)

Köhler, Jochen; Sørensen, John Dalsgaard; Faber, Michael Havbro

2007-01-01

The present paper contains a proposal for the probabilistic modeling of timber material properties. It is produced in the context of the Probabilistic Model Code (PMC) of the Joint Committee on Structural Safety (JCSS) [Joint Committee of Structural Safety. Probabilistic Model Code, Internet...... Publication: www.jcss.ethz.ch; 2001] and of the COST action E24 ‘Reliability of Timber Structures' [COST Action E 24, Reliability of timber structures. Several meetings and Publications, Internet Publication: http://www.km.fgg.uni-lj.si/coste24/coste24.htm; 2005]. The present proposal is based on discussions...... and comments from participants of the COST E24 action and the members of the JCSS. The paper contains a description of the basic reference properties for timber strength parameters and ultimate limit state equations for timber components. The recommended probabilistic model for these basic properties...

Advances in probabilistic risk analysis

International Nuclear Information System (INIS)

Hardung von Hardung, H.

1982-01-01

Probabilistic risk analysis can now look back upon almost a quarter century of intensive development. The early studies, whose methods and results are still referred to occasionally, however, only permitted rough estimates to be made of the probabilities of recognizable accident scenarios, failing to provide a method which could have served as a reference base in calculating the overall risk associated with nuclear power plants. The first truly solid attempt was the Rasmussen Study and, partly based on it, the German Risk Study. In those studies, probabilistic risk analysis has been given a much more precise basis. However, new methodologies have been developed in the meantime, which allow much more informative risk studies to be carried out. They have been found to be valuable tools for management decisions with respect to backfitting, reinforcement and risk limitation. Today they are mainly applied by specialized private consultants and have already found widespread application especially in the USA. (orig.) [de

AST-500 safety analysis experience

Energy Technology Data Exchange (ETDEWEB)

Falikov, A A; Bakhmetiev, A M; Kuul, V S; Samoilov, O B [OKBM, Nizhny Novgorod (Russian Federation)

1997-09-01

Characteristic AST-type NHR safety features and requirements are described briefly. The main approaches and results of design and beyond-design accidents analyses for the AST-500 NHR, and the results of probabilistic safety assessments are considered. It is concluded that the AST-500 possesses a high safety level in virtue of the development and realization in the design of self-protection, passivity and defence-in-depth principles. (author). 9 refs, 2 figs.

Risk-Informed Safety Margin Characterization (RISMC): Integrated Treatment of Aleatory and Epistemic Uncertainty in Safety Analysis

International Nuclear Information System (INIS)

Youngblood, R.W.

2010-01-01

The concept of 'margin' has a long history in nuclear licensing and in the codification of good engineering practices. However, some traditional applications of 'margin' have been carried out for surrogate scenarios (such as design basis scenarios), without regard to the actual frequencies of those scenarios, and have been carried out with in a systematically conservative fashion. This means that the effectiveness of the application of the margin concept is determined in part by the original choice of surrogates, and is limited in any case by the degree of conservatism imposed on the evaluation. In the RISMC project, which is part of the Department of Energy's 'Light Water Reactor Sustainability Program' (LWRSP), we are developing a risk-informed characterization of safety margin. Beginning with the traditional discussion of 'margin' in terms of a 'load' (a physical challenge to system or component function) and a 'capacity' (the capability of that system or component to accommodate the challenge), we are developing the capability to characterize probabilistic load and capacity spectra, reflecting both aleatory and epistemic uncertainty in system response. For example, the probabilistic load spectrum will reflect the frequency of challenges of a particular severity. Such a characterization is required if decision-making is to be informed optimally. However, in order to enable the quantification of probabilistic load spectra, existing analysis capability needs to be extended. Accordingly, the INL is working on a next-generation safety analysis capability whose design will allow for much more efficient parameter uncertainty analysis, and will enable a much better integration of reliability-related and phenomenology-related aspects of margin.

SHEAT: a computer code for probabilistic seismic hazard analysis, user's manual

International Nuclear Information System (INIS)

Ebisawa, Katsumi; Kondo, Masaaki; Abe, Kiyoharu; Tanaka, Toshiaki; Takani, Michio.

1994-08-01

The SHEAT code developed at Japan Atomic Energy Research Institute is for probabilistic seismic hazard analysis which is one of the tasks needed for seismic Probabilistic Safety Assessment (PSA) of a nuclear power plant. Seismic hazard is defined as an annual exceedance frequency of occurrence of earthquake ground motions at various levels of intensity at a given site. With the SHEAT code, seismic hazard is calculated by the following two steps: (1) Modeling of earthquake generation around a site. Future earthquake generation (locations, magnitudes and frequencies of postulated earthquakes) is modelled based on the historical earthquake records, active fault data and expert judgement. (2) Calculation of probabilistic seismic hazard at the site. An earthquake ground motion is calculated for each postulated earthquake using an attenuation model taking into account its standard deviation. Then the seismic hazard at the site is calculated by summing the frequencies of ground motions by all the earthquakes. This document is the user's manual of the SHEAT code. It includes: (1) Outlines of the code, which include overall concept, logical process, code structure, data file used and special characteristics of the code, (2) Functions of subprograms and analytical models in them, (3) Guidance of input and output data, and (4) Sample run results. The code has widely been used at JAERI to analyze seismic hazard at various nuclear power plant sites in japan. (author)

Probabilistic analysis and related topics

CERN Document Server

Bharucha-Reid, A T

1979-01-01

Probabilistic Analysis and Related Topics, Volume 2 focuses on the integrability, continuity, and differentiability of random functions, as well as functional analysis, measure theory, operator theory, and numerical analysis.The selection first offers information on the optimal control of stochastic systems and Gleason measures. Discussions focus on convergence of Gleason measures, random Gleason measures, orthogonally scattered Gleason measures, existence of optimal controls without feedback, random necessary conditions, and Gleason measures in tensor products. The text then elaborates on an

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

Probabilistic Design Analysis (PDA) Approach to Determine the Probability of Cross-System Failures for a Space Launch Vehicle

Science.gov (United States)

Shih, Ann T.; Lo, Yunnhon; Ward, Natalie C.

2010-01-01

Quantifying the probability of significant launch vehicle failure scenarios for a given design, while still in the design process, is critical to mission success and to the safety of the astronauts. Probabilistic risk assessment (PRA) is chosen from many system safety and reliability tools to verify the loss of mission (LOM) and loss of crew (LOC) requirements set by the NASA Program Office. To support the integrated vehicle PRA, probabilistic design analysis (PDA) models are developed by using vehicle design and operation data to better quantify failure probabilities and to better understand the characteristics of a failure and its outcome. This PDA approach uses a physics-based model to describe the system behavior and response for a given failure scenario. Each driving parameter in the model is treated as a random variable with a distribution function. Monte Carlo simulation is used to perform probabilistic calculations to statistically obtain the failure probability. Sensitivity analyses are performed to show how input parameters affect the predicted failure probability, providing insight for potential design improvements to mitigate the risk. The paper discusses the application of the PDA approach in determining the probability of failure for two scenarios from the NASA Ares I project

Probabilistic methods in nuclear power plant component ageing analysis

International Nuclear Information System (INIS)

Simola, K.

1992-03-01

The nuclear power plant ageing research is aimed to ensure that the plant safety and reliability are maintained at a desired level through the designed, and possibly extended lifetime. In ageing studies, the reliability of components, systems and structures is evaluated taking into account the possible time- dependent decrease in reliability. The results of analyses can be used in the evaluation of the remaining lifetime of components and in the development of preventive maintenance, testing and replacement programmes. The report discusses the use of probabilistic models in the evaluations of the ageing of nuclear power plant components. The principles of nuclear power plant ageing studies are described and examples of ageing management programmes in foreign countries are given. The use of time-dependent probabilistic models to evaluate the ageing of various components and structures is described and the application of models is demonstrated with two case studies. In the case study of motor- operated closing valves the analysis are based on failure data obtained from a power plant. In the second example, the environmentally assisted crack growth is modelled with a computer code developed in United States, and the applicability of the model is evaluated on the basis of operating experience

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

Sargent-IV Project. Development of new methodologies for safety analysis of Generation IV reactors; Proyecto SARGEB-IV. Desarrollo de nuevas metodologias de analisis de seguridad para reactores de Generacion IV

Energy Technology Data Exchange (ETDEWEB)

Queral, C.; Gallego, E.; Jimenez, G.

2013-07-01

The main result of this paper is the proposal for the addition of new ingredients in the safety analysis methodologies for Generation-IV reactors that integrates the features of probabilistic safety analysis within deterministic. This ensures a higher degree of integration between the classical deterministic and probabilistic methodologies.

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Applications of probabilistic techniques at NRC

International Nuclear Information System (INIS)

Thadani, A.; Rowsome, F.; Speis, T.

1984-01-01

The NRC is currently making extensive use of probabilistic safety assessment in the reactor regulation. Most of these applications have been introduced in the regulatory activities in the past few years. Plant Probabilistic Safety Studies are being utilized as a design tool for applications for standard designs and for assessment of plants located in regions of particularly high population density. There is considerable motivation for licenses to perform plant-specific probabilistic studies for many, if not all, of the existing operating nuclear power plants as a tool for prioritizing the implementation of the many outstanding licensing actions of these plants as well as recommending the elimination of a number of these issues which are judged to be insignificant in terms of their contribution to safety and risk. Risk assessment perspectives are being used in the priorization of generic safety issues, development of technical resolution of unresolved safety issues, assessing safety significance of proposed new regulatory requirements, assessment of safety significance of some of the occurrences at operating facilities and in environmental impact analyses of license applicants as required by the National Environmental Policy Act. (orig.)

Probabilistic analysis of external events with focus on the Fukushima event

International Nuclear Information System (INIS)

Kollasko, Heiko; Jockenhoevel-Barttfeld, Mariana; Klapp, Ulrich

2014-01-01

External hazards are those natural or man-made hazards to a site and facilities that are originated externally to both the site and its processes, i.e. the duty holder may have very little or no control over the hazard. External hazards can have the potential of causing initiating events at the plant, typically transients like e.g., loss of offsite power. Simultaneously, external events may affect safety systems required to control the initiating event and, where applicable, also back-up systems implemented for risk-reduction. The plant safety may especially be threatened when loads from external hazards exceed the load assumptions considered in the design of safety-related systems, structures and components. Another potential threat is given by hazards inducing initiating events not considered in the safety demonstration otherwise. An example is loss of offsite power combined with prolonged plant isolation. Offsite support, e.g., delivery of diesel fuel oil, usually credited in the deterministic safety analysis may not be possible in this case. As the Fukushima events have shown, the biggest threat is likely given by hazards inducing both effects. Such hazards may well be dominant risk contributors even if their return period is very high. In order to identify relevant external hazards for a certain Nuclear Power Plant (NPP) location, a site specific screening analysis is performed, both for single events and for combinations of external events. As a result of the screening analysis, risk significant and therefore relevant (screened-in) single external events and combinations of them are identified for a site. The screened-in events are further considered in a detailed event tree analysis in the frame of the Probabilistic Safety Analysis (PSA) to calculate the core damage/large release frequency resulting from each relevant external event or from each relevant combination. Screening analyses of external events performed at AREVA are based on the approach provided

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

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.

Japanese round robin analysis for probabilistic fracture mechanics

International Nuclear Information System (INIS)

Yagawa, G.; Yoshimura, S.; Handa, N.

1991-01-01

Recently attention is focused on the probabilistic fracture mechanics, a branch of fracture mechanics with probability theory for a rational mean to assess the strength of components and structures. In particular, the probabilistic fracture mechanics is recognized as the powerful means for quantitative investigation of significance of factors and rational evaluation of life on problems involving a number of uncertainties, such as degradation of material strength, accuracy and frequency of inspection. Comparison with reference experiments are generally employed to assure the analytical accuracy. However, accuracy and reliability of analytical methods in the probabilistic fracture mechanics are hardly verified by experiments. Therefore, it is strongly needed to verify the probabilistic fracture mechanics through the round robin analysis. This paper describes results from the round robin analysis of flat plate with semi-elliptic cracks on the surface, conducted by the PFM Working Group of LE Subcommittee of the Japan Welding Society under the contract of the Japan Atomic Energy Research Institute and participated by Tokyo University, Yokohama National University, the Power Reactor and Nuclear Fuel Corporation, Tokyo Electric Power Co. Central Research Institute of Electric Power Industry, Toshiba Corporation, Kawasaki Heavy Industry Co. and Mitsubishi Heavy Industry Co. (author)

Probabilistic and sensitivity analysis of Botlek Bridge structures

Directory of Open Access Journals (Sweden)

Králik Juraj

2017-01-01

Full Text Available This paper deals with the probabilistic and sensitivity analysis of the largest movable lift bridge of the world. The bridge system consists of six reinforced concrete pylons and two steel decks 4000 tons weight each connected through ropes with counterweights. The paper focuses the probabilistic and sensitivity analysis as the base of dynamic study in design process of the bridge. The results had a high importance for practical application and design of the bridge. The model and resistance uncertainties were taken into account in LHS simulation method.

The application of new mathematical structures to safety analysis

International Nuclear Information System (INIS)

Cooper, J.A.; Ross, T.J.

1997-10-01

Probabilistic safety analyses (PSAs) often depend on significant subjectivity. The recent successes of fuzzy logic and fuzzy and hybrid mathematics in portraying subjectivity is a reminder that a selection made from the most applicable mathematical tools is more important than forced adaptation of conventional tools. In this paper, the authors consider new approaches that enhance conventional and fuzzy PSA by improved handling of subjectivity. The most significant of the mathematical structures were have investigated (from a standpoint of safety analysis applications) will be described, and the general types of applications will be outlined

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

Probabilistic analysis of a materially nonlinear structure

Science.gov (United States)

Millwater, H. R.; Wu, Y.-T.; Fossum, A. F.

1990-01-01

A probabilistic finite element program is used to perform probabilistic analysis of a materially nonlinear structure. The program used in this study is NESSUS (Numerical Evaluation of Stochastic Structure Under Stress), under development at Southwest Research Institute. The cumulative distribution function (CDF) of the radial stress of a thick-walled cylinder under internal pressure is computed and compared with the analytical solution. In addition, sensitivity factors showing the relative importance of the input random variables are calculated. Significant plasticity is present in this problem and has a pronounced effect on the probabilistic results. The random input variables are the material yield stress and internal pressure with Weibull and normal distributions, respectively. The results verify the ability of NESSUS to compute the CDF and sensitivity factors of a materially nonlinear structure. In addition, the ability of the Advanced Mean Value (AMV) procedure to assess the probabilistic behavior of structures which exhibit a highly nonlinear response is shown. Thus, the AMV procedure can be applied with confidence to other structures which exhibit nonlinear behavior.

Probabilistic structural analysis using a general purpose finite element program

Science.gov (United States)

Riha, D. S.; Millwater, H. R.; Thacker, B. H.

1992-07-01

This paper presents an accurate and efficient method to predict the probabilistic response for structural response quantities, such as stress, displacement, natural frequencies, and buckling loads, by combining the capabilities of MSC/NASTRAN, including design sensitivity analysis and fast probability integration. Two probabilistic structural analysis examples have been performed and verified by comparison with Monte Carlo simulation of the analytical solution. The first example consists of a cantilevered plate with several point loads. The second example is a probabilistic buckling analysis of a simply supported composite plate under in-plane loading. The coupling of MSC/NASTRAN and fast probability integration is shown to be orders of magnitude more efficient than Monte Carlo simulation with excellent accuracy.

Extending CANTUP code analysis to probabilistic evaluations

International Nuclear Information System (INIS)

Florea, S.

2001-01-01

The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented

Probabilistic Structural Analysis of SSME Turbopump Blades: Probabilistic Geometry Effects

Science.gov (United States)

Nagpal, V. K.

1985-01-01

A probabilistic study was initiated to evaluate the precisions of the geometric and material properties tolerances on the structural response of turbopump blades. To complete this study, a number of important probabilistic variables were identified which are conceived to affect the structural response of the blade. In addition, a methodology was developed to statistically quantify the influence of these probabilistic variables in an optimized way. The identified variables include random geometric and material properties perturbations, different loadings and a probabilistic combination of these loadings. Influences of these probabilistic variables are planned to be quantified by evaluating the blade structural response. Studies of the geometric perturbations were conducted for a flat plate geometry as well as for a space shuttle main engine blade geometry using a special purpose code which uses the finite element approach. Analyses indicate that the variances of the perturbations about given mean values have significant influence on the response.

Developing reports on safety analysis and probabilistic analysis of safety for operating power units at nuclear power stations with WWER reactors in Russia; Razrabotka otchetov po analizu bezopasnosti i VAB dlya ehkspluatiruyushchikhsya ehnergoblokov AEhS s WWEhR v Rossii

Energy Technology Data Exchange (ETDEWEB)

Malyshev, A B; Morozov, V B [ATOMENERGOPROEKT Institute, Moscow (Russian Federation)

1999-06-01

Report presents the current state-of art in developing safety reports and probabilistic safety analyses for WWER NPPs operated in Russia. Development of these reports and implementation of PSA is done according to the requirements outlined in the basic document `General Statement on Ensuring safety (OPB). At present submitting safety reports to the regulatory authority GAN RF is mandatory for licensing NPPs. Current state of safety reports for the operating WWER type NPPs meets generally the effective Russian standard engineering documents which are approaching the international standards. A mechanism ensuring correspondence of the safety documentation to the current state of operating units is determined. Modernization of the operating units is underway, it is aimed to eliminate existing deviations from requirements of the modern standards in the field of NPP safety

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

Response to Yellman and Murray's comment on 'The meaning of probability in probabilistic risk analysis'

International Nuclear Information System (INIS)

Watson, Stephen R.

1995-01-01

In their comment on a recent contribution of mine, [Watson, S., The meaning of probability in probabilistic safety analysis. Reliab. Engng and System Safety, 45 (1994) 261-269.] Yellman and Murray assert that (1) I argue in favour of a realistic interpretation of probability for PSAs; (2) that the only satisfactory philosophical theory of probability is the relative frequency theory; (3) that I mean the same thing by the words 'uncertainty' and 'probability'; (4) that my argument can easily lead to the belief that the output of PSAs are meaningless. I take issue with all these points, and in this response I set out my arguments

Probabilistic design of nuclear structures: a summary of state of the art and research needs

International Nuclear Information System (INIS)

Ravindra, M.K.; Walser, A.

1978-01-01

This paper provides an overview of ongoing research in probabilistic design of nuclear structures. The main areas of review are (1) loads, (2) load combinations, (3) missiles, (4) design criteria, (5) seismic safety, (6) system reliability, (7) hazard analysis, and (8) probabilistic response. A consistent framework of probabilistic design of nuclear structures is proposed. Areas of further research and data collection are suggested. (Auth.)

Safety analysis methodologies for radioactive waste repositories in shallow ground

International Nuclear Information System (INIS)

1984-01-01

The report is part of the IAEA Safety Series and is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of shallow ground radioactive waste repositories. It discusses approaches that are applicable for safety analysis of a shallow ground repository. The methodologies, analysis techniques and models described are pertinent to the task of predicting the long-term performance of a shallow ground disposal system. They may be used during the processes of selection, confirmation and licensing of new sites and disposal systems or to evaluate the long-term consequences in the post-sealing phase of existing operating or inactive sites. The analysis may point out need for remedial action, or provide information to be used in deciding on the duration of surveillance. Safety analysis both general in nature and specific to a certain repository, site or design concept, are discussed, with emphasis on deterministic and probabilistic studies

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.

Volume 2. Probabilistic analysis of HTGR application studies. Supporting data

International Nuclear Information System (INIS)

1980-09-01

Volume II, Probabilistic Analysis of HTGR Application Studies - Supporting Data, gives the detail data, both deterministic and probabilistic, employed in the calculation presented in Volume I. The HTGR plants and the fossil plants considered in the study are listed. GCRA provided the technical experts from which the data were obtained by MAC personnel. The names of the technical experts (interviewee) and the analysts (interviewer) are given for the probabilistic data

Probabilistic analysis and related topics

CERN Document Server

Bharucha-Reid, A T

1983-01-01

Probabilistic Analysis and Related Topics, Volume 3 focuses on the continuity, integrability, and differentiability of random functions, including operator theory, measure theory, and functional and numerical analysis. The selection first offers information on the qualitative theory of stochastic systems and Langevin equations with multiplicative noise. Discussions focus on phase-space evolution via direct integration, phase-space evolution, linear and nonlinear systems, linearization, and generalizations. The text then ponders on the stability theory of stochastic difference systems and Marko

An approximate methods approach to probabilistic structural analysis

Science.gov (United States)

Mcclung, R. C.; Millwater, H. R.; Wu, Y.-T.; Thacker, B. H.; Burnside, O. H.

1989-01-01

A probabilistic structural analysis method (PSAM) is described which makes an approximate calculation of the structural response of a system, including the associated probabilistic distributions, with minimal computation time and cost, based on a simplified representation of the geometry, loads, and material. The method employs the fast probability integration (FPI) algorithm of Wu and Wirsching. Typical solution strategies are illustrated by formulations for a representative critical component chosen from the Space Shuttle Main Engine (SSME) as part of a major NASA-sponsored program on PSAM. Typical results are presented to demonstrate the role of the methodology in engineering design and analysis.

Application of probabilistic methods to accident analysis at waste management facilities

International Nuclear Information System (INIS)

Banz, I.

1986-01-01

Probabilistic risk assessment is a technique used to systematically analyze complex technical systems, such as nuclear waste management facilities, in order to identify and measure their public health, environmental, and economic risks. Probabilistic techniques have been utilized at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, to evaluate the probability of a catastrophic waste hoist accident. A probability model was developed to represent the hoisting system, and fault trees were constructed to identify potential sequences of events that could result in a hoist accident. Quantification of the fault trees using statistics compiled by the Mine Safety and Health Administration (MSHA) indicated that the annual probability of a catastrophic hoist accident at WIPP is less than one in 60 million. This result allowed classification of a catastrophic hoist accident as ''not credible'' at WIPP per DOE definition. Potential uses of probabilistic techniques at other waste management facilities are discussed

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

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

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

Probabilistic and deterministic soil structure interaction analysis including ground motion incoherency effects

International Nuclear Information System (INIS)

Elkhoraibi, T.; Hashemi, A.; Ostadan, F.

2014-01-01

Soil-structure interaction (SSI) is a major step for seismic design of massive and stiff structures typical of the nuclear facilities and civil infrastructures such as tunnels, underground stations, dams and lock head structures. Currently most SSI analyses are performed deterministically, incorporating limited range of variation in soil and structural properties and without consideration of the ground motion incoherency effects. This often leads to overestimation of the seismic response particularly the In-Structure-Response Spectra (ISRS) with significant impositions of design and equipment qualification costs, especially in the case of high-frequency sensitive equipment at stiff soil or rock sites. The reluctance to incorporate a more comprehensive probabilistic approach is mainly due to the fact that the computational cost of performing probabilistic SSI analysis even without incoherency function considerations has been prohibitive. As such, bounding deterministic approaches have been preferred by the industry and accepted by the regulatory agencies. However, given the recently available and growing computing capabilities, the need for a probabilistic-based approach to the SSI analysis is becoming clear with the advances in performance-based engineering and the utilization of fragility analysis in the decision making process whether by the owners or the regulatory agencies. This paper demonstrates the use of both probabilistic and deterministic SSI analysis techniques to identify important engineering demand parameters in the structure. A typical nuclear industry structure is used as an example for this study. The system is analyzed for two different site conditions: rock and deep soil. Both deterministic and probabilistic SSI analysis approaches are performed, using the program SASSI, with and without ground motion incoherency considerations. In both approaches, the analysis begins at the hard rock level using the low frequency and high frequency hard rock

Probabilistic and deterministic soil structure interaction analysis including ground motion incoherency effects

Energy Technology Data Exchange (ETDEWEB)

Elkhoraibi, T., E-mail: telkhora@bechtel.com; Hashemi, A.; Ostadan, F.

2014-04-01

Soil-structure interaction (SSI) is a major step for seismic design of massive and stiff structures typical of the nuclear facilities and civil infrastructures such as tunnels, underground stations, dams and lock head structures. Currently most SSI analyses are performed deterministically, incorporating limited range of variation in soil and structural properties and without consideration of the ground motion incoherency effects. This often leads to overestimation of the seismic response particularly the In-Structure-Response Spectra (ISRS) with significant impositions of design and equipment qualification costs, especially in the case of high-frequency sensitive equipment at stiff soil or rock sites. The reluctance to incorporate a more comprehensive probabilistic approach is mainly due to the fact that the computational cost of performing probabilistic SSI analysis even without incoherency function considerations has been prohibitive. As such, bounding deterministic approaches have been preferred by the industry and accepted by the regulatory agencies. However, given the recently available and growing computing capabilities, the need for a probabilistic-based approach to the SSI analysis is becoming clear with the advances in performance-based engineering and the utilization of fragility analysis in the decision making process whether by the owners or the regulatory agencies. This paper demonstrates the use of both probabilistic and deterministic SSI analysis techniques to identify important engineering demand parameters in the structure. A typical nuclear industry structure is used as an example for this study. The system is analyzed for two different site conditions: rock and deep soil. Both deterministic and probabilistic SSI analysis approaches are performed, using the program SASSI, with and without ground motion incoherency considerations. In both approaches, the analysis begins at the hard rock level using the low frequency and high frequency hard rock

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.

On the Probabilistic Characterization of Robustness and Resilience

DEFF Research Database (Denmark)

Faber, Michael Havbro; Qin, J.; Miraglia, Simona

2017-01-01

Over the last decade significant research efforts have been devoted to the probabilistic modeling and analysis of system characteristics. Especially performance characteristics of systems subjected to random disturbances, such as robustness and resilience have been in the focus of these efforts...... in the modeling of robustness and resilience in the research areas of natural disaster risk management, socio-ecological systems and social systems and we propose a generic decision analysis framework for the modeling and analysis of systems across application areas. The proposed framework extends the concept...... of direct and indirect consequences and associated risks in probabilistic systems modeling formulated by the Joint Committee on Structural Safety (JCSS) to facilitate the modeling and analysis of resilience in addition to robustness and vulnerability. Moreover, based on recent insights in the modeling...

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.

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)

Probabilistic Accident Progression Analysis with application to a LMFBR design