Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0: Models and Results Database (MAR-D) reference manual. Volume 8

International Nuclear Information System (INIS)

Russell, K.D.; Skinner, N.L.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The primary function of MAR-D is to create a data repository for completed PRAs and Individual Plant Examinations (IPEs) by providing input, conversion, and output capabilities for data used by IRRAS, SARA, SETS, and FRANTIC software. As probabilistic risk assessments and individual plant examinations are submitted to the NRC for review, MAR-D can be used to convert the models and results from the study for use with IRRAS and SARA. Then, these data can be easily accessed by future studies and will be in a form that will enhance the analysis process. This reference manual provides an overview of the functions available within MAR-D and step-by-step operating instructions

Strangeness photoproduction with the SAPHIR-detector

International Nuclear Information System (INIS)

Merkel, H.

1993-12-01

At the ELSA facility at Bonn a photon beam with a high duty cycle up to energies of 3.3 GeV is available. In this energy range the large solid angle detector SAPHIR enables us to investigate the strangeness photoproduction starting from threshold. SAPHIR has already achieved results for the reactions γ+p→K + +Λ and γ+p→K + +Σ 0 . This work investigates the possibilities to measure the related reactions γ+n→K 0 +Λ and γ+n→K 0 +Σ 0 at a deuteron target and to measure the reaction γ+p→K 0 +Σ + at a proton target. For the first time the Σ + polarisation has been measured. With an cross section 10 times smaller compared to the kaon hyperon reactions, the photoproduction of the Φ(1020) meson can be investigated with the SAPHIR detector too. First reconstructed events are shown. (orig.)

All-sky radiance simulation of Megha-Tropiques SAPHIR microwave ...

Indian Academy of Sciences (India)

used as input to the RTTOV model to simulate cloud-affected SAPHIR radiances. ... All-sky radiance simulation; Megha tropiques; microwave SAPHIR sensor; radiative transfer; data ... versions of these non-linear processes (Ohring and.

Development and verification of a leningrad NPP unit 1 living PSA model in the INL SAPHIRE code format for prompt operational safety level monitoring

International Nuclear Information System (INIS)

Bronislav, Vinnikov

2007-01-01

The first part of the paper presents results of the work, that was carried out in complete conformity with the Technical Assignment, which was developed by the Leningrad Nuclear Power Plant. The initial scientific and technical information, contained into the In-Depth Safety Assessment Reports, was given to the author of the work. This information included graphical Fault Trees of Safety Systems and Auxiliary Technical Systems, Event Trees for the necessary number of Initial Events, and also information about failure probabilities of basic components of the nuclear unit. On the basis of this information and fueling it to the Usa Idaho National Laboratory (INL) SAPHIRE code, we have developed an electronic version of the Data Base for failure probabilities of the components of technical systems. Then, we have developed both the electronic versions of the necessary Fault Trees, and an electronic versions of the necessary Event Trees. And at last, we have carried out the linkage of the Event Trees. This work has resulted in the Living PSA (LPSA - Living Probabilistic Safety Assessment) Model of the Leningrad NPP Unit 1. The LPSA-model is completely adapted to be consistent with the USA INL SAPHIRE Risk Monitor. The second part of the paper results in analysis of fire consequences in various places of Leningrad NPP Unit 1. The computations were carried out with the help of the LPSA-model, developed in SAPHIRE code format. On the basis of the computations the order of priority of implementation of fire prevention measures was established. (author)

SAPhIR: a fission-fragment detector

International Nuclear Information System (INIS)

Theisen, Ch.; Gautherin, C.; Houry, M.; Korten, W.; Le Coz, Y.; Lucas, R.; Barreau, G.; Doan, T. P.; Belier, G.; Meot, V.; Ethvignot, Th.; Cahan, B.; Le Coguie, A.; Coppolani, X.; Delaitre, B.; Le Bourlout, P.; Legou, Ph.; Maillard, O.; Durand, G.; Bouillac, A.

1998-01-01

SAPhIR is the acronym for S a clay A q uitaine P ho tovoltaic cells for I s omer R e search. It consists of solar cells, used for fission-fragment detection. It is a collaboration between 3 laboratories: CEA Saclay, CENBG Bordeaux and CEA Bruyeres le Chatel. The coupling of a highly efficient fission-fragment detector like SAPhIR with EUROBALL will provide new insights in the study of very deformed nuclear matter and in the spectroscopy of neutron-rich nuclei

Nucleonic calculations for possible irradiation experiments in SAPHIR

International Nuclear Information System (INIS)

Caro, M.; Pelloni, S.

1990-01-01

Accurate two-dimensional calculations show that a 'neutronic environment' exists in the SAPHIR reactor at the Paul Scherrer Institute (PSI) to simulate the inner surface of a given trepan of the Gundremmingen reactor. Neutron fluences and DPA rates were calculated at two positions in SAPHIR using the modern codes and nuclear data (from JEF-1). A particular region of the reactor can be found in which fluences and DPA rates agree well within a few percent with the Gundremmingen reference case. (author) 13 figs., 4 tabs., 18 refs

Systems analysis programs for hands-on integrated reliability evaluations (SAPHIRE) Version 5.0. Fault tree, event tree, and piping ampersand instrumentation diagram (FEP) editors reference manual: Volume 7

International Nuclear Information System (INIS)

McKay, M.K.; Skinner, N.L.; Wood, S.T.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Fault Tree, Event Tree, and Piping and Instrumentation Diagram (FEP) editors allow the user to graphically build and edit fault trees, and event trees, and piping and instrumentation diagrams (P and IDs). The software is designed to enable the independent use of the graphical-based editors found in the Integrated Reliability and Risk Assessment System (IRRAS). FEP is comprised of three separate editors (Fault Tree, Event Tree, and Piping and Instrumentation Diagram) and a utility module. This reference manual provides a screen-by-screen guide of the entire FEP System

The atmosphere simulation chamber SAPHIR: a tool for the investigation of photochemistry.

Science.gov (United States)

Brauers, T.; Bohn, B.; Johnen, F.-J.; Rohrer, R.; Rodriguez Bares, S.; Tillmann, R.; Wahner, A.

2003-04-01

On the campus of the Forschungszentrum Jülich we constructed SAPHIR (Simulation of Atmospheric PHotochemistry In a large Reaction Chamber) which was accomplished in fall 2001. The chamber consists of a 280-m^3 double-wall Teflon bag of cylindrical shape that is held by a steel frame. Typically 75% of the outside actinic flux (290~nm~--~420~nm) is available inside the chamber. A louvre system allows switching between full sun light and dark within 40 s giving the opportunity to study relaxation processes of the photo chemical system. The SAPHIR chamber is equipped with a comprehensive set of sensitive instruments including the measurements of OH, HO_2, CO, hydrocarbons, aldehydes, nitrogen-oxides and solar radiation. Moreover, the modular concept of SAPHIR allows fast and flexible integration of new instruments and techniques. In this paper we will show the unique and new features of the SAPHIR chamber, namely the clean air supply and high purity water vapor supply providing a wide range of trace gas concentrations being accessible through the experiments. We will also present examples from the first year of SAPHIR experiment showing the scope of application from high quality instrument inter-comparison and kinetic studies to the simulation of complex mixtures of trace gases at ambient concentrations.

Integration of Probabilistic Exposure Assessment and Probabilistic Hazard Characterization

NARCIS (Netherlands)

Voet, van der H.; Slob, W.

2007-01-01

A method is proposed for integrated probabilistic risk assessment where exposure assessment and hazard characterization are both included in a probabilistic way. The aim is to specify the probability that a random individual from a defined (sub)population will have an exposure high enough to cause a

Development of the software of the data taking system SOS for the SAPHIR experiment. Entwicklung der Software des Datennahmesystems SOS fuer das SAPHIR-Experiment

Energy Technology Data Exchange (ETDEWEB)

Manns, J.

1989-02-01

The data acquistion system SOS has been developed for the SAPHIR experiment at the Bonn stretcher ring ELSA. It can handle up to 280 kilobytes of data per second or a maximum triggerrate of 200 Hz. The multiprocessor based online system consists of twenty VIP-microprocessors and two VAX-computers. Each component of the SAPHIR experiment has at least one program in the online system to maintain special functions for this specific component. All of these programs can receive event data without interfering with the transfer of events to a mass storage for offline analysis. A special program SOL has been developed to serve as a user interface to the data acquisition system and as a status display for most of the programs of the online system. Using modern features like windowing and mouse control on a VAX-station the SAPHIR online SOL establishes an easy way of controlling the data acquisition system. (orig.).

Simulation of the beam guiding of the SAPHIR experiment by means of a differential-equation model; Simulation der Strahlfuehrung des SAPHIR-Experiments mittels eines Differentialgleichungsmodells

Energy Technology Data Exchange (ETDEWEB)

Greve, T.

1991-08-01

This paper shows the numerical simulation of a beam line by means of a model of differential equations simulating the beam line from the Bonn Electron Stretcher Accelerator ELSA to the SAPHIR spectrometer. Furthermore a method for calculating the initial values based on measurements of beam profiles is being discussed. (orig.). [Deutsch] Diese Arbeit zeigt die numerische Simulation einer Strahlfuehrung mittels eines Differentialgleichungsmodells anhand der Strahlfuehrung vom Bonner ELSA-Beschleuniger zum SAPHIR-Experiment. Weiterhin wird eine Methode zur Gewinnung der Startwerte aus Strahlprofilmessungen diskutiert. (orig.).

The scintillation counter system at the SAPHIR detector

International Nuclear Information System (INIS)

Bour, D.

1989-10-01

The scintillation-counters system of the SAPHIR-detector at the stretcher accelerator ELSA in Bonn consists of 64 counters. It supplies a fast hadronic trigger and is utilizised for the particle identification by time of flight measurements. Prototypes of the counters (340x21.25 x 6.0 cm 3 ) had been tested. The contribution to the resolution of the time of flight measurement was measured to σ=125 ps, the effective light velocity to 17.5 ns/cm and the attenuation length of 7.8 m. A pion kaon separation is possible up to a momentum of 1 GeV/c with time of flight measurement. With the first photon-beam at SAPHIR the counters were tested, first triggers were obtained and evaluated. (orig.) [de

The trigger and data acquisition system of the SAPHIR detector

International Nuclear Information System (INIS)

Honscheid, K.

1988-10-01

At present SAPHIR, a new experimental facility for medium energy physics is under construction at the Bonn electron accelerator ELSA (energy ≤ 3.5 GeV, duty cycle ≅ 100%). SAPHIR combines a large solid angle coverage with a tagging system and is therefore suited to investigate reactions with multi-particle final states. Structure and function of the multi-stage trigger system, which is used to select such processes, are described in this paper. With this system the trigger decision can be based on the number of charged particles as well as on the number of neutral particle detected. Several VMEbus modules have been developed, using memory look-up tables to make fast trigger decisions possible. In order to determine the number of neutral particles from the cluster distribution in the electromagnetic calorimeter some ideas of cellular had to be added. The system has a modular structure, so it can easily be extended. In the second part of this thesis the SAPHIR data acquisition system is discussed. It consists of a multiprocessor system with the VIP microcomputer as central element. The VIP is a VMEbus modul optimized for a multiprocessor environment. Its description as well as that of the other VMEbus boards developed for the SAPHIR online system can be found in this paper. As a basis for software development the operating system SOS is supplied. With SOS it is possible to write programs independent of the actual hardware configuration and so the complicated multiprocessor environment is hidden. To the user the system looks like a simple multi-tasking system. SOS is not restricted to the VIPs but can also be installed on computers of the VAX family, so that efficient mixed configurations are possible. The SAPHIR online system, based on the VIP microcomputer and the SOS operating system, is presented in the last part of this paper. This includes the read-out system, the monitoring of the different components etc. (orig./HSI) [de

Implications of probabilistic risk assessment

International Nuclear Information System (INIS)

Cullingford, M.C.; Shah, S.M.; Gittus, J.H.

1987-01-01

Probabilistic risk assessment (PRA) is an analytical process that quantifies the likelihoods, consequences and associated uncertainties of the potential outcomes of postulated events. Starting with planned or normal operation, probabilistic risk assessment covers a wide range of potential accidents and considers the whole plant and the interactions of systems and human actions. Probabilistic risk assessment can be applied in safety decisions in design, licensing and operation of industrial facilities, particularly nuclear power plants. The proceedings include a review of PRA procedures, methods and technical issues in treating uncertainties, operating and licensing issues and future trends. Risk assessment for specific reactor types or components and specific risks (eg aircraft crashing onto a reactor) are used to illustrate the points raised. All 52 articles are indexed separately. (U.K.)

Independent Verification and Validation Of SAPHIRE 8 Software Requirements Project Number: N6423 U.S. Nuclear Regulatory Commission

Energy Technology Data Exchange (ETDEWEB)

Kent Norris

2009-09-01

The purpose of the Independent Verification and Validation (IV&V) role in the evaluation of the SAPHIRE requirements definition is to assess the activities that results in the specification, documentation, and review of the requirements that the software product must satisfy, including functionality, performance, design constraints, attributes and external interfaces. The IV&V team began this endeavor after the software engineering and software development of SAPHIRE had already been in production. IV&V reviewed the requirements specified in the NRC Form 189s to verify these requirements were included in SAPHIRE’s Software Verification and Validation Plan (SVVP).

Systems analysis programs for hands-on integrated reliability evaluations (SAPHIRE) version 5.0, technical reference manual

International Nuclear Information System (INIS)

Russell, K.D.; Atwood, C.L.; Galyean, W.J.; Sattison, M.B.; Rasmuson, D.M.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. This volume provides information on the principles used in the construction and operation of Version 5.0 of the Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) system. It summarizes the fundamental mathematical concepts of sets and logic, fault trees, and probability. This volume then describes the algorithms that these programs use to construct a fault tree and to obtain the minimal cut sets. It gives the formulas used to obtain the probability of the top event from the minimal cut sets, and the formulas for probabilities that are appropriate under various assumptions concerning repairability and mission time. It defines the measures of basic event importance that these programs can calculate. This volume gives an overview of uncertainty analysis using simple Monte Carlo sampling or Latin Hypercube sampling, and states the algorithms used by these programs to generate random basic event probabilities from various distributions. Further references are given, and a detailed example of the reduction and quantification of a simple fault tree is provided in an appendix

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)

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0: Integrated Reliability and Risk Analysis System (IRRAS) reference manual. Volume 2

International Nuclear Information System (INIS)

Russell, K.D.; Kvarfordt, K.J.; Skinner, N.L.; Wood, S.T.; Rasmuson, D.M.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the use the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification to report generation. Version 1.0 of the IRRAS program was released in February of 1987. Since then, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 5.0 and is the subject of this Reference Manual. Version 5.0 of IRRAS provides the same capabilities as earlier versions and ads the ability to perform location transformations, seismic analysis, and provides enhancements to the user interface as well as improved algorithm performance. Additionally, version 5.0 contains new alphanumeric fault tree and event used for event tree rules, recovery rules, and end state partitioning

Saphire models and software for ASP evaluations

International Nuclear Information System (INIS)

Sattison, M.B.

1997-01-01

The Idaho National Engineering Laboratory (INEL) over the three years has created 75 plant-specific Accident Sequence Precursor (ASP) models using the SAPHIRE suite of PRA codes. Along with the new models, the INEL has also developed a new module for SAPHIRE which is tailored specifically to the unique needs of ASP evaluations. These models and software will be the next generation of risk tools for the evaluation of accident precursors by both the U.S. Nuclear Regulatory Commission's (NRC's) Office of Nuclear Reactor Regulation (NRR) and the Office for Analysis and Evaluation of Operational Data (AEOD). This paper presents an overview of the models and software. Key characteristics include: (1) classification of the plant models according to plant response with a unique set of event trees for each plant class, (2) plant-specific fault trees using supercomponents, (3) generation and retention of all system and sequence cutsets, (4) full flexibility in modifying logic, regenerating cutsets, and requantifying results, and (5) user interface for streamlined evaluation of ASP events. Future plans for the ASP models is also presented

Assimilation of SAPHIR radiance: impact on hyperspectral radiances in 4D-VAR

Science.gov (United States)

Indira Rani, S.; Doherty, Amy; Atkinson, Nigel; Bell, William; Newman, Stuart; Renshaw, Richard; George, John P.; Rajagopal, E. N.

2016-04-01

Assimilation of a new observation dataset in an NWP system may affect the quality of an existing observation data set against the model background (short forecast), which in-turn influence the use of an existing observation in the NWP system. Effect of the use of one data set on the use of another data set can be quantified as positive, negative or neutral. Impact of the addition of new dataset is defined as positive if the number of assimilated observations of an existing type of observation increases, and bias and standard deviation decreases compared to the control (without the new dataset) experiment. Recently a new dataset, Megha Tropiques SAPHIR radiances, which provides atmospheric humidity information, is added in the Unified Model 4D-VAR assimilation system. In this paper we discuss the impact of SAPHIR on the assimilation of hyper-spectral radiances like AIRS, IASI and CrIS. Though SAPHIR is a Microwave instrument, its impact can be clearly seen in the use of hyper-spectral radiances in the 4D-VAR data assimilation systems in addition to other Microwave and InfraRed observation. SAPHIR assimilation decreased the standard deviation of the spectral channels of wave number from 650 -1600 cm-1 in all the three hyperspectral radiances. Similar impact on the hyperspectral radiances can be seen due to the assimilation of other Microwave radiances like from AMSR2 and SSMIS Imager.

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)

Characterisation of the photolytic HONO-source in the atmosphere simulation chamber SAPHIR

Directory of Open Access Journals (Sweden)

F. Rohrer

2005-01-01

Full Text Available HONO formation has been proposed as an important OH radical source in simulation chambers for more than two decades. Besides the heterogeneous HONO formation by the dark reaction of NO2 and adsorbed water, a photolytic source has been proposed to explain the elevated reactivity in simulation chamber experiments. However, the mechanism of the photolytic process is not well understood so far. As expected, production of HONO and NOx was also observed inside the new atmospheric simulation chamber SAPHIR under solar irradiation. This photolytic HONO and NOx formation was studied with a sensitive HONO instrument under reproducible controlled conditions at atmospheric concentrations of other trace gases. It is shown that the photolytic HONO source in the SAPHIR chamber is not caused by NO2 reactions and that it is the only direct NOy source under illuminated conditions. In addition, the photolysis of nitrate which was recently postulated for the observed photolytic HONO formation on snow, ground, and glass surfaces, can be excluded in the chamber. A photolytic HONO source at the surface of the chamber is proposed which is strongly dependent on humidity, on light intensity, and on temperature. An empirical function describes these dependencies and reproduces the observed HONO formation rates to within 10%. It is shown that the photolysis of HONO represents the dominant radical source in the SAPHIR chamber for typical tropospheric O3/H2O concentrations. For these conditions, the HONO concentrations inside SAPHIR are similar to recent observations in ambient air.

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)

Review of the Brunswick Steam Electric Plant Probabilistic Risk Assessment

International Nuclear Information System (INIS)

Sattison, M.B.; Davis, P.R.; Satterwhite, D.G.; Gilmore, W.E.; Gregg, R.E.

1989-11-01

A review of the Brunswick Steam Electric Plant probabilistic risk Assessment was conducted with the objective of confirming the safety perspectives brought to light by the probabilistic risk assessment. The scope of the review included the entire Level I probabilistic risk assessment including external events. This is consistent with the scope of the probabilistic risk assessment. The review included an assessment of the assumptions, methods, models, and data used in the study. 47 refs., 14 figs., 15 tabs

Design and construction of the SAPHIR detector

International Nuclear Information System (INIS)

Schwille, W.J.; Bockhorst, M.; Burbach, G.; Burgwinkel, R.; Empt, J.; Guse, B.; Haas, K.M.; Hannappel, J.; Heinloth, K.; Hey, T.; Honscheid, K.; Jahnen, T.; Jakob, H.P.; Joepen, N.; Juengst, H.; Kirch, U.; Klein, F.J.; Kostrewa, D.; Lindemann, L.; Link, J.; Manns, J.; Menze, D.; Merkel, H.; Merkel, R.; Neuerburg, W.; Paul, E.; Ploetzke, R.; Schenk, U.; Schmidt, S.; Scholmann, J.; Schuetz, P.; Schultz-Coulon, H.C.; Schweitzer, M.; Tran, M.Q.; Vogl, W.; Wedemeyer, R.; Wehnes, F.; Wisskirchen, J.; Wolf, A.

1994-01-01

The design, construction, and performance of the large solid angle magnetic spectrometer SAPHIR is described. It was built for the investigation of photon-induced reactions on nucleous and light nuclei with mulit-particle final states up to photon energies of 3.1 GeV. The detector is equipped with a tagged photon beam facility and is operated at the stretcher ring ELSA in Bonn. (orig.)

Design and construction of the SAPHIR detector

Energy Technology Data Exchange (ETDEWEB)

Schwille, W.J. (Bonn Univ. (Germany). Physikalisches Inst.); Bockhorst, M. (Bonn Univ. (Germany). Physikalisches Inst.); Burbach, G. (Bonn Univ. (Germany). Physikalisches Inst.); Burgwinkel, R. (Bonn Univ. (Germany). Physikalisches Inst.); Empt, J. (Bonn Univ. (Germany). Physikalisches Inst.); Guse, B. (Bonn Univ. (Germany). Physikalisches Inst.); Haas, K.M. (Bonn Univ. (Germany). Physikalisches Inst.); Hannappel, J. (Bonn Univ. (Germany). Physikalisches Inst.); Heinloth, K. (Bonn Univ. (Germany). Physikalisches Inst.); Hey, T. (Bonn Univ. (Germany). Physikalisches Inst.); Honscheid, K. (Bonn Univ. (Germany). Physikalisches Inst.); Jahnen, T. (Bonn Univ. (Germany). Physikalisches Inst.); Jakob, H.P. (Bonn Univ. (Germany). Physikalisches Inst.); Joepen, N. (Bonn Univ. (Germany). Physikalisches Inst.); Juengst, H. (Bonn Univ. (Germany). Physikalisches Inst.); Kirch, U. (Bonn Univ. (Germany). Physikalisches Inst.); Klein, F.J. (Bonn Univ. (Germany). Physikalisches Inst.)

1994-05-15

The design, construction, and performance of the large solid angle magnetic spectrometer SAPHIR is described. It was built for the investigation of photon-induced reactions on nucleous and light nuclei with mulit-particle final states up to photon energies of 3.1 GeV. The detector is equipped with a tagged photon beam facility and is operated at the stretcher ring ELSA in Bonn. (orig.)

Probabilistic assessment of pressure vessel and piping reliability

International Nuclear Information System (INIS)

Sundararajan, C.

1986-01-01

The paper presents a critical review of the state-of-the-art in probabilistic assessment of pressure vessel and piping reliability. First the differences in assessing the reliability directly from historical failure data and indirectly by a probabilistic analysis of the failure phenomenon are discussed and the advantages and disadvantages are pointed out. The rest of the paper deals with the latter approach of reliability assessment. Methods of probabilistic reliability assessment are described and major projects where these methods are applied for pressure vessel and piping problems are discussed. An extensive list of references is provided at the end of the paper

Development of a model-independent evaluation of photon-deuteron reactions for the SAPHIR detector

International Nuclear Information System (INIS)

Wolf, A.

1993-01-01

The SAPHIR detector measures photon induced reactions with many particles in the final state. Thus a detailed investigation of those processes at photon energies between 0.4 and 3.3 GeV is possible. The interpretation of the distribution of the sample of events, which SAPHIR is able to reconstruct, has to be done after a correction of influences induced by the detector acceptance. In this work a model independent method of correcting and analysing the data is discussed. The implementation of the basic tools of this analysis is described and first tests with simulated and real events are performed. SAPHIR uses a time-of-flight system for the identification of particles. This work describes the structure of a program library, which supports an easy way of decoding the digitizations of this system (including calibration of the hardware) and obtaining the flight time for a particle in a event. The necessary step for calibrating the system are outlined, too. (orig.)

Development of the software of the data taking system SOS for the SAPHIR experiment

International Nuclear Information System (INIS)

Manns, J.

1989-02-01

The data acquistion system SOS has been developed for the SAPHIR experiment at the Bonn stretcher ring ELSA. It can handle up to 280 kilobytes of data per second or a maximum triggerrate of 200 Hz. The multiprocessor based online system consists of twenty VIP-microprocessors and two VAX-computers. Each component of the SAPHIR experiment has at least one program in the online system to maintain special functions for this specific component. All of these programs can receive event data without interfering with the transfer of events to a mass storage for offline analysis. A special program SOL has been developed to serve as a user interface to the data acquisition system and as a status display for most of the programs of the online system. Using modern features like windowing and mouse control on a VAX-station the SAPHIR online SOL establishes an easy way of controlling the data acquisition system. (orig.)

Analysis of truncation limit in probabilistic safety assessment

International Nuclear Information System (INIS)

Cepin, Marko

2005-01-01

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

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)

Independent Verification and Validation SAPHIRE Version 8 Final Report Project Number: N6423 U.S. Nuclear Regulatory Commission

Energy Technology Data Exchange (ETDEWEB)

Kent Norris

2010-04-01

This report provides an evaluation of the SAPHIRE version 8 software product. SAPHIRE version 8 is being developed with a phased or cyclic iterative rapid application development methodology. Due to this approach, a similar approach has been taken for the IV&V activities on each vital software object. IV&V and Software Quality Assurance (SQA) activities occur throughout the entire development life cycle and therefore, will be required through the full development of SAPHIRE version 8. Later phases of the software life cycle, the operation and maintenance phases, are not applicable in this effort since the IV&V is being done prior to releasing Version 8.

Documentation design for probabilistic risk assessment

International Nuclear Information System (INIS)

Parkinson, W.J.; von Herrmann, J.L.

1985-01-01

This paper describes a framework for documentation design of probabilistic risk assessment (PRA) and is based on the EPRI document NP-3470 ''Documentation Design for Probabilistic Risk Assessment''. The goals for PRA documentation are stated. Four audiences are identified which PRA documentation must satisfy, and the documentation consistent with the needs of the various audiences are discussed, i.e., the Summary Report, the Executive Summary, the Main Report, and Appendices. The authors recommend the documentation specifications discussed herein as guides rather than rigid definitions

The photon detection system of the SAPHIR spectrometer

International Nuclear Information System (INIS)

Joepen, N.

1990-09-01

Worldwide a new generation of Electron Accelerators with energies below 5 GeV and a high duty cycle up to 100% is being built or planned. The first machine of this kind is ELSA, the Electron Stretcher and Accelerator, at the Physics Institute of Bonn University. Due to the high duty cycle of ELSA, experiments with tagged photon beams and a large angular acceptance become possible. At present SAPHIR, a new magnetic detector, especially layed out to detect multi-particle final states with good accuracy, is going into operation. Besides a large arrangement of drift chambers, for a good momentum resolution, and a trigger- and time-of-flight counter system, for particle identification, one of the main features of SAPHIR is a good photon detection capability. This is accomplished by a large electromagnetic calorimeter consisting of 98 modules covering a detection area of about 16 m 2 in forward direction. For the calorimeter a brass-gas-sandwich detector was developed. Its signal wires are strung perpendicular to the converter planes. The chambers are filled with a standard gas mixture Ar/CH 4 (90:10) at atmospheric pressure and operated with a considerably high voltage in the semi-proportional mode. A sample of nine shower counter modules was tested at the electron test beam of the Bonn 2.5 GeV electron synchrotron. An energy resolution of σ(E)/(E*√E(GeV)) = 13.55 ± 0.6% for a single module was achieved. The incident angle of the electrons was varied between 0 and 45 degrees. No significant change of energy resolution and linearity was observed. Combining the information from wire and cathode signals a position resolution (E = 1 GeV:Φ=0deg → σ = 15 mm, Φ=45deg → σ x = 19 mm) was reached. The second part of this paper gives a description of the shower counter arrangement in the SAPHIR detector. It requires a sophisticated control and calibration system, whose details are presented. Further on some aspects of the calorimeter calibration procedure are discussed

PRECIS -- A probabilistic risk assessment system

International Nuclear Information System (INIS)

Peterson, D.M.; Knowlton, R.G. Jr.

1996-01-01

A series of computer tools has been developed to conduct the exposure assessment and risk characterization phases of human health risk assessments within a probabilistic framework. The tools are collectively referred to as the Probabilistic Risk Evaluation and Characterization Investigation System (PRECIS). With this system, a risk assessor can calculate the doses and risks associated with multiple environmental and exposure pathways, for both chemicals and radioactive contaminants. Exposure assessment models in the system account for transport of contaminants to receptor points from a source zone originating in unsaturated soils above the water table. In addition to performing calculations of dose and risk based on initial concentrations, PRECIS can also be used in an inverse manner to compute soil concentrations in the source area that must not be exceeded if prescribed limits on dose or risk are to be met. Such soil contaminant levels, referred to as soil guidelines, are computed for both single contaminants and chemical mixtures and can be used as action levels or cleanup levels. Probabilistic estimates of risk, dose and soil guidelines are derived using Monte Carlo techniques

Implementation of the FASTBUS data-acquisition system in the readout of the SAPHIR detector

International Nuclear Information System (INIS)

Empt, J.

1993-12-01

The magnetic detector SAPHIR is layed out to detect multiparticle final states with good acuracy, especially a good photon detection capability is designed. Therefore a large electromagnetic calorimeter was built, consisting of 98 modules covering a detection area of about 16 m 2 in forward direction. For this calorimeter a brass-gas-sandwich detector was developed with signal wires perpendicular to the converter planes. For data acquisition of a major part of this calorimeter a modular FASTBUS system is used. In this report the FASTBUS system and its installation in the SAPHIR Online Program are described. (orig.)

Non-probabilistic defect assessment for structures with cracks based on interval model

International Nuclear Information System (INIS)

Dai, Qiao; Zhou, Changyu; Peng, Jian; Chen, Xiangwei; He, Xiaohua

2013-01-01

Highlights: • Non-probabilistic approach is introduced to defect assessment. • Definition and establishment of IFAC are put forward. • Determination of assessment rectangle is proposed. • Solution of non-probabilistic reliability index is presented. -- Abstract: Traditional defect assessment methods conservatively treat uncertainty of parameters as safety factors, while the probabilistic method is based on the clear understanding of detailed statistical information of parameters. In this paper, the non-probabilistic approach is introduced to the failure assessment diagram (FAD) to propose a non-probabilistic defect assessment method for structures with cracks. This novel defect assessment method contains three critical processes: establishment of the interval failure assessment curve (IFAC), determination of the assessment rectangle, and solution of the non-probabilistic reliability degree. Based on the interval theory, uncertain parameters such as crack sizes, material properties and loads are considered as interval variables. As a result, the failure assessment curve (FAC) will vary in a certain range, which is defined as IFAC. And the assessment point will vary within a rectangle zone which is defined as an assessment rectangle. Based on the interval model, the establishment of IFAC and the determination of the assessment rectangle are presented. Then according to the interval possibility degree method, the non-probabilistic reliability degree of IFAC can be determined. Meanwhile, in order to clearly introduce the non-probabilistic defect assessment method, a numerical example for the assessment of a pipe with crack is given. In addition, the assessment result of the proposed method is compared with that of the traditional probabilistic method, which confirms that this non-probabilistic defect assessment can reasonably resolve the practical problem with interval variables

Non-probabilistic defect assessment for structures with cracks based on interval model

Energy Technology Data Exchange (ETDEWEB)

Dai, Qiao; Zhou, Changyu, E-mail: changyu_zhou@163.com; Peng, Jian; Chen, Xiangwei; He, Xiaohua

2013-09-15

Highlights: • Non-probabilistic approach is introduced to defect assessment. • Definition and establishment of IFAC are put forward. • Determination of assessment rectangle is proposed. • Solution of non-probabilistic reliability index is presented. -- Abstract: Traditional defect assessment methods conservatively treat uncertainty of parameters as safety factors, while the probabilistic method is based on the clear understanding of detailed statistical information of parameters. In this paper, the non-probabilistic approach is introduced to the failure assessment diagram (FAD) to propose a non-probabilistic defect assessment method for structures with cracks. This novel defect assessment method contains three critical processes: establishment of the interval failure assessment curve (IFAC), determination of the assessment rectangle, and solution of the non-probabilistic reliability degree. Based on the interval theory, uncertain parameters such as crack sizes, material properties and loads are considered as interval variables. As a result, the failure assessment curve (FAC) will vary in a certain range, which is defined as IFAC. And the assessment point will vary within a rectangle zone which is defined as an assessment rectangle. Based on the interval model, the establishment of IFAC and the determination of the assessment rectangle are presented. Then according to the interval possibility degree method, the non-probabilistic reliability degree of IFAC can be determined. Meanwhile, in order to clearly introduce the non-probabilistic defect assessment method, a numerical example for the assessment of a pipe with crack is given. In addition, the assessment result of the proposed method is compared with that of the traditional probabilistic method, which confirms that this non-probabilistic defect assessment can reasonably resolve the practical problem with interval variables.

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)

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)

Probabilistic risk assessment in nuclear power plant regulation

Energy Technology Data Exchange (ETDEWEB)

Wall, J B

1980-09-01

A specific program is recommended to utilize more effectively probabilistic risk assessment in nuclear power plant regulation. It is based upon the engineering insights from the Reactor Safety Study (WASH-1400) and some follow-on risk assessment research by USNRC. The Three Mile Island accident is briefly discussed from a risk viewpoint to illustrate a weakness in current practice. The development of a probabilistic safety goal is recommended with some suggestions on underlying principles. Some ongoing work on risk perception and the draft probabilistic safety goal being reviewed on Canada is described. Some suggestions are offered on further risk assessment research. Finally, some recent U.S. Nuclear Regulatory Commission actions are described.

Probabilistic Flood Defence Assessment Tools

Directory of Open Access Journals (Sweden)

Slomp Robert

2016-01-01

institutions managing flood the defences, and not by just a small number of experts in probabilistic assessment. Therefore, data management and use of software are main issues that have been covered in courses and training in 2016 and 2017. All in all, this is the largest change in the assessment of Dutch flood defences since 1996. In 1996 probabilistic techniques were first introduced to determine hydraulic boundary conditions (water levels and waves (wave height, wave period and direction for different return periods. To simplify the process, the assessment continues to consist of a three-step approach, moving from simple decision rules, to the methods for semi-probabilistic assessment, and finally to a fully probabilistic analysis to compare the strength of flood defences with the hydraulic loads. The formal assessment results are thus mainly based on the fully probabilistic analysis and the ultimate limit state of the strength of a flood defence. For complex flood defences, additional models and software were developed. The current Hydra software suite (for policy analysis, formal flood defence assessment and design will be replaced by the model Ringtoets. New stand-alone software has been developed for revetments, geotechnical analysis and slope stability of the foreshore. Design software and policy analysis software, including the Delta model, will be updated in 2018. A fully probabilistic method results in more precise assessments and more transparency in the process of assessment and reconstruction of flood defences. This is of increasing importance, as large-scale infrastructural projects in a highly urbanized environment are increasingly subject to political and societal pressure to add additional features. For this reason, it is of increasing importance to be able to determine which new feature really adds to flood protection, to quantify how much its adds to the level of flood protection and to evaluate if it is really worthwhile. Please note: The Netherlands

Probabilistic tsunami hazard assessment for Point Lepreau Generating Station

Energy Technology Data Exchange (ETDEWEB)

Mullin, D., E-mail: dmullin@nbpower.com [New Brunswick Power Corporation, Point Lepreau Generating Station, Point Lepreau (Canada); Alcinov, T.; Roussel, P.; Lavine, A.; Arcos, M.E.M.; Hanson, K.; Youngs, R., E-mail: trajce.alcinov@amecfw.com, E-mail: patrick.roussel@amecfw.com [AMEC Foster Wheeler Environment & Infrastructure, Dartmouth, NS (Canada)

2015-07-01

In 2012 the Geological Survey of Canada published a preliminary probabilistic tsunami hazard assessment in Open File 7201 that presents the most up-to-date information on all potential tsunami sources in a probabilistic framework on a national level, thus providing the underlying basis for conducting site-specific tsunami hazard assessments. However, the assessment identified a poorly constrained hazard for the Atlantic Coastline and recommended further evaluation. As a result, NB Power has embarked on performing a Probabilistic Tsunami Hazard Assessment (PTHA) for Point Lepreau Generating Station. This paper provides the methodology and progress or hazard evaluation results for Point Lepreau G.S. (author)

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

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

SAPHIR, a simulator for engineering and training on N4-type nuclear power plants

International Nuclear Information System (INIS)

Vovan, C.

1999-01-01

SAPHIR, the new simulator developed by FRAMATOME, has been designed to be a convenient tool for engineering and training for different types of nuclear power plants. Its first application is for the French 'N4' four-loop 1500MWe PWR. The basic features of SAPHIR are: (1) Use of advanced codes for modelling He primary and secondary systems' including an axial steam generator model, (2) Use of a simulation workshop containing different tools for modelling fluid, electrical, instrument and control networks, (3) A Man-Machine Interface designed for an easy and convivial use which can simulate the different computerized control consoles of the 'N4' control room. This paper outlines features and capabilities of this tool, both for engineering and training purposes. (author)

The alarm system of the SAPHIR detector

International Nuclear Information System (INIS)

Schultz-Coulon, H.C.

1993-06-01

In order to obtain an effective control of the different detector components an alarm system was built and implemented into the data acquisition system of the SAPHIR experiment. It provides an easy way of indicating errors by either adequate library calls or an appropriate hardware signal, both leading to an active alarm. This allows to react directly to any error detected by one of the specific control systems. In addition for selected kinds of errors the data run can be stopped automatically. Concept and construction of this system are described and some examples for its application are given. (orig.)

Advanced Test Reactor probabilistic risk assessment

International Nuclear Information System (INIS)

Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

1993-01-01

This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

Construction and calibration studies of the SAPHIR scintillation counters

International Nuclear Information System (INIS)

Kostrewa, D.

1988-03-01

For the scintillation counter system of the SAPHIR detector at the stretcher ring ELSA in Bonn 50 time of flight counters and 12 trigger counters have been built. Each of them has two photomultipliers, one at each side. A laser calibration system with a pulsed nitrogen laser as central light source to monitor these photomultipliers has been optimized. It was used to adjust the photomultipliers and to test their long and short time instabilities. (orig.)

Probabilistic risk assessment as an aid to risk management

International Nuclear Information System (INIS)

Garrick, B.J.

1982-01-01

Probabilistic risk assessments are providing important insights into nuclear power plant safety. Their value is two-fold: first as a means of quantifying nuclear plant risk including contributors to risk, and second as an aid to risk management. A risk assessment provides an analytical plant model that can be the basis for performing meaningful decision analyses for controlling safety. It is the aspect of quantitative risk management that makes probabilistic risk assessment an important technical discipline of the future

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

Track recognition in the central drift chamber of the SAPHIR detector at ELSA and first reconstruction of real tracks

International Nuclear Information System (INIS)

Korn, P.

1991-02-01

The FORTRAN program for pattern recognition in the central drift chamber of SAPHIR has been modified in order to find tracks with more than one missing wire signal and has been optimized in resolving the left/right ambiguities. The second part of this report deals with the reconstruction of some real tracks (γ → e + e - ), which were measured with SAPHIR. The efficiency of the central drift chamber and the space-to-drift time-relation are discussed. (orig.)

The international probabilistic system assessment group. Background and results 1990

International Nuclear Information System (INIS)

1991-01-01

The OECD Nuclear Energy Agency (NEA) devotes considerable effort to the further development of methodologies to assess the performance of radioactive waste disposal systems, and to increase confidence in their application and results. The NEA provides an international forum for the exchange of information and experience among national experts of its twenty-three Member countries and conducts joint studies of issues important for safety assessment. In 1985, the NEA Radioactive Waste Management Committee set up the Probabilistic System Assessment Code User Group (PSAC), in order to help coordinate the development of probabilistic system assessment codes. The activities of the Group include exchange of information, code and experience, discussion of relevant technical issues, and the conduct of code comparison (PSACOIN) exercises designed to build confidence in the correct operation of these tools for safety assessment. The Group is now known simply as the Probabilistic System Assessment Group (PSAG). This report has been prepared to inform interested parties, beyond the group of specialists directly involved, about probabilistic system assessment techniques as used for performance assessment of waste disposal systems, and to give a summary of the objectives and achievements of PSAG. The report is published under the responsibility of the Secretary General of the OECD

Probabilistic risk assessment, Volume I

International Nuclear Information System (INIS)

Anon.

1982-01-01

This book contains 158 papers presented at the International Topical Meeting on Probabilistic Risk Assessment held by the American Nuclear Society (ANS) and the European Nuclear Society (ENS) in Port Chester, New York in 1981. The meeting was second in a series of three. The main focus of the meeting was on the safety of light water reactors. The papers discuss safety goals and risk assessment. Quantitative safety goals, risk assessment in non-nuclear technologies, and operational experience and data base are also covered. Included is an address by Dr. Chauncey Starr

Human reliability assessment and probabilistic risk assessment

International Nuclear Information System (INIS)

Embrey, D.E.; Lucas, D.A.

1989-01-01

Human reliability assessment (HRA) is used within Probabilistic Risk Assessment (PRA) to identify the human errors (both omission and commission) which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. There exist a variey of HRA techniques and the selection of an appropriate one is often difficult. This paper reviews a number of available HRA techniques and discusses their strengths and weaknesses. The techniques reviewed include: decompositional methods, time-reliability curves and systematic expert judgement techniques. (orig.)

A review of probabilistic risk assessment of contaminated land

International Nuclear Information System (INIS)

Oeberg, T.; Bergbaeck, B.

2005-01-01

Background, Aims and Scope. The management and decisions concerning restoration of contaminated land often require indepth risk analyses. An environmental risk assessment is generally described as proceeding in four separate steps: hazard identification, dose-response assessment, exposure assessment, and risk characterization. The risk assessment should acknowledge and quantify the uncertainty in risk predictions. This can be achieved by applying probabilistic methods which, although they have been available for many years, are still not generally used. Risk assessment of contaminated land is an area where probabilistic methods have proved particularly useful. Many reports have appeared in the literature, mostly by North American researchers. The aim of this review is to summarize the experience gained so far, provide a number of useful examples, and suggest what may be done to promote probabilistic methods in Europe and the rest of the world. Methods. The available literature has been explored through searches in the major scientific and technical databases, WWW resources, textbooks and direct contacts with active researchers. A calculation example was created using standard simulation software. (orig.)

Probabilistic safety assessment

International Nuclear Information System (INIS)

Hoertner, H.; Schuetz, B.

1982-09-01

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

Limited probabilistic risk assessment applications in plant backfitting

International Nuclear Information System (INIS)

Desaedeleer, G.

1987-01-01

Plant backfitting programs are defined on the basis of deterministic (e.g. Systematic Evaluation Program) or probabilistic (e.g. Probabilistic Risk Assessment) approaches. Each approach provides valuable assets in defining the program and has its own advantages and disadvantages. Ideally one should combine the strong points of each approach. This chapter summarizes actual experience gained from combinations of deterministic and probabilistic approaches to define and implement PWR backfitting programs. Such combinations relate to limited applications of probabilistic techniques and are illustrated for upgrading fluid systems. These evaluations allow sound and rational optimization systems upgrade. However, the boundaries of the reliability analysis need to be clearly defined and system reliability may have to go beyond classical boundaries (e.g. identification of weak links in support systems). Also the implementation of upgrade on a system per system basis is not necessarily cost-effective. (author)

Uncertainty and sensitivity analysis using probabilistic system assessment code. 1

International Nuclear Information System (INIS)

Honma, Toshimitsu; Sasahara, Takashi.

1993-10-01

This report presents the results obtained when applying the probabilistic system assessment code under development to the PSACOIN Level 0 intercomparison exercise organized by the Probabilistic System Assessment Code User Group in the Nuclear Energy Agency (NEA) of OECD. This exercise is one of a series designed to compare and verify probabilistic codes in the performance assessment of geological radioactive waste disposal facilities. The computations were performed using the Monte Carlo sampling code PREP and post-processor code USAMO. The submodels in the waste disposal system were described and coded with the specification of the exercise. Besides the results required for the exercise, further additional uncertainty and sensitivity analyses were performed and the details of these are also included. (author)

Probabilistic risk assessment (PRA) reference document. Final report

International Nuclear Information System (INIS)

Murphy, J.A.

1984-09-01

This document describes the current status of probabilistic risk assessment (PRA) as practiced in the nuclear reactor regulatory process. The PRA studies that have been completed or are under way are reviewed. The levels of maturity of the methodologies used in a PRA are discussed. Insights derived from PRAs are listed. The potential uses of PRA results for regulatory purposes are discussed. This document was issued for comment in February 1984 entitled Probabilistic Risk Assessment (PRA): Status Report and Guidance for Regulatory Application. The comments received on the draft have been considered for this final version of the report

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

A methodology for reviewing probabilistic risk assessments

International Nuclear Information System (INIS)

Derby, S.L.

1983-01-01

The starting point for peer review of a Probabilistic Risk Assessment (PRA) is a clear understanding of how the risk estimate was prepared and of what contributions dominate the calculation. The problem facing the reviewers is how to cut through the complex details of a PRA to gain this understanding. This paper presents a structured, analytical procedure that solves this problem. The effectiveness of this solution is demonstrated by an application on the Zion Probabilistic Safety Study. The procedure found the three dominant initiating events and provided a simplified reconstruction of the calculation of the risk estimate. Significant assessments of uncertainty were also identified. If peer review disputes the accuracy of these judgments, then the revised risk estimate could significantly increase

Development of a FASTBUS data acquisition system for the SAPHIR calorimeter

International Nuclear Information System (INIS)

Klein, F.J.

1992-01-01

Due to the high duty cycle of the new Electron Accelerator at the Physics Institute of Bonn University, ELSA, experiments with tagged photon beams and a large angular acceptance become possible. The new magnetic detector SAPHIR is layed out to detect multi-particle final states with good accuracy, especially a good photon detection capability is designed. Therefore a large electromagnetic calorimeter is built, consisting of 98 modules covering a detection area of about 16 m 2 in forward direction. For this calorimeter a brass-gas-sandwich detector was developed with signal wires perpendicular to the converter planes. The chambers are filled with a standard gas mixture Ar/CH 4 (90:10) at atmospheric pressure and operated with a considerably high voltage in the semi-proportional mode. A modified shower counter module, containing 20 μm thick signal wires, was tested at the electron test beam of the Bonn 2.5 GeV electron synchrotron. An energy resolution of σ(E)/E*√E(GeV) = 12.2±0.5% was achieved. For data acquisition a modular FASTBUS system was used, which will be installed in the SAPHIR Online Program. (orig.) [de

The graphics system and the data saving for the SAPHIR experiment

International Nuclear Information System (INIS)

Albold, D.

1990-08-01

Important extensions have been made to the data acquisition system SOS for the SAPHIR experiment at the Bonn ELSA facilities. As support for various online-programs, controlling components of the detector, a graphic system for presenting data was developed. This enables any program in the system to use all graphic devices. Main component is a program serving requests for presentation on a 19 inch color monitor. Window-technique allows a presentation of several graphics on one screen. Equipped with a trackball and using menus, this is an easy to use and powerful tool in controlling the experiment. Other important extensions concern data storage. A huge amount of event data can be stored on 8 mm cassettes by the program Eventsaver. This program can be controlled by a component of the SAPHIR-Online SOL running on a VAX-Computer and using windows and menus. The smaller amount of data, containing parameters and programs, which should be accessible within a small period of time, can be stored on a magnetic disk. A program supporting a file-structure for access to this disk is described. (orig./HSI) [de

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)

Probabilistic safety assessment goals in Canada

International Nuclear Information System (INIS)

Snell, V.G.

1986-01-01

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

A probabilistic tsunami hazard assessment for Indonesia

Science.gov (United States)

Horspool, N.; Pranantyo, I.; Griffin, J.; Latief, H.; Natawidjaja, D. H.; Kongko, W.; Cipta, A.; Bustaman, B.; Anugrah, S. D.; Thio, H. K.

2014-11-01

Probabilistic hazard assessments are a fundamental tool for assessing the threats posed by hazards to communities and are important for underpinning evidence-based decision-making regarding risk mitigation activities. Indonesia has been the focus of intense tsunami risk mitigation efforts following the 2004 Indian Ocean tsunami, but this has been largely concentrated on the Sunda Arc with little attention to other tsunami prone areas of the country such as eastern Indonesia. We present the first nationally consistent probabilistic tsunami hazard assessment (PTHA) for Indonesia. This assessment produces time-independent forecasts of tsunami hazards at the coast using data from tsunami generated by local, regional and distant earthquake sources. The methodology is based on the established monte carlo approach to probabilistic seismic hazard assessment (PSHA) and has been adapted to tsunami. We account for sources of epistemic and aleatory uncertainty in the analysis through the use of logic trees and sampling probability density functions. For short return periods (100 years) the highest tsunami hazard is the west coast of Sumatra, south coast of Java and the north coast of Papua. For longer return periods (500-2500 years), the tsunami hazard is highest along the Sunda Arc, reflecting the larger maximum magnitudes. The annual probability of experiencing a tsunami with a height of > 0.5 m at the coast is greater than 10% for Sumatra, Java, the Sunda islands (Bali, Lombok, Flores, Sumba) and north Papua. The annual probability of experiencing a tsunami with a height of > 3.0 m, which would cause significant inundation and fatalities, is 1-10% in Sumatra, Java, Bali, Lombok and north Papua, and 0.1-1% for north Sulawesi, Seram and Flores. The results of this national-scale hazard assessment provide evidence for disaster managers to prioritise regions for risk mitigation activities and/or more detailed hazard or risk assessment.

A~probabilistic tsunami hazard assessment for Indonesia

Science.gov (United States)

Horspool, N.; Pranantyo, I.; Griffin, J.; Latief, H.; Natawidjaja, D. H.; Kongko, W.; Cipta, A.; Bustaman, B.; Anugrah, S. D.; Thio, H. K.

2014-05-01

Probabilistic hazard assessments are a fundamental tool for assessing the threats posed by hazards to communities and are important for underpinning evidence based decision making on risk mitigation activities. Indonesia has been the focus of intense tsunami risk mitigation efforts following the 2004 Indian Ocean Tsunami, but this has been largely concentrated on the Sunda Arc, with little attention to other tsunami prone areas of the country such as eastern Indonesia. We present the first nationally consistent Probabilistic Tsunami Hazard Assessment (PTHA) for Indonesia. This assessment produces time independent forecasts of tsunami hazard at the coast from tsunami generated by local, regional and distant earthquake sources. The methodology is based on the established monte-carlo approach to probabilistic seismic hazard assessment (PSHA) and has been adapted to tsunami. We account for sources of epistemic and aleatory uncertainty in the analysis through the use of logic trees and through sampling probability density functions. For short return periods (100 years) the highest tsunami hazard is the west coast of Sumatra, south coast of Java and the north coast of Papua. For longer return periods (500-2500 years), the tsunami hazard is highest along the Sunda Arc, reflecting larger maximum magnitudes along the Sunda Arc. The annual probability of experiencing a tsunami with a height at the coast of > 0.5 m is greater than 10% for Sumatra, Java, the Sunda Islands (Bali, Lombok, Flores, Sumba) and north Papua. The annual probability of experiencing a tsunami with a height of >3.0 m, which would cause significant inundation and fatalities, is 1-10% in Sumatra, Java, Bali, Lombok and north Papua, and 0.1-1% for north Sulawesi, Seram and Flores. The results of this national scale hazard assessment provide evidence for disaster managers to prioritise regions for risk mitigation activities and/or more detailed hazard or risk assessment.

TOPAS 2 - a high-resolution tagging system at the Bonn SAPHIR detector

International Nuclear Information System (INIS)

Rappenecker, G.

1989-02-01

For the SAPHIR-arrangement in Bonn a high resolving tagging system has been developed achieving an energy resolution of 2 MeV, covering the range of (0.94-0.34) E 0 photon energy (1.0 GeV 0 2 , ArCH 4 and ArC 2 H 6 in concern of performance, clustersize and coincidence width. (orig.)

A new plant chamber facility PLUS coupled to the atmospheric simulation chamber SAPHIR

Science.gov (United States)

Hohaus, T.; Kuhn, U.; Andres, S.; Kaminski, M.; Rohrer, F.; Tillmann, R.; Wahner, A.; Wegener, R.; Yu, Z.; Kiendler-Scharr, A.

2015-11-01

A new PLant chamber Unit for Simulation (PLUS) for use with the atmosphere simulation chamber SAPHIR (Simulation of Atmospheric PHotochemistry In a large Reaction Chamber) has been build and characterized at the Forschungszentrum Jülich GmbH, Germany. The PLUS chamber is an environmentally controlled flow through plant chamber. Inside PLUS the natural blend of biogenic emissions of trees are mixed with synthetic air and are transferred to the SAPHIR chamber where the atmospheric chemistry and the impact of biogenic volatile organic compounds (BVOC) can be studied in detail. In PLUS all important enviromental parameters (e.g. temperature, PAR, soil RH etc.) are well-controlled. The gas exchange volume of 9.32 m3 which encloses the stem and the leafes of the plants is constructed such that gases are exposed to FEP Teflon film and other Teflon surfaces only to minimize any potential losses of BVOCs in the chamber. Solar radiation is simulated using 15 LED panels which have an emission strength up to 800 μmol m-2 s-1. Results of the initial characterization experiments are presented in detail. Background concentrations, mixing inside the gas exchange volume, and transfer rate of volatile organic compounds (VOC) through PLUS under different humidity conditions are explored. Typical plant characteristics such as light and temperature dependent BVOC emissions are studied using six Quercus Ilex trees and compared to previous studies. Results of an initial ozonolysis experiment of BVOC emissions from Quercus Ilex at typical atmospheric concentrations inside SAPHIR are presented to demonstrate a typical experimental set up and the utility of the newly added plant chamber.

A new plant chamber facility, PLUS, coupled to the atmosphere simulation chamber SAPHIR

Science.gov (United States)

Hohaus, T.; Kuhn, U.; Andres, S.; Kaminski, M.; Rohrer, F.; Tillmann, R.; Wahner, A.; Wegener, R.; Yu, Z.; Kiendler-Scharr, A.

2016-03-01

A new PLant chamber Unit for Simulation (PLUS) for use with the atmosphere simulation chamber SAPHIR (Simulation of Atmospheric PHotochemistry In a large Reaction Chamber) has been built and characterized at the Forschungszentrum Jülich GmbH, Germany. The PLUS chamber is an environmentally controlled flow-through plant chamber. Inside PLUS the natural blend of biogenic emissions of trees is mixed with synthetic air and transferred to the SAPHIR chamber, where the atmospheric chemistry and the impact of biogenic volatile organic compounds (BVOCs) can be studied in detail. In PLUS all important environmental parameters (e.g., temperature, photosynthetically active radiation (PAR), soil relative humidity (RH)) are well controlled. The gas exchange volume of 9.32 m3 which encloses the stem and the leaves of the plants is constructed such that gases are exposed to only fluorinated ethylene propylene (FEP) Teflon film and other Teflon surfaces to minimize any potential losses of BVOCs in the chamber. Solar radiation is simulated using 15 light-emitting diode (LED) panels, which have an emission strength up to 800 µmol m-2 s-1. Results of the initial characterization experiments are presented in detail. Background concentrations, mixing inside the gas exchange volume, and transfer rate of volatile organic compounds (VOCs) through PLUS under different humidity conditions are explored. Typical plant characteristics such as light- and temperature- dependent BVOC emissions are studied using six Quercus ilex trees and compared to previous studies. Results of an initial ozonolysis experiment of BVOC emissions from Quercus ilex at typical atmospheric concentrations inside SAPHIR are presented to demonstrate a typical experimental setup and the utility of the newly added plant chamber.

Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

International Nuclear Information System (INIS)

Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

2008-01-01

The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

Energy Technology Data Exchange (ETDEWEB)

Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

2008-10-15

The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

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)

A Unified Probabilistic Framework for Dose–Response Assessment of Human Health Effects

Science.gov (United States)

Slob, Wout

2015-01-01

Background When chemical health hazards have been identified, probabilistic dose–response assessment (“hazard characterization”) quantifies uncertainty and/or variability in toxicity as a function of human exposure. Existing probabilistic approaches differ for different types of endpoints or modes-of-action, lacking a unifying framework. Objectives We developed a unified framework for probabilistic dose–response assessment. Methods We established a framework based on four principles: a) individual and population dose responses are distinct; b) dose–response relationships for all (including quantal) endpoints can be recast as relating to an underlying continuous measure of response at the individual level; c) for effects relevant to humans, “effect metrics” can be specified to define “toxicologically equivalent” sizes for this underlying individual response; and d) dose–response assessment requires making adjustments and accounting for uncertainty and variability. We then derived a step-by-step probabilistic approach for dose–response assessment of animal toxicology data similar to how nonprobabilistic reference doses are derived, illustrating the approach with example non-cancer and cancer datasets. Results Probabilistically derived exposure limits are based on estimating a “target human dose” (HDMI), which requires risk management–informed choices for the magnitude (M) of individual effect being protected against, the remaining incidence (I) of individuals with effects ≥ M in the population, and the percent confidence. In the example datasets, probabilistically derived 90% confidence intervals for HDMI values span a 40- to 60-fold range, where I = 1% of the population experiences ≥ M = 1%–10% effect sizes. Conclusions Although some implementation challenges remain, this unified probabilistic framework can provide substantially more complete and transparent characterization of chemical hazards and support better-informed risk

Probabilistic generation assessment system of renewable energy in Korea

Directory of Open Access Journals (Sweden)

Yeonchan Lee

2016-01-01

Full Text Available This paper proposes probabilistic generation assessment system introduction of renewable energy generators. This paper is focused on wind turbine generator and solar cell generator. The proposed method uses an assessment model based on probabilistic model considering uncertainty of resources (wind speed and solar radiation. Equivalent generation function of the wind and solar farms are evaluated. The equivalent generation curves of wind farms and solar farms are assessed using regression analysis method using typical least square method from last actual generation data for wind farms. The proposed model is applied to Korea Renewable Generation System of 8 grouped 41 wind farms and 9 grouped around 600 solar farms in South Korea.

Research on probabilistic assessment method based on the corroded pipeline assessment criteria

International Nuclear Information System (INIS)

Zhang Guangli; Luo, Jinheng; Zhao Xinwei; Zhang Hua; Zhang Liang; Zhang Yi

2012-01-01

Pipeline integrity assessments are performed using conventional deterministic approaches, even though there are many uncertainties about the parameters in the pipeline integrity assessment. In this paper, a probabilistic assessment method is provided for the gas pipeline with corrosion defects based on the current corroded pipe evaluation criteria, and the failure probability of corroded pipelines due to the uncertainties of loadings, material property and measurement accuracy is estimated using Monte-Carlo technique. Furthermore, the sensitivity analysis approach is introduced to rank the influence of various random variables to the safety of pipeline. And the method to determine the critical defect size based on acceptable failure probability is proposed. Highlights: ► The folias factor in pipeline corrosion assessment methods was analyzed. ► The probabilistic method was applied in corrosion assessment methods. ► The influence of assessment variables to the reliability of pipeline was ranked. ► The acceptable failure probability was used to determine the critical defect size.

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

A Unified Probabilistic Framework for Dose-Response Assessment of Human Health Effects.

Science.gov (United States)

Chiu, Weihsueh A; Slob, Wout

2015-12-01

When chemical health hazards have been identified, probabilistic dose-response assessment ("hazard characterization") quantifies uncertainty and/or variability in toxicity as a function of human exposure. Existing probabilistic approaches differ for different types of endpoints or modes-of-action, lacking a unifying framework. We developed a unified framework for probabilistic dose-response assessment. We established a framework based on four principles: a) individual and population dose responses are distinct; b) dose-response relationships for all (including quantal) endpoints can be recast as relating to an underlying continuous measure of response at the individual level; c) for effects relevant to humans, "effect metrics" can be specified to define "toxicologically equivalent" sizes for this underlying individual response; and d) dose-response assessment requires making adjustments and accounting for uncertainty and variability. We then derived a step-by-step probabilistic approach for dose-response assessment of animal toxicology data similar to how nonprobabilistic reference doses are derived, illustrating the approach with example non-cancer and cancer datasets. Probabilistically derived exposure limits are based on estimating a "target human dose" (HDMI), which requires risk management-informed choices for the magnitude (M) of individual effect being protected against, the remaining incidence (I) of individuals with effects ≥ M in the population, and the percent confidence. In the example datasets, probabilistically derived 90% confidence intervals for HDMI values span a 40- to 60-fold range, where I = 1% of the population experiences ≥ M = 1%-10% effect sizes. Although some implementation challenges remain, this unified probabilistic framework can provide substantially more complete and transparent characterization of chemical hazards and support better-informed risk management decisions.

Engineering aspects of probabilistic risk assessment

International Nuclear Information System (INIS)

vonHerrmann, J.L.; Wood, P.J.

1984-01-01

Over the last decade, the use of probabilistic risk assessment (PRA) in the nuclear industry has expanded significantly. In these analyses the probabilities of experiencing certain undesired events (for example, a plant accident which results in damage to the nuclear fuel) are estimated and the consequences of these events are evaluated in terms of some common measure. These probabilities and consequences are then combined to form a representation of the risk associated with the plant studied. In the relatively short history of probabilistic risk assessment of nuclear power plants, the primary motivation for these studies has been the quantitative assessment of public risk associated with a single plant or group of plants. Accordingly, the primary product of most PRAs performed to date has been a 'risk curve' in which the probability (or expected frequency) of exceeding a certain consequence level is plotted against that consequence. The most common goal of these assessments has been to demonstrate the 'acceptability' of the calculated risk by comparison of the resultant risk curve to risk curves associated with other plants or with other societal risks. Presented here are brief descriptions of some alternate applications of PRAs, a discussion of how these other applications compare or contrast with the currently popular uses of PRA, and a discussion of the relative benefits of each

Probabilistic Tsunami Hazard Assessment: the Seaside, Oregon Pilot Study

Science.gov (United States)

Gonzalez, F. I.; Geist, E. L.; Synolakis, C.; Titov, V. V.

2004-12-01

A pilot study of Seaside, Oregon is underway, to develop methodologies for probabilistic tsunami hazard assessments that can be incorporated into Flood Insurance Rate Maps (FIRMs) developed by FEMA's National Flood Insurance Program (NFIP). Current NFIP guidelines for tsunami hazard assessment rely on the science, technology and methodologies developed in the 1970s; although generally regarded as groundbreaking and state-of-the-art for its time, this approach is now superseded by modern methods that reflect substantial advances in tsunami research achieved in the last two decades. In particular, post-1990 technical advances include: improvements in tsunami source specification; improved tsunami inundation models; better computational grids by virtue of improved bathymetric and topographic databases; a larger database of long-term paleoseismic and paleotsunami records and short-term, historical earthquake and tsunami records that can be exploited to develop improved probabilistic methodologies; better understanding of earthquake recurrence and probability models. The NOAA-led U.S. National Tsunami Hazard Mitigation Program (NTHMP), in partnership with FEMA, USGS, NSF and Emergency Management and Geotechnical agencies of the five Pacific States, incorporates these advances into site-specific tsunami hazard assessments for coastal communities in Alaska, California, Hawaii, Oregon and Washington. NTHMP hazard assessment efforts currently focus on developing deterministic, "credible worst-case" scenarios that provide valuable guidance for hazard mitigation and emergency management. The NFIP focus, on the other hand, is on actuarial needs that require probabilistic hazard assessments such as those that characterize 100- and 500-year flooding events. There are clearly overlaps in NFIP and NTHMP objectives. NTHMP worst-case scenario assessments that include an estimated probability of occurrence could benefit the NFIP; NFIP probabilistic assessments of 100- and 500-yr

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 risk assessment of HTGRs

International Nuclear Information System (INIS)

Fleming, K.N.; Houghton, W.J.; Hannaman, G.W.; Joksimovic, V.

1980-08-01

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the US Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed

Use of limited data to construct Bayesian networks for probabilistic risk assessment.

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina M.; Swiler, Laura Painton

2013-03-01

Probabilistic Risk Assessment (PRA) is a fundamental part of safety/quality assurance for nuclear power and nuclear weapons. Traditional PRA very effectively models complex hardware system risks using binary probabilistic models. However, traditional PRA models are not flexible enough to accommodate non-binary soft-causal factors, such as digital instrumentation&control, passive components, aging, common cause failure, and human errors. Bayesian Networks offer the opportunity to incorporate these risks into the PRA framework. This report describes the results of an early career LDRD project titled %E2%80%9CUse of Limited Data to Construct Bayesian Networks for Probabilistic Risk Assessment%E2%80%9D. The goal of the work was to establish the capability to develop Bayesian Networks from sparse data, and to demonstrate this capability by producing a data-informed Bayesian Network for use in Human Reliability Analysis (HRA) as part of nuclear power plant Probabilistic Risk Assessment (PRA). This report summarizes the research goal and major products of the research.

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

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

Computer aided probabilistic assessment of containment integrity

International Nuclear Information System (INIS)

Tsai, J.C.; Touchton, R.A.

1984-01-01

In the probabilistic risk assessment (PRA) of a nuclear power plant, there are three probability-based techniques which are widely used for event sequence frequency quantification (including nodal probability estimation). These three techniques are the event tree analysis, the fault tree analysis and the Bayesian approach for database development. In the barrier analysis for assessing radionuclide release to the environment in a PRA study, these techniques are employed to a greater extent in estimating conditions which could lead to failure of the fuel cladding and the reactor coolant system (RCS) pressure boundary, but to a lesser degree in the containment pressure boundary failure analysis. The main reason is that containment issues are currently still in a state of flux. In this paper, the authors describe briefly the computer programs currently used by the nuclear industry to do event tree analyses, fault tree analyses and the Bayesian update. The authors discuss how these computer aided probabilistic techniques might be adopted for failure analysis of the containment pressure boundary

A tiered approach for probabilistic ecological risk assessment of contaminated sites

International Nuclear Information System (INIS)

Zolezzi, M.; Nicolella, C.; Tarazona, J.V.

2005-01-01

This paper presents a tiered methodology for probabilistic ecological risk assessment. The proposed approach starts from deterministic comparison (ratio) of single exposure concentration and threshold or safe level calculated from a dose-response relationship, goes through comparison of probabilistic distributions that describe exposure values and toxicological responses of organisms to the chemical of concern, and finally determines the so called distribution-based quotients (DBQs). In order to illustrate the proposed approach, soil concentrations of 1,2,4-trichlorobenzene (1,2,4- TCB) measured in an industrial contaminated site were used for site-specific probabilistic ecological risks assessment. By using probabilistic distributions, the risk, which exceeds a level of concern for soil organisms with the deterministic approach, is associated to the presence of hot spots reaching concentrations able to affect acutely more than 50% of the soil species, while the large majority of the area presents 1,2,4- TCB concentrations below those reported as toxic [it

Probabilistic Modeling and Risk Assessment of Cable Icing

DEFF Research Database (Denmark)

Roldsgaard, Joan Hee

This dissertation addresses the issues related to icing of structures with special emphasis on bridge cables. Cable supported bridges in cold climate suffers for ice accreting on the cables, this poses three different undesirable situations. Firstly the changed shape of the cable due to ice...... preliminary framework is modified for assessing the probability of occurrence of in-cloud and precipitation icing and its duration. Different probabilistic models are utilized for the representation of the meteorological variables and their appropriateness is evaluated both through goodness-of-fit tests...... are influencing the two icing mechanisms and their duration. The model is found to be more sensitive to changes in the discretization levels of the input variables. Thirdly the developed operational probabilistic framework for the assessment of the expected number of occurrences of ice/snow accretion on bridge...

Probabilistic risk assessment of HTGRs

International Nuclear Information System (INIS)

Fleming, K.N.; Houghton, W.J.; Hannaman, G.W.; Joksimovic, V.

1981-01-01

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the U.S. Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed. (author)

Review of the Diablo Canyon probabilistic risk assessment

International Nuclear Information System (INIS)

Bozoki, G.E.; Fitzpatrick, R.G.; Bohn, M.P.; Sabek, M.G.; Ravindra, M.K.; Johnson, J.J.

1994-08-01

This report details the review of the Diablo Canyon Probabilistic Risk Assessment (DCPRA). The study was performed under contract from the Probabilistic Risk Analysis Branch, Office of Nuclear Reactor Research, USNRC by Brookhaven National Laboratory. The DCPRA is a full scope Level I effort and although the review touched on all aspects of the PRA, the internal events and seismic events received the vast majority of the review effort. The report includes a number of independent systems analyses sensitivity studies, importance analyses as well as conclusions on the adequacy of the DCPRA for use in the Diablo Canyon Long Term Seismic Program

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.

Comparative study of probabilistic methodologies for small signal stability assessment

Energy Technology Data Exchange (ETDEWEB)

Rueda, J.L.; Colome, D.G. [Universidad Nacional de San Juan (IEE-UNSJ), San Juan (Argentina). Inst. de Energia Electrica], Emails: joseluisrt@iee.unsj.edu.ar, colome@iee.unsj.edu.ar

2009-07-01

Traditional deterministic approaches for small signal stability assessment (SSSA) are unable to properly reflect the existing uncertainties in real power systems. Hence, the probabilistic analysis of small signal stability (SSS) is attracting more attention by power system engineers. This paper discusses and compares two probabilistic methodologies for SSSA, which are based on the two point estimation method and the so-called Monte Carlo method, respectively. The comparisons are based on the results obtained for several power systems of different sizes and with different SSS performance. It is demonstrated that although with an analytical approach the amount of computation of probabilistic SSSA can be reduced, the different degrees of approximations that are adopted, lead to deceptive results. Conversely, Monte Carlo based probabilistic SSSA can be carried out with reasonable computational effort while holding satisfactory estimation precision. (author)

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)

Probabilistic risk assessment for six vapour intrusion algorithms

NARCIS (Netherlands)

Provoost, J.; Reijnders, L.; Bronders, J.; Van Keer, I.; Govaerts, S.

2014-01-01

A probabilistic assessment with sensitivity analysis using Monte Carlo simulation for six vapour intrusion algorithms, used in various regulatory frameworks for contaminated land management, is presented here. In addition a deterministic approach with default parameter sets is evaluated against

Integrated probabilistic risk assessment for nanoparticles: the case of nanosilica in food.

Science.gov (United States)

Jacobs, Rianne; van der Voet, Hilko; Ter Braak, Cajo J F

Insight into risks of nanotechnology and the use of nanoparticles is an essential condition for the social acceptance and safe use of nanotechnology. One of the problems with which the risk assessment of nanoparticles is faced is the lack of data, resulting in uncertainty in the risk assessment. We attempt to quantify some of this uncertainty by expanding a previous deterministic study on nanosilica (5-200 nm) in food into a fully integrated probabilistic risk assessment. We use the integrated probabilistic risk assessment method in which statistical distributions and bootstrap methods are used to quantify uncertainty and variability in the risk assessment. Due to the large amount of uncertainty present, this probabilistic method, which separates variability from uncertainty, contributed to a better understandable risk assessment. We found that quantifying the uncertainties did not increase the perceived risk relative to the outcome of the deterministic study. We pinpointed particular aspects of the hazard characterization that contributed most to the total uncertainty in the risk assessment, suggesting that further research would benefit most from obtaining more reliable data on those aspects.

Integrated probabilistic risk assessment for nanoparticles: the case of nanosilica in food

International Nuclear Information System (INIS)

Jacobs, Rianne; Voet, Hilko van der; Braak, Cajo J. F. ter

2015-01-01

Insight into risks of nanotechnology and the use of nanoparticles is an essential condition for the social acceptance and safe use of nanotechnology. One of the problems with which the risk assessment of nanoparticles is faced is the lack of data, resulting in uncertainty in the risk assessment. We attempt to quantify some of this uncertainty by expanding a previous deterministic study on nanosilica (5–200 nm) in food into a fully integrated probabilistic risk assessment. We use the integrated probabilistic risk assessment method in which statistical distributions and bootstrap methods are used to quantify uncertainty and variability in the risk assessment. Due to the large amount of uncertainty present, this probabilistic method, which separates variability from uncertainty, contributed to a better understandable risk assessment. We found that quantifying the uncertainties did not increase the perceived risk relative to the outcome of the deterministic study. We pinpointed particular aspects of the hazard characterization that contributed most to the total uncertainty in the risk assessment, suggesting that further research would benefit most from obtaining more reliable data on those aspects

Integrated probabilistic risk assessment for nanoparticles: the case of nanosilica in food

Energy Technology Data Exchange (ETDEWEB)

Jacobs, Rianne, E-mail: rianne.jacobs@wur.nl; Voet, Hilko van der; Braak, Cajo J. F. ter [Wageningen University and Research Centre, Biometris (Netherlands)

2015-06-15

Insight into risks of nanotechnology and the use of nanoparticles is an essential condition for the social acceptance and safe use of nanotechnology. One of the problems with which the risk assessment of nanoparticles is faced is the lack of data, resulting in uncertainty in the risk assessment. We attempt to quantify some of this uncertainty by expanding a previous deterministic study on nanosilica (5–200 nm) in food into a fully integrated probabilistic risk assessment. We use the integrated probabilistic risk assessment method in which statistical distributions and bootstrap methods are used to quantify uncertainty and variability in the risk assessment. Due to the large amount of uncertainty present, this probabilistic method, which separates variability from uncertainty, contributed to a better understandable risk assessment. We found that quantifying the uncertainties did not increase the perceived risk relative to the outcome of the deterministic study. We pinpointed particular aspects of the hazard characterization that contributed most to the total uncertainty in the risk assessment, suggesting that further research would benefit most from obtaining more reliable data on those aspects.

Comparative Probabilistic Assessment of Occupational Pesticide Exposures Based on Regulatory Assessments

Science.gov (United States)

Pouzou, Jane G.; Cullen, Alison C.; Yost, Michael G.; Kissel, John C.; Fenske, Richard A.

2018-01-01

Implementation of probabilistic analyses in exposure assessment can provide valuable insight into the risks of those at the extremes of population distributions, including more vulnerable or sensitive subgroups. Incorporation of these analyses into current regulatory methods for occupational pesticide exposure is enabled by the exposure data sets and associated data currently used in the risk assessment approach of the Environmental Protection Agency (EPA). Monte Carlo simulations were performed on exposure measurements from the Agricultural Handler Exposure Database and the Pesticide Handler Exposure Database along with data from the Exposure Factors Handbook and other sources to calculate exposure rates for three different neurotoxic compounds (azinphos methyl, acetamiprid, emamectin benzoate) across four pesticide-handling scenarios. Probabilistic estimates of doses were compared with the no observable effect levels used in the EPA occupational risk assessments. Some percentage of workers were predicted to exceed the level of concern for all three compounds: 54% for azinphos methyl, 5% for acetamiprid, and 20% for emamectin benzoate. This finding has implications for pesticide risk assessment and offers an alternative procedure that may be more protective of those at the extremes of exposure than the current approach. PMID:29105804

Comparative Probabilistic Assessment of Occupational Pesticide Exposures Based on Regulatory Assessments.

Science.gov (United States)

Pouzou, Jane G; Cullen, Alison C; Yost, Michael G; Kissel, John C; Fenske, Richard A

2017-11-06

Implementation of probabilistic analyses in exposure assessment can provide valuable insight into the risks of those at the extremes of population distributions, including more vulnerable or sensitive subgroups. Incorporation of these analyses into current regulatory methods for occupational pesticide exposure is enabled by the exposure data sets and associated data currently used in the risk assessment approach of the Environmental Protection Agency (EPA). Monte Carlo simulations were performed on exposure measurements from the Agricultural Handler Exposure Database and the Pesticide Handler Exposure Database along with data from the Exposure Factors Handbook and other sources to calculate exposure rates for three different neurotoxic compounds (azinphos methyl, acetamiprid, emamectin benzoate) across four pesticide-handling scenarios. Probabilistic estimates of doses were compared with the no observable effect levels used in the EPA occupational risk assessments. Some percentage of workers were predicted to exceed the level of concern for all three compounds: 54% for azinphos methyl, 5% for acetamiprid, and 20% for emamectin benzoate. This finding has implications for pesticide risk assessment and offers an alternative procedure that may be more protective of those at the extremes of exposure than the current approach. © 2017 Society for Risk Analysis.

Application of probabilistic risk assessment to advanced liquid metal reactor designs

International Nuclear Information System (INIS)

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

1987-01-01

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

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)

Probabilistic assessment of SGTR management

International Nuclear Information System (INIS)

Champ, M.; Cornille, Y.; Lanore, J.M.

1989-04-01

In case of steam generator tube rupture (SGTR) event, in France, the mitigation of accident relies on operator intervention, by applying a specific accidental procedure. A detailed probabilistic analysis has been conducted which required the assessment of the failure probability of the operator actions, and for that purpose it was necessary to estimate the time available for the operator to apply the adequate procedure for various sequences. The results indicate that by taking into account the delays and the existence of adequate accidental procedures, the risk is reduced to a reasonably low level

Probabilistic Seismic Risk Assessment in Manizales, Colombia:Quantifying Losses for Insurance Purposes

Institute of Scientific and Technical Information of China (English)

Mario A.Salgado-Gálvez; Gabriel A.Bernal; Daniela Zuloaga; Mabel C.Marulanda; Omar-Darío Cardona; Sebastián Henao

2017-01-01

A fully probabilistic seismic risk assessment was developed in Manizales,Colombia,considering assets of different types.The first type includes elements that are part of the water and sewage network,and the second type includes public and private buildings.This assessment required the development of a probabilistic seismic hazard analysis that accounts for the dynamic soil response,assembling high resolution exposure databases,and the development of damage models for different types of elements.The economic appraisal of the exposed assets was developed together with specialists of the water utilities company of Manizales and the city administration.The risk assessment was performed using several Comprehensive Approach to Probabilistic Risk Assessment modules as well as the R-System,obtaining results in terms of traditional metrics such as loss exceedance curve,average annual loss,and probable maximum loss.For the case of pipelines,repair rates were also estimated.The results for the water and sewage network were used in activities related to the expansion and maintenance strategies,as well as for the exploration of financial retention and transfer alternatives using insurance schemes based on technical,probabilistic,and prospective damage and loss estimations.In the case of the buildings,the results were used in the update of the technical premium values of the existing collective insurance scheme.

Failure probability assessment of wall-thinned nuclear pipes using probabilistic fracture mechanics

International Nuclear Information System (INIS)

Lee, Sang-Min; Chang, Yoon-Suk; Choi, Jae-Boong; Kim, Young-Jin

2006-01-01

The integrity of nuclear piping system has to be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc., are required. Up to now, this has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approaches are considered as an appropriate method for piping system evaluation. The objectives of this paper are to estimate the failure probabilities of wall-thinned pipes in nuclear secondary systems and to propose limited operating conditions under different types of loadings. To do this, a probabilistic assessment program using reliability index and simulation techniques was developed and applied to evaluate failure probabilities of wall-thinned pipes subjected to internal pressure, bending moment and combined loading of them. The sensitivity analysis results as well as prototypal integrity assessment results showed a promising applicability of the probabilistic assessment program, necessity of practical evaluation reflecting combined loading condition and operation considering limited condition

Insights gained through probabilistic risk assessments

International Nuclear Information System (INIS)

Hitchler, M.J.; Burns, N.L.; Liparulo, N.J.; Mink, F.J.

1987-01-01

The insights gained through a comparison of seven probabilistic risk assessments (PRA) studies (Italian PUN, Sizewell B, Ringhals 2, Millstone 3, Zion 1 and 2, Oconee 3, and Seabrook) included insights regarding the adequacy of the PRA technology utilized in the studies and the potential areas for improvement and insights regarding the adequacy of plant designs and how PRA has been utilized to enhance the design and operation of nuclear power plants

Exploring the uncertainties in cancer risk assessment using the integrated probabilistic risk assessment (IPRA) approach.

Science.gov (United States)

Slob, Wout; Bakker, Martine I; Biesebeek, Jan Dirk Te; Bokkers, Bas G H

2014-08-01

Current methods for cancer risk assessment result in single values, without any quantitative information on the uncertainties in these values. Therefore, single risk values could easily be overinterpreted. In this study, we discuss a full probabilistic cancer risk assessment approach in which all the generally recognized uncertainties in both exposure and hazard assessment are quantitatively characterized and probabilistically evaluated, resulting in a confidence interval for the final risk estimate. The methodology is applied to three example chemicals (aflatoxin, N-nitrosodimethylamine, and methyleugenol). These examples illustrate that the uncertainty in a cancer risk estimate may be huge, making single value estimates of cancer risk meaningless. Further, a risk based on linear extrapolation tends to be lower than the upper 95% confidence limit of a probabilistic risk estimate, and in that sense it is not conservative. Our conceptual analysis showed that there are two possible basic approaches for cancer risk assessment, depending on the interpretation of the dose-incidence data measured in animals. However, it remains unclear which of the two interpretations is the more adequate one, adding an additional uncertainty to the already huge confidence intervals for cancer risk estimates. © 2014 Society for Risk Analysis.

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

Probabilistic cumulative risk assessment of anti-androgenic pesticides in food

DEFF Research Database (Denmark)

Müller, Anne Kirstine; Nielsen, Elsa

2008-01-01

A cumulative risk assessment of three anti-androgenic pesticides vinclozolin, procymidone and prochloraz in combination has been carried out using an Integrated Probabilistic Risk Assessment (IPRA) model. In the model, variability in both exposure and sensitivity between individuals were combined...

Probabilistic cumulative risk assessment of anti-androgenic pesticides in food.

NARCIS (Netherlands)

Müller, A.K.; Bosgra, S.; Boon, P.E.; van der Voet, H.; Nielsen, E.; Ladefoged, O.

2009-01-01

In this paper, we present a cumulative risk assessment of three anti-androgenic pesticides (vinclozolin, procymidone and prochloraz) using the relative potency factor (RPF) approach and an integrated probabilistic risk assessment (IPRA) model. RPFs for each substance were estimated for three

Probabilistic cumulative risk assessment of anti-androgenic pesticides in food

NARCIS (Netherlands)

Muller, A.K.; Bosgra, S.; Boon, P.E.; Voet, van der H.; Nielsen, E.; Ladefoged, O.

2009-01-01

In this paper, we present a cumulative risk assessment of three anti-androgenic pesticides (vinclozolin, procymidone and prochloraz) using the relative potency factor (RPF) approach and an integrated probabilistic risk assessment (IPRA) model. RPFs for each substance were estimated for three

Overview of the probabilistic risk assessment approach

International Nuclear Information System (INIS)

Reed, J.W.

1985-01-01

The techniques of probabilistic risk assessment (PRA) are applicable to Department of Energy facilities. The background and techniques of PRA are given with special attention to seismic, wind and flooding external events. A specific application to seismic events is provided to demonstrate the method. However, the PRA framework is applicable also to wind and external flooding. 3 references, 8 figures, 1 table

Probabilistic approach to EMP assessment

International Nuclear Information System (INIS)

Bevensee, R.M.; Cabayan, H.S.; Deadrick, F.J.; Martin, L.C.; Mensing, R.W.

1980-09-01

The development of nuclear EMP hardness requirements must account for uncertainties in the environment, in interaction and coupling, and in the susceptibility of subsystems and components. Typical uncertainties of the last two kinds are briefly summarized, and an assessment methodology is outlined, based on a probabilistic approach that encompasses the basic concepts of reliability. It is suggested that statements of survivability be made compatible with system reliability. Validation of the approach taken for simple antenna/circuit systems is performed with experiments and calculations that involve a Transient Electromagnetic Range, numerical antenna modeling, separate device failure data, and a failure analysis computer program

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)

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

Development of a Quantitative Framework for Regulatory Risk Assessments: Probabilistic Approaches

International Nuclear Information System (INIS)

Wilmot, R.D.

2003-11-01

The Swedish regulators have been active in the field of performance assessment for many years and have developed sophisticated approaches to the development of scenarios and other aspects of assessments. These assessments have generally used dose as the assessment end-point and have been based on deterministic calculations. Recently introduced Swedish regulations have introduced a risk criterion for radioactive waste disposal: the annual risk of harmful effects after closure of a disposal facility should not exceed 10 -6 for a representative individual in the group exposed to the greatest risk. A recent review of the overall structure of risk assessments in safety cases concluded that there are a number of decisions and assumptions in the development of a risk assessment methodology that could potentially affect the calculated results. Regulatory understanding of these issues, potentially supported by independent calculations, is important in preparing for review of a proponent's risk assessment. One approach to evaluating risk in performance assessments is to use the concept of probability to express uncertainties, and to propagate these probabilities through the analysis. This report describes the various approaches available for undertaking such probabilistic analyses, both as a means of accounting for uncertainty in the determination of risk and more generally as a means of sensitivity and uncertainty analysis. The report discusses the overall nature of probabilistic analyses and how they are applied to both the calculation of risk and sensitivity analyses. Several approaches are available, including differential analysis, response surface methods and simulation. Simulation is the approach most commonly used, both in assessments for radioactive waste disposal and in other subject areas, and the report describes the key stages of this approach in detail. Decisions relating to the development of input PDFs, sampling methods (including approaches to the treatment

Simulation of the beam guiding of the SAPHIR experiment by means of a differential-equation model

International Nuclear Information System (INIS)

Greve, T.

1991-08-01

This paper shows the numerical simulation of a beam line by means of a model of differential equations simulating the beam line from the Bonn Electron Stretcher Accelerator ELSA to the SAPHIR spectrometer. Furthermore a method for calculating the initial values based on measurements of beam profiles is being discussed. (orig.) [de

The central drift chamber of the SAPHIR detector - implementation into the experiment and study of its properties

International Nuclear Information System (INIS)

Haas, K.M.

1992-01-01

At the Bonn accelerator facility ELSA the large solid angle detector SAPHIR was built for the investigation of photon induced reactions. A main component of SAPHIR is the central drift chamber (CDC) matching the magneto gap of 1m 3 . The diameter of the in total 1828 hexagonal drift cells is about 18 mm. The subject of this paper is the implementation of the CDC in the experiment. After the description of the hardware follows the presentation of the software tools for filtering and monitoring the data, which have been developed and tested. An algorithm for extracting the space time relationship is presented. The properties of the chamber with an improved gas mixture (Helium/Neon/Isobutane8 21.25:63.75:15) have been investigated. A spatial resolution of about 200 μm was achieved. The efficiency of the chamber is 97% at a tagged photon of 5x10 4 per second crossing the chamber. (orig.) [de

Current status and future expectation concerning probabilistic risk assessment of NPPs. 1. Features and issues of probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Yamashita, Masahiro

2012-01-01

Probabilistic risk assessment (PRA) of Nuclear Power Plants (NPPs) could play an important role in assuring safety of NPPs. However PRA had not always effectively used, which was indicated in Japanese government's report on Fukushima Daiichi NPP accident. At the Risk Technical Committee (RTC) of Standards Committee of Atomic Energy Society of Japan, preparation of standards (implementing criteria) focusing on PRA methodology and investigation on basic philosophy for use of PRA had been in progress. Based on activities of RTC, a serial in three articles including this described current status and future expectation concerning probabilistic risk assessment of NPPs. This article introduced features and issues of PRA methodology related to the use of PRA. Features of PRA methodology could be shown as (1) systematic and comprehensive understanding of risk, (2) support of grading approach, (3) identification of effective safety upgrade measures and (4) quantitative understanding of effects of uncertainty. Issues of PRA methodology were (1) extension of PRA application area, (2) upgrade of PRA methodology, (3) quality assurance of PRA, (4) treatment of uncertainty and (5) quantitative evaluation criteria. (T. Tanaka)

Contribution to the microwave characterisation of superconductive materials by means of sapphire resonators; Contribution a la caracterisation hyperfrequence de materiaux supraconducteurs par des resonateurs-saphirs

Energy Technology Data Exchange (ETDEWEB)

Hanus, Xavier

1993-12-06

The objective of this research thesis is to find a compact resonant structure which would allow the residual surface impedance of superconductive samples to be simply, quickly and economically characterised. The author first explains why he decided to use a sapphire single-crystal as inner dielectric, given some performance reached by resonant structures equipped with such inner dielectrics, and given constraints adopted from the start. He explains the origin of microwave losses which appear in this type of resonant structure, i.e. respectively the surface impedance as far as metallic losses are concerned, and the sapphire dielectric loss angle for as far as dielectric losses are concerned. The experimental installation and the principle of microwave measurements are described. The performance of different possible solutions of resonant structures from starting criteria is presented. The solution of the cavity-sapphire with a TE{sub 011} resonant mode is derived [French] Le but de cette etude est de trouver une structure resonnante compacte permettant de caracteriser simplement, rapidement et economiquement l'impedance de surface residuelle d'echantillons supraconducteurs. Les contraintes de mise en oeuvre et les performances atteintes par des resonateurs avec saphirs synthetiques justifient le choix d'un tel dielectrique a faible angle de perte. L'evaluation des performances experimentales appuyee par des modelesanalytiques permet de rejeter differentes solutions. Ainsi les resonateurs fermes avec saphirs minces sont rejetes en raison des mauvais contacts metalliques. Les resonateurs ouverts avec saphirs minces et epais sont egalement rejetes, meme pour les modes de resonance en principe confines, en raison des pertes par rayonnement. La seule solution est donc d'utiliser une cavite-saphir TE{sub 011} qui offre une configuration de champs naturellement confines. Des mesures sur une premiere cavite en niobium massif ont permis de selectionner un saphir obtenu par

Comparative analysis of deterministic and probabilistic fracture mechanical assessment tools

Energy Technology Data Exchange (ETDEWEB)

Heckmann, Klaus [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany); Saifi, Qais [VTT Technical Research Centre of Finland, Espoo (Finland)

2016-11-15

Uncertainties in material properties, manufacturing processes, loading conditions and damage mechanisms complicate the quantification of structural reliability. Probabilistic structure mechanical computing codes serve as tools for assessing leak- and break probabilities of nuclear piping components. Probabilistic fracture mechanical tools were compared in different benchmark activities, usually revealing minor, but systematic discrepancies between results of different codes. In this joint paper, probabilistic fracture mechanical codes are compared. Crack initiation, crack growth and the influence of in-service inspections are analyzed. Example cases for stress corrosion cracking and fatigue in LWR conditions are analyzed. The evolution of annual failure probabilities during simulated operation time is investigated, in order to identify the reasons for differences in the results of different codes. The comparison of the tools is used for further improvements of the codes applied by the partners.

The muon trigger of the SAPHIR shower detector

International Nuclear Information System (INIS)

Rufeger-Hurek, H.

1989-12-01

The muon trigger system of the SAPHIR shower counter consists of 4 scintillation counters. The total trigger rate of cosmic muons is about 55 Hz which is reduced to about 45 Hz by the selecting algorithms. This rate of clean muon events allows a simultaneous monitoring of the whole electronics system and the calibration of the gas sandwich detector by measuring the gas gain. The dependences of the signals on the geometry have been simulated with the help of a Monte Carlo program. The comparison of simulated and measured pulse heights shows that faults in the electronics as well as defects in the detector hardware, e.g., the HV system, or temperature effects, can be recognized at the level of a few percent. In addition the muon signals are used to determine the calibration factor for each cathode channel individually. (orig.) [de

The probabilistic approach in the licensing process and the development of probabilistic risk assessment methodology in Japan

International Nuclear Information System (INIS)

Togo, Y.; Sato, K.

1981-01-01

The probabilistic approach has long seemed to be one of the most comprehensive methods for evaluating the safety of nuclear plants. So far, most of the guidelines and criteria for licensing are based on the deterministic concept. However, there have been a few examples to which the probabilistic approach was directly applied, such as the evaluation of aircraft crashes and turbine missiles. One may find other examples of such applications. However, a much more important role is now to be played by this concept, in implementing the 52 recommendations from the lessons learned from the TMI accident. To develop the probabilistic risk assessment methodology most relevant to Japanese situations, a five-year programme plan has been adopted and is to be conducted by the Japan Atomic Research Institute from fiscal 1980. Various problems have been identified and are to be solved through this programme plan. The current status of developments is described together with activities outside the government programme. (author)

First measurement of the reactions γp→K+Λ and γp→K+Σ0 with SAPHIR at ELSA

International Nuclear Information System (INIS)

Lindemann, L.

1993-04-01

This report can be subdivided into two main parts. The first part concerns the reconstruction program which has been developed to analyse the data taken with the large solid angle detector SAPHIR which is in operation at the Bonn electron accelerator facility ELSA. A survey on this program is given and some improvements as well as the efficiency concerning real data are discussed. The subject of the second part concerns the measurements of the reactions γp→K + Λand γp→K + Σ 0 . The analysis of a sample of data taken with the SAPHIR in June 1992 is discussed in detail. As a result of this analysis total and differential cross sections as well as the recoil polarization for the two processes are presented. In particular the first measurement of the Σ 0 polarization in photoproduction can be reported. (orig.)

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 assessment in the nuclear power industry

International Nuclear Information System (INIS)

Fullwood, R.R.; Hall, R.E.

1988-01-01

This book describes the more important improvements in risk assessment methodology developed over the last decade. The book covers the following areas - a general view of risk pertaining to nuclear power, mathematics necessary to understand the text, a concise overview of the light water reactors and their features for protecting the public, probabilities and consequences calculated to form risk assessment to the plant, and 34 applications of probabilistic risk assessment (PRA) in the power generation industry. There is a glossary of acronyms and unusual words and a list of references. (author)

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)

A methodology for reviewing Probabilistic Risk Assessments

International Nuclear Information System (INIS)

Derby, S.L.

1983-01-01

The starting point for peer review of a Probabilistic Risk Assessment (PRA) is a clear understanding of how the risk estimate was prepared and of what contributions dominate the calculation. The problem facing the reviewers is how to cut through the complex details of a PRA to gain this understanding. This paper presents a structured, analytical procedure that solves this problem. The effectiveness of this solution is demonstrated by an application on the Zion Probabilistic Safety Study. The procedure found the three dominant initiating events and provided a simplified reconstruction of the calculation of the risk estimate. Significant assessments of uncertainty were also identified. If peer review disputes the accuracy of these judgments, then the revised risk estimate could significantly increase. The value of this procedure comes from having a systematic framework for the PRA review. Practical constraints limit the time and qualified people needed for an adequate review. Having the established framework from this procedure as a starting point, reviewers can focus most of their attention on the accuracy and the completeness of the calculation. Time wasted at the start of the review is reduced by first using this procedure to sort through the technical details of the PRA and to reconstruct the risk estimate from dominant contributions

Methods improvements incorporated into the SAPHIRE ASP models

International Nuclear Information System (INIS)

Sattison, M.B.; Blackman, H.S.; Novack, S.D.; Smith, C.L.; Rasmuson, D.M.

1994-01-01

The Office for Analysis and Evaluation of Operational Data (AEOD) has sought the assistance of the Idaho National Engineering Laboratory (INEL) to make some significant enhancements to the SAPHIRE-based Accident Sequence Precursor (ASP) models recently developed by the INEL. The challenge of this project is to provide the features of a full-scale PRA within the framework of the simplified ASP models. Some of these features include: (1) uncertainty analysis addressing the standard PRA uncertainties and the uncertainties unique to the ASP models and methodology, (2) incorporation and proper quantification of individual human actions and the interaction among human actions, (3) enhanced treatment of common cause failures, and (4) extension of the ASP models to more closely mimic full-scale PRAs (inclusion of more initiators, explicitly modeling support system failures, etc.). This paper provides an overview of the methods being used to make the above improvements

Methods improvements incorporated into the SAPHIRE ASP models

International Nuclear Information System (INIS)

Sattison, M.B.; Blackman, H.S.; Novack, S.D.

1995-01-01

The Office for Analysis and Evaluation of Operational Data (AEOD) has sought the assistance of the Idaho National Engineering Laboratory (INEL) to make some significant enhancements to the SAPHIRE-based Accident Sequence Precursor (ASP) models recently developed by the INEL. The challenge of this project is to provide the features of a full-scale PRA within the framework of the simplified ASP models. Some of these features include: (1) uncertainty analysis addressing the standard PRA uncertainties and the uncertainties unique to the ASP models and methods, (2) incorporation and proper quantification of individual human actions and the interaction among human actions, (3) enhanced treatment of common cause failures, and (4) extension of the ASP models to more closely mimic full-scale PRAs (inclusion of more initiators, explicitly modeling support system failures, etc.). This paper provides an overview of the methods being used to make the above improvements

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 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)

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)

Probabilistic Risk Assessment to Inform Decision Making: Frequently Asked Questions

Science.gov (United States)

General concepts and principles of Probabilistic Risk Assessment (PRA), describe how PRA can improve the bases of Agency decisions, and provide illustrations of how PRA has been used in risk estimation and in describing the uncertainty in decision making.

Deterministic and Probabilistic Serviceability Assessment of Footbridge Vibrations due to a Single Walker Crossing

Directory of Open Access Journals (Sweden)

Cristoforo Demartino

2018-01-01

Full Text Available This paper presents a numerical study on the deterministic and probabilistic serviceability assessment of footbridge vibrations due to a single walker crossing. The dynamic response of the footbridge is analyzed by means of modal analysis, considering only the first lateral and vertical modes. Single span footbridges with uniform mass distribution are considered, with different values of the span length, natural frequencies, mass, and structural damping and with different support conditions. The load induced by a single walker crossing the footbridge is modeled as a moving sinusoidal force either in the lateral or in the vertical direction. The variability of the characteristics of the load induced by walkers is modeled using probability distributions taken from the literature defining a Standard Population of walkers. Deterministic and probabilistic approaches were adopted to assess the peak response. Based on the results of the simulations, deterministic and probabilistic vibration serviceability assessment methods are proposed, not requiring numerical analyses. Finally, an example of the application of the proposed method to a truss steel footbridge is presented. The results highlight the advantages of the probabilistic procedure in terms of reliability quantification.

Bayesian Hierarchical Structure for Quantifying Population Variability to Inform Probabilistic Health Risk Assessments.

Science.gov (United States)

Shao, Kan; Allen, Bruce C; Wheeler, Matthew W

2017-10-01

Human variability is a very important factor considered in human health risk assessment for protecting sensitive populations from chemical exposure. Traditionally, to account for this variability, an interhuman uncertainty factor is applied to lower the exposure limit. However, using a fixed uncertainty factor rather than probabilistically accounting for human variability can hardly support probabilistic risk assessment advocated by a number of researchers; new methods are needed to probabilistically quantify human population variability. We propose a Bayesian hierarchical model to quantify variability among different populations. This approach jointly characterizes the distribution of risk at background exposure and the sensitivity of response to exposure, which are commonly represented by model parameters. We demonstrate, through both an application to real data and a simulation study, that using the proposed hierarchical structure adequately characterizes variability across different populations. © 2016 Society for Risk Analysis.

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 Radiological Performance Assessment Modeling and Uncertainty

Science.gov (United States)

Tauxe, J.

2004-12-01

A generic probabilistic radiological Performance Assessment (PA) model is presented. The model, built using the GoldSim systems simulation software platform, concerns contaminant transport and dose estimation in support of decision making with uncertainty. Both the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE) require assessments of potential future risk to human receptors of disposal of LLW. Commercially operated LLW disposal facilities are licensed by the NRC (or agreement states), and the DOE operates such facilities for disposal of DOE-generated LLW. The type of PA model presented is probabilistic in nature, and hence reflects the current state of knowledge about the site by using probability distributions to capture what is expected (central tendency or average) and the uncertainty (e.g., standard deviation) associated with input parameters, and propagating through the model to arrive at output distributions that reflect expected performance and the overall uncertainty in the system. Estimates of contaminant release rates, concentrations in environmental media, and resulting doses to human receptors well into the future are made by running the model in Monte Carlo fashion, with each realization representing a possible combination of input parameter values. Statistical summaries of the results can be compared to regulatory performance objectives, and decision makers are better informed of the inherently uncertain aspects of the model which supports their decision-making. While this information may make some regulators uncomfortable, they must realize that uncertainties which were hidden in a deterministic analysis are revealed in a probabilistic analysis, and the chance of making a correct decision is now known rather than hoped for. The model includes many typical features and processes that would be part of a PA, but is entirely fictitious. This does not represent any particular site and is meant to be a generic example. A

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

Probabilistic inhalation risk assessment due to radioactivity released from coal fired thermal power plants

International Nuclear Information System (INIS)

Tiwari, M.; Ajmal, P.Y.; Bhangare, R.C.; Sahu, S.K.; Pandit, G.G.

2014-01-01

This paper deals with assessment of radiological risk to the general public around in the neighborhood of a 1000 MWe coal-based thermal power plant. We have used Monte Carlo simulation for characterization of uncertainty in inhalation risk due to radionuclide escaping from the stack of thermal power plant. Monte Carlo simulation treats parameters as random variables bound to a given probabilistic distribution to evaluate the distribution of the resulting output. Risk assessment is the process that estimates the likelihood of occurrence of adverse effects to humans and ecological receptors as a result of exposure to hazardous chemical, radiation, and/or biological agents. Quantitative risk characterization involves evaluating exposure estimates against a benchmark of toxicity, such as a cancer slope factor. Risk is calculated by multiplying the carcinogenic slope factor (SF) of the radionuclide by the dose an individual receives. The collective effective doses to the population living in the neighborhood of coal-based thermal power plant were calculated using Gaussian plume dispersion model. Monte Carlo Analysis is the most widely used probabilistic method in risk assessment. The MCA technique treats any uncertain parameter as random variable that obeys a given probabilistic distribution. This technique is widely used for analyzing probabilistic uncertainty. In MCA computer simulation are used to combine multiple probability distributions associated with the dose and SF depicted in risk equation. Thus we get a probabilistic distribution for the risk

Risk assessment using probabilistic standards

International Nuclear Information System (INIS)

Avila, R.

2004-01-01

A core element of risk is uncertainty represented by plural outcomes and their likelihood. No risk exists if the future outcome is uniquely known and hence guaranteed. The probability that we will die some day is equal to 1, so there would be no fatal risk if sufficiently long time frame is assumed. Equally, rain risk does not exist if there was 100% assurance of rain tomorrow, although there would be other risks induced by the rain. In a formal sense, any risk exists if, and only if, more than one outcome is expected at a future time interval. In any practical risk assessment we have to deal with uncertainties associated with the possible outcomes. One way of dealing with the uncertainties is to be conservative in the assessments. For example, we may compare the maximal exposure to a radionuclide with a conservatively chosen reference value. In this case, if the exposure is below the reference value then it is possible to assure that the risk is low. Since single values are usually compared; this approach is commonly called 'deterministic'. Its main advantage lies in the simplicity and in that it requires minimum information. However, problems arise when the reference values are actually exceeded or might be exceeded, as in the case of potential exposures, and when the costs for realizing the reference values are high. In those cases, the lack of knowledge on the degree of conservatism involved impairs a rational weighing of the risks against other interests. In this presentation we will outline an approach for dealing with uncertainties that in our opinion is more consistent. We will call it a 'fully probabilistic risk assessment'. The essence of this approach consists in measuring the risk in terms of probabilities, where the later are obtained from comparison of two probabilistic distributions, one reflecting the uncertainties in the outcomes and one reflecting the uncertainties in the reference value (standard) used for defining adverse outcomes. Our first aim

Use of Probabilistic Risk Assessment in Shuttle Decision Making Process

Science.gov (United States)

Boyer, Roger L.; Hamlin, Teri, L.

2011-01-01

This slide presentation reviews the use of Probabilistic Risk Assessment (PRA) to assist in the decision making for the shuttle design and operation. Probabilistic Risk Assessment (PRA) is a comprehensive, structured, and disciplined approach to identifying and analyzing risk in complex systems and/or processes that seeks answers to three basic questions: (i.e., what can go wrong? what is the likelihood of these occurring? and what are the consequences that could result if these occur?) The purpose of the Shuttle PRA (SPRA) is to provide a useful risk management tool for the Space Shuttle Program (SSP) to identify strengths and possible weaknesses in the Shuttle design and operation. SPRA was initially developed to support upgrade decisions, but has evolved into a tool that supports Flight Readiness Reviews (FRR) and near real-time flight decisions. Examples of the use of PRA for the shuttle are reviewed.

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

Wind effects on long-span bridges: Probabilistic wind data format for buffeting and VIV load assessments

Science.gov (United States)

Hoffmann, K.; Srouji, R. G.; Hansen, S. O.

2017-12-01

The technology development within the structural design of long-span bridges in Norwegian fjords has created a need for reformulating the calculation format and the physical quantities used to describe the properties of wind and the associated wind-induced effects on bridge decks. Parts of a new probabilistic format describing the incoming, undisturbed wind is presented. It is expected that a fixed probabilistic format will facilitate a more physically consistent and precise description of the wind conditions, which in turn increase the accuracy and considerably reduce uncertainties in wind load assessments. Because the format is probabilistic, a quantification of the level of safety and uncertainty in predicted wind loads is readily accessible. A simple buffeting response calculation demonstrates the use of probabilistic wind data in the assessment of wind loads and responses. Furthermore, vortex-induced fatigue damage is discussed in relation to probabilistic wind turbulence data and response measurements from wind tunnel tests.

Method and system for dynamic probabilistic risk assessment

Science.gov (United States)

Dugan, Joanne Bechta (Inventor); Xu, Hong (Inventor)

2013-01-01

The DEFT methodology, system and computer readable medium extends the applicability of the PRA (Probabilistic Risk Assessment) methodology to computer-based systems, by allowing DFT (Dynamic Fault Tree) nodes as pivot nodes in the Event Tree (ET) model. DEFT includes a mathematical model and solution algorithm, supports all common PRA analysis functions and cutsets. Additional capabilities enabled by the DFT include modularization, phased mission analysis, sequence dependencies, and imperfect coverage.

Application of probabilistic risk assessment to reprocessing

International Nuclear Information System (INIS)

Perkins, W.C.

1984-01-01

The Savannah River Laboratory uses probabilistic methods of risk assessment in safety analyses of reprocessing facilities at the Savannah River Plant. This method uses both the probability of an accident and its consequence to calculate the risks from radiological, chemical, and industrial hazards. The three principal steps in such an assesment are identification of accidents, calculation of frequencies, and consequence quantification. The tools used at SRL include several databanks, logic tree methods, and computer-assisted methods for calculating both frequencies and consequences. 5 figures

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)

Quantification of Wave Model Uncertainties Used for Probabilistic Reliability Assessments of Wave Energy Converters

DEFF Research Database (Denmark)

Ambühl, Simon; Kofoed, Jens Peter; Sørensen, John Dalsgaard

2015-01-01

Wave models used for site assessments are subjected to model uncertainties, which need to be quantified when using wave model results for probabilistic reliability assessments. This paper focuses on determination of wave model uncertainties. Four different wave models are considered, and validation...... data are collected from published scientific research. The bias and the root-mean-square error, as well as the scatter index, are considered for the significant wave height as well as the mean zero-crossing wave period. Based on an illustrative generic example, this paper presents how the quantified...... uncertainties can be implemented in probabilistic reliability assessments....

Determination of Wave Model Uncertainties used for Probabilistic Reliability Assessments of Wave Energy Devices

DEFF Research Database (Denmark)

Ambühl, Simon; Kofoed, Jens Peter; Sørensen, John Dalsgaard

2014-01-01

Wave models used for site assessments are subject to model uncertainties, which need to be quantified when using wave model results for probabilistic reliability assessments. This paper focuses on determination of wave model uncertainties. Considered are four different wave models and validation...... data is collected from published scientific research. The bias, the root-mean-square error as well as the scatter index are considered for the significant wave height as well as the mean zero-crossing wave period. Based on an illustrative generic example it is shown how the estimated uncertainties can...... be implemented in probabilistic reliability assessments....

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)

A reconnaissance assessment of probabilistic earthquake accelerations at the Nevada Test Site

International Nuclear Information System (INIS)

Perkins, D.M.; Thenhaus, P.C.; Hanson, S.L.; Algermissen, S.T.

1986-01-01

We have made two interim assessments of the probabilistic ground-motion hazard for the potential nuclear-waste disposal facility at the Nevada Test Site (NTS). The first assessment used historical seismicity and generalized source zones and source faults in the immediate vicinity of the facility. This model produced relatively high probabilistic ground motions, comparable to the higher of two earlier estimates, which was obtained by averaging seismicity in a 400-km-radius circle around the site. The high ground-motion values appear to be caused in part by nuclear-explosion aftershocks remaining in the catalog even after the explosions themselves have been removed. The second assessment used particularized source zones and source faults in a region substantially larger than NTS to provide a broad context of probabilistic ground motion estimates at other locations of the study region. Source faults are mapped or inferred faults having lengths of 5 km or more. Source zones are defined by boundaries separating fault groups on the basis of direction and density. For this assessment, earthquake recurrence has been estimated primarily from historic seismicity prior to nuclear testing. Long-term recurrence for large-magnitude events is constrained by geological estimates of recurrence in a regime in which the large-magnitude earthquakes would occur with predominately normal mechanisms. 4 refs., 10 figs

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)

Performing Probabilistic Risk Assessment Through RAVEN

Energy Technology Data Exchange (ETDEWEB)

A. Alfonsi; C. Rabiti; D. Mandelli; J. Cogliati; R. Kinoshita

2013-06-01

The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data mining module

Multi-Hazard Advanced Seismic Probabilistic Risk Assessment Tools and Applications

International Nuclear Information System (INIS)

Coleman, Justin L.; Bolisetti, Chandu; Veeraraghavan, Swetha; Parisi, Carlo; Prescott, Steven R.; Gupta, Abhinav

2016-01-01

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States (US), but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission (NRC) guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. This framework will continue to evolve as a result of improved knowledge and newly identified regulatory needs and objectives, most notably in response to the NRC activities developed in response to the 2011 Fukushima accident in Japan. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is significant room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. This is particularly true if PRA is applied to natural hazards other than seismic loading. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the SSCs and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms.

Multi-Hazard Advanced Seismic Probabilistic Risk Assessment Tools and Applications

Energy Technology Data Exchange (ETDEWEB)

Coleman, Justin L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolisetti, Chandu [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gupta, Abhinav [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States (US), but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission (NRC) guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. This framework will continue to evolve as a result of improved knowledge and newly identified regulatory needs and objectives, most notably in response to the NRC activities developed in response to the 2011 Fukushima accident in Japan. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is significant room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. This is particularly true if PRA is applied to natural hazards other than seismic loading. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the SSCs and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms.

A probabilistic approach to Radiological Environmental Impact Assessment

International Nuclear Information System (INIS)

Avila, Rodolfo; Larsson, Carl-Magnus

2001-01-01

Since a radiological environmental impact assessment typically relies on limited data and poorly based extrapolation methods, point estimations, as implied by a deterministic approach, do not suffice. To be of practical use for risk management, it is necessary to quantify the uncertainty margins of the estimates as well. In this paper we discuss how to work out a probabilistic approach for dealing with uncertainties in assessments of the radiological risks to non-human biota of a radioactive contamination. Possible strategies for deriving the relevant probability distribution functions from available empirical data and theoretical knowledge are outlined

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

Exploration Health Risks: Probabilistic Risk Assessment

Science.gov (United States)

Rhatigan, Jennifer; Charles, John; Hayes, Judith; Wren, Kiley

2006-01-01

Maintenance of human health on long-duration exploration missions is a primary challenge to mission designers. Indeed, human health risks are currently the largest risk contributors to the risks of evacuation or loss of the crew on long-duration International Space Station missions. We describe a quantitative assessment of the relative probabilities of occurrence of the individual risks to human safety and efficiency during space flight to augment qualitative assessments used in this field to date. Quantitative probabilistic risk assessments will allow program managers to focus resources on those human health risks most likely to occur with undesirable consequences. Truly quantitative assessments are common, even expected, in the engineering and actuarial spheres, but that capability is just emerging in some arenas of life sciences research, such as identifying and minimize the hazards to astronauts during future space exploration missions. Our expectation is that these results can be used to inform NASA mission design trade studies in the near future with the objective of preventing the higher among the human health risks. We identify and discuss statistical techniques to provide this risk quantification based on relevant sets of astronaut biomedical data from short and long duration space flights as well as relevant analog populations. We outline critical assumptions made in the calculations and discuss the rationale for these. Our efforts to date have focussed on quantifying the probabilities of medical risks that are qualitatively perceived as relatively high risks of radiation sickness, cardiac dysrhythmias, medically significant renal stone formation due to increased calcium mobilization, decompression sickness as a result of EVA (extravehicular activity), and bone fracture due to loss of bone mineral density. We present these quantitative probabilities in order-of-magnitude comparison format so that relative risk can be gauged. We address the effects of

Human reliability. Is probabilistic human reliability assessment possible?

International Nuclear Information System (INIS)

Mosneron Dupin, F.

1996-01-01

The possibility of carrying out Probabilistic Human Reliability Assessments (PHRA) is often doubted. Basing ourselves on the experience Electricite de France (EDF) has acquired in Probabilistic Safety Assessments for nuclear power plants, we show why the uncertainty of PHRA is very high. We then specify the limits of generic data and models for PHRA: very important factors are often poorly taken into account. To account for them, you need to have proper understanding of the actual context in which operators work. This demands surveys on the field (power plant and simulator) all of which must be carried out with behaviours science skills. The idea of estimating the probabilities of operator failure must not be abandoned, but probabilities must be given less importance, for they are only approximate indications. The qualitative aspects of PHRA should be given greater value (analysis process and qualitative insights). That is why the description (illustrated by case histories) of the main mechanisms of human behaviour, and of their manifestations in the nuclear power plant context (in terms of habits, attitudes, and informal methods and organization in particular) should be an important part of PHRA handbooks. These handbooks should also insist more on methods for gathering information on the actual context of the work of operators. Under these conditions, the PHRA should be possible and even desirable as a process for systematic analysis and assessment of human intervention. (author). 24 refs, 2 figs, 1 tab

Human reliability analysis methods for probabilistic safety assessment

International Nuclear Information System (INIS)

Pyy, P.

2000-11-01

Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

Human reliability. Is probabilistic human reliability assessment possible?

Energy Technology Data Exchange (ETDEWEB)

Mosneron Dupin, F

1997-12-31

The possibility of carrying out Probabilistic Human Reliability Assessments (PHRA) is often doubted. Basing ourselves on the experience Electricite de France (EDF) has acquired in Probabilistic Safety Assessments for nuclear power plants, we show why the uncertainty of PHRA is very high. We then specify the limits of generic data and models for PHRA: very important factors are often poorly taken into account. To account for them, you need to have proper understanding of the actual context in which operators work. This demands surveys on the field (power plant and simulator) all of which must be carried out with behaviours science skills. The idea of estimating the probabilities of operator failure must not be abandoned, but probabilities must be given less importance, for they are only approximate indications. The qualitative aspects of PHRA should be given greater value (analysis process and qualitative insights). That is why the description (illustrated by case histories) of the main mechanisms of human behaviour, and of their manifestations in the nuclear power plant context (in terms of habits, attitudes, and informal methods and organization in particular) should be an important part of PHRA handbooks. These handbooks should also insist more on methods for gathering information on the actual context of the work of operators. Under these conditions, the PHRA should be possible and even desirable as a process for systematic analysis and assessment of human intervention. (author). 24 refs, 2 figs, 1 tab.

A Probabilistic Assessment of the Next Geomagnetic Reversal

OpenAIRE

Buffett, B; Davis, W

2018-01-01

©2018. American Geophysical Union. All Rights Reserved. Deterministic forecasts for the next geomagnetic reversal are not feasible due to large uncertainties in the present-day state of the Earth's core. A more practical approach relies on probabilistic assessments using paleomagnetic observations to characterize the amplitude of fluctuations in the geomagnetic dipole. We use paleomagnetic observations for the past 2 Myr to construct a stochastic model for the axial dipole field and apply wel...

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 cumulative risk assessment of anti-androgenic pesticides in food

DEFF Research Database (Denmark)

Müller, Anne Kirstine; Bosgra, Sieto; Boon, Polly E.

2009-01-01

In this paper, we present a cumulative risk assessment of three anti-androgenic pesticides (vinclozolin, procymidone and prochloraz) using the relative potency factor (RPF) approach and an integrated probabilistic risk assessment (IPRA) model. RPFs for each substance were estimated for three......) and the fraction of individuals with IMoEs vinclozolin, procymidone and prochloraz is not likely to be of concern for the reproductive development of their male foetuses. However...

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

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 Risk Assessment Procedures Guide for NASA Managers and Practitioners (Second Edition)

Science.gov (United States)

Stamatelatos,Michael; Dezfuli, Homayoon; Apostolakis, George; Everline, Chester; Guarro, Sergio; Mathias, Donovan; Mosleh, Ali; Paulos, Todd; Riha, David; Smith, Curtis;

Architecture for Integrated Medical Model Dynamic Probabilistic Risk Assessment

Science.gov (United States)

Jaworske, D. A.; Myers, J. G.; Goodenow, D.; Young, M.; Arellano, J. D.

2016-01-01

Probabilistic Risk Assessment (PRA) is a modeling tool used to predict potential outcomes of a complex system based on a statistical understanding of many initiating events. Utilizing a Monte Carlo method, thousands of instances of the model are considered and outcomes are collected. PRA is considered static, utilizing probabilities alone to calculate outcomes. Dynamic Probabilistic Risk Assessment (dPRA) is an advanced concept where modeling predicts the outcomes of a complex system based not only on the probabilities of many initiating events, but also on a progression of dependencies brought about by progressing down a time line. Events are placed in a single time line, adding each event to a queue, as managed by a planner. Progression down the time line is guided by rules, as managed by a scheduler. The recently developed Integrated Medical Model (IMM) summarizes astronaut health as governed by the probabilities of medical events and mitigation strategies. Managing the software architecture process provides a systematic means of creating, documenting, and communicating a software design early in the development process. The software architecture process begins with establishing requirements and the design is then derived from the requirements.

Probabilistic Assessment of the Occurrence and Duration of Ice Accretion on Cables

DEFF Research Database (Denmark)

Roldsgaard, Joan Hee; Georgakis, Christos Thomas; Faber, Michael Havbro

2015-01-01

This paper presents an operational framework for assessing the probability of occurrence of in-cloud and precipitation icing and its duration. The framework utilizes the features of the Bayesian Probabilistic Networks. and its performance is illustrated through a case study of the cable-stayed...... Oresund Bridge. The Bayesian Probabilistic Network model used for the estimation of the occurrence and duration probabilities is studied and it is found to be robust with respect to changes in the choice of distribution types used to model the meteorological variables that influence the two icing...

Assessing dynamic postural control during exergaming in older adults: A probabilistic approach.

Science.gov (United States)

Soancatl Aguilar, V; Lamoth, C J C; Maurits, N M; Roerdink, J B T M

2018-02-01

Digital games controlled by body movements (exergames) have been proposed as a way to improve postural control among older adults. Exergames are meant to be played at home in an unsupervised way. However, only few studies have investigated the effect of unsupervised home-exergaming on postural control. Moreover, suitable methods to dynamically assess postural control during exergaming are still scarce. Dynamic postural control (DPC) assessment could be used to provide both meaningful feedback and automatic adjustment of exergame difficulty. These features could potentially foster unsupervised exergaming at home and improve the effectiveness of exergames as tools to improve balance control. The main aim of this study is to investigate the effect of six weeks of unsupervised home-exergaming on DPC as assessed by a recently developed probabilistic model. High probability values suggest 'deteriorated' postural control, whereas low probability values suggest 'good' postural control. In a pilot study, ten healthy older adults (average 77.9, SD 7.2 years) played an ice-skating exergame at home half an hour per day, three times a week during six weeks. The intervention effect on DPC was assessed using exergaming trials recorded by Kinect at baseline and every other week. Visualization of the results suggests that the probabilistic model is suitable for real-time DPC assessment. Moreover, linear mixed model analysis and parametric bootstrapping suggest a significant intervention effect on DPC. In conclusion, these results suggest that unsupervised exergaming for improving DPC among older adults is indeed feasible and that probabilistic models could be a new approach to assess DPC. Copyright © 2017 Elsevier B.V. All rights reserved.

Uncertainty analysis on probabilistic fracture mechanics assessment methodology

International Nuclear Information System (INIS)

Rastogi, Rohit; Vinod, Gopika; Chandra, Vikas; Bhasin, Vivek; Babar, A.K.; Rao, V.V.S.S.; Vaze, K.K.; Kushwaha, H.S.; Venkat-Raj, V.

1999-01-01

Fracture Mechanics has found a profound usage in the area of design of components and assessing fitness for purpose/residual life estimation of an operating component. Since defect size and material properties are statistically distributed, various probabilistic approaches have been employed for the computation of fracture probability. Monte Carlo Simulation is one such procedure towards the analysis of fracture probability. This paper deals with uncertainty analysis using the Monte Carlo Simulation methods. These methods were developed based on the R6 failure assessment procedure, which has been widely used in analysing the integrity of structures. The application of this method is illustrated with a case study. (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

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-09-01

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

Probabilistic Multi-Hazard Assessment of Dry Cask Structures

Energy Technology Data Exchange (ETDEWEB)

Bencturk, Bora [Univ. of Houston, TX (United States); Padgett, Jamie [Rice Univ., Houston, TX (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States).

2017-01-10

systems the concrete shall not only provide shielding but insures stability of the upright canister, facilitates anchoring, allows ventilation, and provides physical protection against theft, severe weather and natural (seismic) as well as man-made events (blast incidences). Given the need to remain functional for 40 years or even longer in case of interim storage, the concrete outerpack and the internal canister components need to be evaluated with regard to their long-term ability to perform their intended design functions. Just as evidenced by deteriorating concrete bridges, there are reported visible degradation mechanisms of dry storage systems especially when high corrosive environments are considered in maritime locations. The degradation of reinforced concrete is caused by multiple physical and chemical mechanisms, which may be summarized under the heading of environmental aging. The underlying hygro-thermal transport processes are accelerated by irradiation effects, hence creep and shrinkage need to include the effect of chloride penetration, alkali aggregate reaction as well as corrosion of the reinforcing steel. In light of the above, the two main objectives of this project are to (1) develop a probabilistic multi-hazard assessment framework, and (2) through experimental and numerical research perform a comprehensive assessment under combined earthquake loads and aging induced deterioration, which will also provide data for the development and validation of the probabilistic framework.

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

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

Psacoin level 1A intercomparison probabilistic system assessment code (PSAC) user group

International Nuclear Information System (INIS)

Nies, A.; Laurens, J.M.; Galson, D.A.; Webster, S.

1990-01-01

This report describes an international code intercomparison exercise conducted by the NEA Probabilistic System Assessment Code (PSAC) User Group. The PSACOIN Level 1A exercise is the third of a series designed to contribute to the verification of probabilistic codes that may be used in assessing the safety of radioactive waste disposal systems or concepts. Level 1A is based on a more realistic system model than that used in the two previous exercises, and involves deep geological disposal concepts with a relatively complex structure of the repository vault. The report compares results and draws conclusions with regard to the use of different modelling approaches and the possible importance to safety of various processes within and around a deep geological repository. In particular, the relative significance of model uncertainty and data variability is discussed

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)

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 assessment of leak-before-break

International Nuclear Information System (INIS)

Bush, S.H.

1984-01-01

A summary of results illustrating what might be derived from a probabilistic risk assessment (PRA) study follows. The failure probabilities for larger sizes of nuclear piping are considered to be in the range of 10 -4 to 10 -6 per reactor-year (exclusive of intergranular stress corrosion cracking (IGSCC). Smaller pipe sizes, of lesser safety significance, have much higher failure rates. In BWRs, IGSCC can cause failure rates much higher than 10 -4 in piping 4 to 10 in. in size. Suggested failure mechanisms apply in most instances, exclusive of IGSCC. Catastrophic failures would appear more likely from operator error or design and construction errors (water hammer, improper handling of dynamic loads, and undetected fabrication defects) rather than conventional flaw initiation and growth by fatigue

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Probabilistic seismic hazard assessment for Point Lepreau Generating Station

Energy Technology Data Exchange (ETDEWEB)

Mullin, D. [New Brunswick Power Corp., Point Lepreau Generating Station, Lepreau, New Brunswick (Canada); Lavine, A. [AMEC Foster Wheeler Environment and Infrastructure Americas, Oakland, California (United States); Egan, J. [SAGE Engineers, Oakland, California (United States)

2015-09-15

A Probabilistic Seismic Hazard Assessment (PSHA) has been performed for the Point Lepreau Generating Station (PLGS). The objective is to provide characterization of the earthquake ground shaking that will be used to evaluate seismic safety. The assessment is based on the current state of knowledge of the informed scientific and engineering community regarding earthquake hazards in the site region, and includes two primary components-a seismic source model and a ground motion model. This paper provides the methodology and results of the PLGS PSHA. The implications of the updated hazard information for site safety are discussed in a separate paper. (author)

Probabilistic seismic hazard assessment for Point Lepreau Generating Station

Energy Technology Data Exchange (ETDEWEB)

Mullin, D., E-mail: dmullin@nbpower.com [New Brunswick Power Corporation, Point Lepreau Generating Station, Point Lepreau, NB (Canada); Lavine, A., E-mail: alexis.lavine@amecfw.com [AMEC Foster Wheeler Environment & Infrastructure Americas, Oakland, CA (United States); Egan, J., E-mail: jegan@sageengineers.com [SAGE Engineers, Oakland, CA (United States)

2015-07-01

A Probabilistic Seismic Hazard Assessment (PSHA) has been performed for the Point Lepreau Generating Station (PLGS). The objective is to provide characterization of the earthquake ground shaking that will be used to evaluate seismic safety. The assessment is based on the current state of knowledge of the informed scientific and engineering community regarding earthquake hazards in the site region, and includes two primary components--a seismic source model and a ground motion model. This paper provides the methodology and results of the PLGS PSHA. The implications of the updated hazard information for site safety are discussed in a separate paper. (author)

Evaluation of nuclear power plant siting by probabilistic assessment of environmental impacts

International Nuclear Information System (INIS)

Vuori, S.

1978-01-01

The work consists of the description of a probabilistic consequence assessment model ARANO, and the individual calculation schemes therein included. This assessment model has been applied to the risk/benefit and cost/benefit analyses of the siting of nuclear power plants. In addition, there have been made some comparisons with the alternative fossil fuelled energy production scenarios. (author)

Probabilistic seismic hazard assessment. Gentilly 2

International Nuclear Information System (INIS)

1996-03-01

Results of this probabilistic seismic hazard assessment were determined using a suite of conservative assumptions. The intent of this study was to perform a limited hazard assessment that incorporated a range of technically defensible input parameters. To best achieve this goal, input selected for the hazard assessment tended to be conservative with respect to selection of attenuation modes, and seismicity parameters. Seismic hazard estimates at Gentilly 2 were most affected by selection of the attenuation model. Alternative definitions of seismic source zones had a relatively small impact on seismic hazard. A St. Lawrence Rift model including a maximum magnitude of 7.2 m b in the zone containing the site had little effect on the hazard estimate relative to other seismic source zonation models. Mean annual probabilities of exceeding the design peak ground acceleration, and the design response spectrum for the Gentilly 2 site were computed to lie in the range of 0.001 to 0.0001. This hazard result falls well within the range determined to be acceptable for nuclear reactor sites located throughout the eastern United States. (author) 34 refs., 6 tabs., 28 figs

Assessing performance and validating finite element simulations using probabilistic knowledge

Energy Technology Data Exchange (ETDEWEB)

Dolin, Ronald M.; Rodriguez, E. A. (Edward A.)

2002-01-01

Two probabilistic approaches for assessing performance are presented. The first approach assesses probability of failure by simultaneously modeling all likely events. The probability each event causes failure along with the event's likelihood of occurrence contribute to the overall probability of failure. The second assessment method is based on stochastic sampling using an influence diagram. Latin-hypercube sampling is used to stochastically assess events. The overall probability of failure is taken as the maximum probability of failure of all the events. The Likelihood of Occurrence simulation suggests failure does not occur while the Stochastic Sampling approach predicts failure. The Likelihood of Occurrence results are used to validate finite element predictions.

Bayesian parameter estimation in probabilistic risk assessment

International Nuclear Information System (INIS)

Siu, Nathan O.; Kelly, Dana L.

1998-01-01

Bayesian statistical methods are widely used in probabilistic risk assessment (PRA) because of their ability to provide useful estimates of model parameters when data are sparse and because the subjective probability framework, from which these methods are derived, is a natural framework to address the decision problems motivating PRA. This paper presents a tutorial on Bayesian parameter estimation especially relevant to PRA. It summarizes the philosophy behind these methods, approaches for constructing likelihood functions and prior distributions, some simple but realistic examples, and a variety of cautions and lessons regarding practical applications. References are also provided for more in-depth coverage of various topics

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

Probabilistic assessments of climate change impacts on durum wheat in the Mediterranean region

Directory of Open Access Journals (Sweden)

R. Ferrise

2011-05-01

Full Text Available Recently, the availability of multi-model ensemble prediction methods has permitted a shift from a scenario-based approach to a risk-based approach in assessing the effects of climate change. This provides more useful information to decision-makers who need probability estimates to assess the seriousness of the projected impacts.

In this study, a probabilistic framework for evaluating the risk of durum wheat yield shortfall over the Mediterranean Basin has been exploited. An artificial neural network, trained to emulate the outputs of a process-based crop growth model, has been adopted to create yield response surfaces which are then overlaid with probabilistic projections of future temperature and precipitation changes in order to estimate probabilistic projections of future yields. The risk is calculated as the relative frequency of projected yields below a selected threshold.

In contrast to previous studies, which suggest that the beneficial effects of elevated atmospheric CO₂ concentration over the next few decades would outweigh the detrimental effects of the early stages of climatic warming and drying, the results of this study are of greater concern.

Probabilistic Seismic Hazard Assessment for Iraq

Energy Technology Data Exchange (ETDEWEB)

Onur, Tuna [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Gok, Rengin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Abdulnaby, Wathiq [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shakir, Ammar M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mahdi, Hanan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Numan, Nazar M.S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Al-Shukri, Haydar [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chlaib, Hussein K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ameen, Taher H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Abd, Najah A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-05-06

Probabilistic Seismic Hazard Assessments (PSHA) form the basis for most contemporary seismic provisions in building codes around the world. The current building code of Iraq was published in 1997. An update to this edition is in the process of being released. However, there are no national PSHA studies in Iraq for the new building code to refer to for seismic loading in terms of spectral accelerations. As an interim solution, the new draft building code was considering to refer to PSHA results produced in the late 1990s as part of the Global Seismic Hazard Assessment Program (GSHAP; Giardini et al., 1999). However these results are: a) more than 15 years outdated, b) PGA-based only, necessitating rough conversion factors to calculate spectral accelerations at 0.3s and 1.0s for seismic design, and c) at a probability level of 10% chance of exceedance in 50 years, not the 2% that the building code requires. Hence there is a pressing need for a new, updated PSHA for Iraq.

Assessment and presentation of uncertainties in probabilistic risk assessment: how should this be done

International Nuclear Information System (INIS)

Garlick, A.R.; Holloway, N.J.

1987-01-01

Despite continuing improvements in probabilistic risk assessment (PRA) techniques, PRA results, particularly those including degraded core analysis, will have maximum uncertainties of several orders of magnitude. This makes the expression of results, a matter no less important than their estimation. We put forward some ideas on the assessment and expression of highly uncertain quantities, such as probabilities of outcomes of a severe accident. These do not form a consistent set, but rather a number of alternative approaches aimed at stimulating discussion. These include non-probability expressions, such as fuzzy logic or Schafer's support and plausibility which abandon the purely probabilistic expression of risk for a more flexible type of expression, in which other types of measure are possible. The 'risk equivalent plant' concepts represent the opposite approach. Since uncertainty in a risk measure is in itself a form of risk, an attempt is made to define a 'risk equivalent' which is a risk with perfectly defined parameters, regarded (by means of suitable methods of judgement) as 'equally undesirable' with the actual plant. Some guidelines are given on the use of Bayesian methods in data-free or limited data situations. (author)

77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

Science.gov (United States)

2012-10-09

... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

Track finding and track reconstruction in the internal forward drift chamber of SAPHIR

International Nuclear Information System (INIS)

Umlauf, G.

1993-03-01

A track finding algorithm has been developed for the inner forward drift chamber of the SAPHIR detector (at ELSA in Bonn) using the Principal Components Analysis as a tool for interpolating track coordinates. The drift chamber consists of twelve planar layers with six different inclinations and is being operated in an inhomogenous magnetic field. The task of track finding is basicly split into a primary stage that defines track candidates without the use of drift-time information and a second stage that serves to verify the track candidate and to resolve the intrinsic left-right ambiguities of the drift chamber signals. Tracks with at most three missing signals can be found. (orig.) [de

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)

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

Transmission capacity assessment by probabilistic planning. An approach

International Nuclear Information System (INIS)

Lammintausta, M.

2002-01-01

The Finnish electricity markets participate in the Scandinavian markets, Nord-Pool. The Finnish market is free for marketers, producers and consumers. All these participants can be seen as customers of the transmission network, which in turn can be considered to be a market place in which electricity can be sold and bought. The Finnish transmission network is owned and operated by an independent company, Fingrid that has the full responsibility of the Finnish transmission system. The available transfer capacity of a transmission route is traditionally limited by deterministic security constraints. More efficient and flexible network utilisation could be achieved with probabilistic planning methods. This report introduces a simple and practical probabilistic approach for transfer limit and risk assessment. The method is based on the economical benefit and risk predictions. It uses also the existing results of deterministic data and it could be used side by side with the deterministic method. The basic concept and necessary equations for expected risks of various market players have been derived for further developments. The outage costs and thereby the risks of the market participants depend on how the system operator reacts to the faults. In the Finnish power system consumers will usually experience no costs due to the faults because of meshed network and counter trade method preferred by the system operator. The costs to the producers and dealers are also low because of the counter trade method. The network company will lose the cost of reparation, additional losses and cost of regulation power because of counter trades. In case power flows will be rearranged drastically because of aggressive strategies used in the electricity markets, the only way to fulfil the needs of free markets is that the network operator buys regulation power for short-term problems and reinforces the network in the long-term situations. The reinforcement is done if the network can not be

Particle identification by time-of-flight measurement in the SAPHIR

International Nuclear Information System (INIS)

Hoffmann-Rothe, P.

1993-02-01

Using photoproduction data which have been measured with the SAPHIR-detector with different target materials (C H 2 solid , H 2 liquid , D 2 liquid ) a detailed investigation and discussion of the detectors performance to measure the time of flight of charged particles and to separate between particles of different mass has been accomplished. A FORTRAN program has been written which provides a calibration of the scintillator panels of the TOF hodoscopes, calculates correction factors for the time-walk effect an finally, by combining the time of flight with track momentum measurement, determines particle masses. The current configuration of the detector makes it possible to separate between proton and pion up to a particle momentum of 1.6 GeV/c. Proton and kaon can be separated up to a momentum of 1.3 GeV/c, kaon and pion up to a momentum of 0.85 GeV/c. (prog.) [de

Probabilistic studies of accident sequences

International Nuclear Information System (INIS)

Villemeur, A.; Berger, J.P.

1986-01-01

For several years, Electricite de France has carried out probabilistic assessment of accident sequences for nuclear power plants. In the framework of this program many methods were developed. As the interest in these studies was increasing and as adapted methods were developed, Electricite de France has undertaken a probabilistic safety assessment of a nuclear power plant [fr

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

Developing and evaluating distributions for probabilistic human exposure assessments

Energy Technology Data Exchange (ETDEWEB)

Maddalena, Randy L.; McKone, Thomas E.

2002-08-01

This report describes research carried out at the Lawrence Berkeley National Laboratory (LBNL) to assist the U. S. Environmental Protection Agency (EPA) in developing a consistent yet flexible approach for evaluating the inputs to probabilistic risk assessments. The U.S. EPA Office of Emergency and Remedial Response (OERR) recently released Volume 3 Part A of Risk Assessment Guidance for Superfund (RAGS), as an update to the existing two-volume set of RAGS. The update provides policy and technical guidance on performing probabilistic risk assessment (PRA). Consequently, EPA risk managers and decision-makers need to review and evaluate the adequacy of PRAs for supporting regulatory decisions. A critical part of evaluating a PRA is the problem of evaluating or judging the adequacy of input distributions PRA. Although the overarching theme of this report is the need to improve the ease and consistency of the regulatory review process, the specific objectives are presented in two parts. The objective of Part 1 is to develop a consistent yet flexible process for evaluating distributions in a PRA by identifying the critical attributes of an exposure factor distribution and discussing how these attributes relate to the task-specific adequacy of the input. This objective is carried out with emphasis on the perspective of a risk manager or decision-maker. The proposed evaluation procedure provides consistency to the review process without a loss of flexibility. As a result, the approach described in Part 1 provides an opportunity to apply a single review framework for all EPA regions and yet provide the regional risk manager with the flexibility to deal with site- and case-specific issues in the PRA process. However, as the number of inputs to a PRA increases, so does the complexity of the process for calculating, communicating and managing risk. As a result, there is increasing effort required of both the risk professionals performing the analysis and the risk manager

Uncertainty propagation in probabilistic risk assessment: A comparative study

International Nuclear Information System (INIS)

Ahmed, S.; Metcalf, D.R.; Pegram, J.W.

1982-01-01

Three uncertainty propagation techniques, namely method of moments, discrete probability distribution (DPD), and Monte Carlo simulation, generally used in probabilistic risk assessment, are compared and conclusions drawn in terms of the accuracy of the results. For small uncertainty in the basic event unavailabilities, the three methods give similar results. For large uncertainty, the method of moments is in error, and the appropriate method is to propagate uncertainty in the discrete form either by DPD method without sampling or by Monte Carlo. (orig.)

Insights from Guideline for Performance of Internal Flooding Probabilistic Risk Assessment (IFPRA)

International Nuclear Information System (INIS)

Choi, Sun Yeong; Yang, Joo Eon

2009-01-01

An internal flooding (IF) risk assessment refers to the quantitative probabilistic safety assessment (PSA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (ASME-RA-Sb-2005) includes high-level and supporting technical requirements for developing internal flooding probabilistic risk assessment (IFPRA). This industry standard is endorsed in Regulatory Guide 1.200, Revision 1 as an acceptable approach for addressing the risk contribution from IF events for risk informed applications that require U.S. Nuclear Regulatory commission (NRC) approval. In 2006, EPRI published a draft report for IFPRA that addresses the requirements of the ASME PRA consensus standard and have made efforts to refine and update the final EPRI IFPRA guideline. Westinghouse has performed an IFPRA analysis for several nuclear power plants (NPPs), such as Watts Bar and Fort Calhoun, using the draft EPRI guidelines for development of an IFPRA. Proprietary methodologies have been developed to apply the EPRI guidelines. The objectives of the draft report for IFPRA guideline are to: · Provide guidance for PSA practitioners in the performance of the elements of a PRA associated with internal flooding events consistent with the current state of the art for internal flooding PRA · Provide guidance regarding acceptable approaches that is sufficient to meeting the requirements of the ASME PRA Standard associated with internal flooding · Incorporate lessons learned in the performance of internal flooding PRAs including those identified as pilot applications of earlier drafts of this procedures guide The purpose of this paper is to present a vision for domestic nuclear power plants' IFPRA by comparing the method of the draft EPRI guidelines with the existing IFPRA method for domestic NPPs

Insights from Guideline for Performance of Internal Flooding Probabilistic Risk Assessment (IFPRA)

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Yeong; Yang, Joo Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-10-15

An internal flooding (IF) risk assessment refers to the quantitative probabilistic safety assessment (PSA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (ASME-RA-Sb-2005) includes high-level and supporting technical requirements for developing internal flooding probabilistic risk assessment (IFPRA). This industry standard is endorsed in Regulatory Guide 1.200, Revision 1 as an acceptable approach for addressing the risk contribution from IF events for risk informed applications that require U.S. Nuclear Regulatory commission (NRC) approval. In 2006, EPRI published a draft report for IFPRA that addresses the requirements of the ASME PRA consensus standard and have made efforts to refine and update the final EPRI IFPRA guideline. Westinghouse has performed an IFPRA analysis for several nuclear power plants (NPPs), such as Watts Bar and Fort Calhoun, using the draft EPRI guidelines for development of an IFPRA. Proprietary methodologies have been developed to apply the EPRI guidelines. The objectives of the draft report for IFPRA guideline are to: {center_dot} Provide guidance for PSA practitioners in the performance of the elements of a PRA associated with internal flooding events consistent with the current state of the art for internal flooding PRA {center_dot} Provide guidance regarding acceptable approaches that is sufficient to meeting the requirements of the ASME PRA Standard associated with internal flooding {center_dot} Incorporate lessons learned in the performance of internal flooding PRAs including those identified as pilot applications of earlier drafts of this procedures guide The purpose of this paper is to present a vision for domestic nuclear power plants' IFPRA by comparing the method of the draft EPRI guidelines with the existing IFPRA method for domestic NPPs.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-02-15

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

Comparison of seismic margin assessment and probabilistic risk assessment in seismic IPE

International Nuclear Information System (INIS)

Reed, J.W.; Kassawara, R.P.

1993-01-01

A comparison of technical requirements and managerial issues between seismic margin assessment (SMA) and seismic probabilistic risk assessment (SPRA) in a seismic Individual Plant Examination (IPE) is presented and related to requirements for an Unresolved Safety Issue (USI) A-46 review which is required for older nuclear power plants. Advantages and disadvantages are discussed for each approach. Technical requirements reviewed for a seismic IPE include: scope of plants covered, seismic input, scope of review, selection of equipment, required experience and training of engineers, walkdown procedure, evaluation of components, relay review, containment review, quality assurance, products, documentation requirements, and closure procedure. Managerial issues discussed include regulatory acceptability, compatibility with seismic IPE, compliance with seismic IPE requirements, ease of use by utilities, and relative cost

Structural Analysis of PTM Hotspots (SAPH-ire) – A Quantitative Informatics Method Enabling the Discovery of Novel Regulatory Elements in Protein Families*

Science.gov (United States)

Dewhurst, Henry M.; Choudhury, Shilpa; Torres, Matthew P.

2015-01-01

Predicting the biological function potential of post-translational modifications (PTMs) is becoming increasingly important in light of the exponential increase in available PTM data from high-throughput proteomics. We developed structural analysis of PTM hotspots (SAPH-ire)—a quantitative PTM ranking method that integrates experimental PTM observations, sequence conservation, protein structure, and interaction data to allow rank order comparisons within or between protein families. Here, we applied SAPH-ire to the study of PTMs in diverse G protein families, a conserved and ubiquitous class of proteins essential for maintenance of intracellular structure (tubulins) and signal transduction (large and small Ras-like G proteins). A total of 1728 experimentally verified PTMs from eight unique G protein families were clustered into 451 unique hotspots, 51 of which have a known and cited biological function or response. Using customized software, the hotspots were analyzed in the context of 598 unique protein structures. By comparing distributions of hotspots with known versus unknown function, we show that SAPH-ire analysis is predictive for PTM biological function. Notably, SAPH-ire revealed high-ranking hotspots for which a functional impact has not yet been determined, including phosphorylation hotspots in the N-terminal tails of G protein gamma subunits—conserved protein structures never before reported as regulators of G protein coupled receptor signaling. To validate this prediction we used the yeast model system for G protein coupled receptor signaling, revealing that gamma subunit–N-terminal tail phosphorylation is activated in response to G protein coupled receptor stimulation and regulates protein stability in vivo. These results demonstrate the utility of integrating protein structural and sequence features into PTM prioritization schemes that can improve the analysis and functional power of modification-specific proteomics data. PMID:26070665

Probabilistic record linkage.

Science.gov (United States)

Sayers, Adrian; Ben-Shlomo, Yoav; Blom, Ashley W; Steele, Fiona

2016-06-01

Studies involving the use of probabilistic record linkage are becoming increasingly common. However, the methods underpinning probabilistic record linkage are not widely taught or understood, and therefore these studies can appear to be a 'black box' research tool. In this article, we aim to describe the process of probabilistic record linkage through a simple exemplar. We first introduce the concept of deterministic linkage and contrast this with probabilistic linkage. We illustrate each step of the process using a simple exemplar and describe the data structure required to perform a probabilistic linkage. We describe the process of calculating and interpreting matched weights and how to convert matched weights into posterior probabilities of a match using Bayes theorem. We conclude this article with a brief discussion of some of the computational demands of record linkage, how you might assess the quality of your linkage algorithm, and how epidemiologists can maximize the value of their record-linked research using robust record linkage methods. © The Author 2015; Published by Oxford University Press on behalf of the International Epidemiological Association.

Site-specific probabilistic ecological risk assessment of a volatile chlorinated hydrocarbon-contaminated tidal estuary.

Science.gov (United States)

Hunt, James; Birch, Gavin; Warne, Michael St J

2010-05-01

Groundwater contaminated with volatile chlorinated hydrocarbons (VCHs) was identified as discharging to Penrhyn Estuary, an intertidal embayment of Botany Bay, New South Wales, Australia. A screening-level hazard assessment of surface water in Penrhyn Estuary identified an unacceptable hazard to marine organisms posed by VCHs. Given the limitations of hazard assessments, the present study conducted a higher-tier, quantitative probabilistic risk assessment using the joint probability curve (JPC) method that accounted for variability in exposure and toxicity profiles to quantify risk (delta). Risk was assessed for 24 scenarios, including four areas of the estuary based on three exposure scenarios (low tide, high tide, and both low and high tides) and two toxicity scenarios (chronic no-observed-effect concentrations [NOEC] and 50% effect concentrations [EC50]). Risk (delta) was greater at low tide than at high tide and varied throughout the tidal cycle. Spatial distributions of risk in the estuary were similar using both NOEC and EC50 data. The exposure scenario including data combined from both tides was considered the most accurate representation of the ecological risk in the estuary. When assessing risk using data across both tides, the greatest risk was identified in the Springvale tributary (delta=25%)-closest to the source area-followed by the inner estuary (delta=4%) and the Floodvale tributary (delta=2%), with the lowest risk in the outer estuary (delta=0.1%), farthest from the source area. Going from the screening level ecological risk assessment (ERA) to the probabilistic ERA changed the risk from unacceptable to acceptable in 50% of exposure scenarios in two of the four areas within the estuary. The probabilistic ERA provided a more realistic assessment of risk than the screening-level hazard assessment. Copyright (c) 2010 SETAC.

Integrated probabilistic risk assessment for nanoparticles: the case of nanosilica in food

OpenAIRE

Jacobs, R.; Voet, van der, H.; Braak, ter, C.J.F.

2015-01-01

Insight into risks of nanotechnology and the use of nanoparticles is an essential condition for the social acceptance and safe use of nanotechnology. One of the problems with which the risk assessment of nanoparticles is faced is the lack of data, resulting in uncertainty in the risk assessment. We attempt to quantify some of this uncertainty by expanding a previous deterministic study on nanosilica (5?200?nm) in food into a fully integrated probabilistic risk assessment. We use the integrate...

Dietary Exposure Assessment of Danish Consumers to Dithiocarbamate Residues in Food: a Comparison of the Deterministic and Probabilistic Approach

DEFF Research Database (Denmark)

Jensen, Bodil Hamborg; Andersen, Jens Hinge; Petersen, Annette

2008-01-01

Probabilistic and deterministic estimates of the acute and chronic exposure of the Danish populations to dithiocarbamate residues were performed. The Monte Carlo Risk Assessment programme (MCRA 4.0) was used for the probabilistic risk assessment. Food consumption data were obtained from...... the nationwide dietary survey conducted in 2000-02. Residue data for 5721 samples from the monitoring programme conducted in the period 1998-2003 were used for dithiocarbamates, which had been determined as carbon disulphide. Contributions from 26 commodities were included in the calculations. Using...... the probabilistic approach, the daily acute intakes at the 99.9% percentile for adults and children were 11.2 and 28.2 mu g kg(-1) body weight day(-1), representing 5.6% and 14.1% of the ARfD for maneb, respectively. When comparing the point estimate approach with the probabilistic approach, the outcome...

Dealing with uncertainty arising out of probabilistic risk assessment

International Nuclear Information System (INIS)

Solomon, K.A.; Kastenberg, W.E.; Nelson, P.F.

1984-03-01

In addressing the area of safety goal implementation, the question of uncertainty arises. This report suggests that the Nuclear Regulatory Commission (NRC) should examine how other regulatory organizations have addressed the issue. Several examples are given from the chemical industry, and comparisons are made to nuclear power risks. Recommendations are made as to various considerations that the NRC should require in probabilistic risk assessments in order to properly treat uncertainties in the implementation of the safety goal policy. 40 references, 7 figures, 5 tables

Probabilistic risk assessment of earthquakes at the Rocky Flats Plant and subsequent upgrade to reduce risk

International Nuclear Information System (INIS)

Day, S.A.

1989-01-01

An analysis to determine the risk associated with earthquakes at the Rocky Flats Plant was performed. Seismic analyses and structural evaluations were used to postulate building and equipment damage and radiological releases to the environment from various magnitudes of earthquakes. Dispersion modeling and dose assessment to the public were then calculated. The frequency of occurrence of various magnitudes of earthquakes were determined from the Department of Energy natural Phenomena Hazards Modeling Project. Risk to the public was probabilistically assessed for each magnitude of earthquake and for overall seismic risk. Based on the results of this Probabilistic Risk Assessment and a cost/benefit analysis, seismic upgrades are being implemented for several plutonium-handling facilities for the purpose of risk reduction

N reactor external events probabilistic risk assessment

International Nuclear Information System (INIS)

Baxter, J.T.

1989-01-01

An external events probabilistic risk assessment of the N Reactor has been completed. The methods used are those currently being proposed for external events analysis in NUREG-1150. Results are presented for the external hazards that survived preliminary screening. They are earthquake, fire, and external flood. Core damage frequencies for these hazards are shown to be comparable to those for commercial pressurized water reactors. Dominant fire sequences are described and related to 10 CFR 50, Appendix R design requirements. Potential remedial measures that reduce fire core damage risk are described including modifications to fire protection systems, procedure changes, and addition of new administrative controls. Dominant seismic sequences are described. The effect of non-safety support system dependencies on seismic risk is presented

PRA-Code Upgrade to Handle a Generic Problem

International Nuclear Information System (INIS)

Wilson, J. R.

1999-01-01

During the probabilistic risk assessment (PRA) for the proposed Yucca Mountain nuclear waste repository, a problem came up that could not be handled by most PRA computer codes. This problem deals with dependencies between sequential events in time. Two similar scenarios that illustrate this problem are LOOP nonrecovery and sequential wearout failures with units of time. The purpose of this paper is twofold: To explain the problem generically, and to show how the PRA code at the INEEL, SAPHIRE, has been modified to solve this problem correctly

Probabilistic tsunami hazard assessment at Seaside, Oregon, for near-and far-field seismic sources

Science.gov (United States)

Gonzalez, F.I.; Geist, E.L.; Jaffe, B.; Kanoglu, U.; Mofjeld, H.; Synolakis, C.E.; Titov, V.V.; Areas, D.; Bellomo, D.; Carlton, D.; Horning, T.; Johnson, J.; Newman, J.; Parsons, T.; Peters, R.; Peterson, C.; Priest, G.; Venturato, A.; Weber, J.; Wong, F.; Yalciner, A.

2009-01-01

The first probabilistic tsunami flooding maps have been developed. The methodology, called probabilistic tsunami hazard assessment (PTHA), integrates tsunami inundation modeling with methods of probabilistic seismic hazard assessment (PSHA). Application of the methodology to Seaside, Oregon, has yielded estimates of the spatial distribution of 100- and 500-year maximum tsunami amplitudes, i.e., amplitudes with 1% and 0.2% annual probability of exceedance. The 100-year tsunami is generated most frequently by far-field sources in the Alaska-Aleutian Subduction Zone and is characterized by maximum amplitudes that do not exceed 4 m, with an inland extent of less than 500 m. In contrast, the 500-year tsunami is dominated by local sources in the Cascadia Subduction Zone and is characterized by maximum amplitudes in excess of 10 m and an inland extent of more than 1 km. The primary sources of uncertainty in these results include those associated with interevent time estimates, modeling of background sea level, and accounting for temporal changes in bathymetry and topography. Nonetheless, PTHA represents an important contribution to tsunami hazard assessment techniques; viewed in the broader context of risk analysis, PTHA provides a method for quantifying estimates of the likelihood and severity of the tsunami hazard, which can then be combined with vulnerability and exposure to yield estimates of tsunami risk. Copyright 2009 by the American Geophysical Union.

Probabilistic disaggregation model with application to natural hazard risk assessment of portfolios

DEFF Research Database (Denmark)

Custer, Rocco; Nishijima, Kazuyoshi

In natural hazard risk assessment, a resolution mismatch between hazard data and aggregated exposure data is often observed. A possible solution to this issue is the disaggregation of exposure data to match the spatial resolution of hazard data. Disaggregation models available in literature...... disaggregation model that considers the uncertainty in the disaggregation, taking basis in the scaled Dirichlet distribution. The proposed probabilistic disaggregation model is applied to a portfolio of residential buildings in the Canton Bern, Switzerland, subject to flood risk. Thereby, the model is verified...... are usually deterministic and make use of auxiliary indicator, such as land cover, to spatially distribute exposures. As the dependence between auxiliary indicator and disaggregated number of exposures is generally imperfect, uncertainty arises in disaggregation. This paper therefore proposes a probabilistic...

Probabilistic logics and probabilistic networks

CERN Document Server

Haenni, Rolf; Wheeler, Gregory; Williamson, Jon; Andrews, Jill

2014-01-01

Probabilistic Logic and Probabilistic Networks presents a groundbreaking framework within which various approaches to probabilistic logic naturally fit. Additionally, the text shows how to develop computationally feasible methods to mesh with this framework.

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)

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

Probabilistic Assessment of Structural Seismic Damage for Buildings in Mid-America

International Nuclear Information System (INIS)

Bai, Jong-Wha; Hueste, Mary Beth D.; Gardoni, Paolo

2008-01-01

This paper provides an approach to conduct a probabilistic assessment of structural damage due to seismic events with an application to typical building structures in Mid-America. The developed methodology includes modified damage state classifications based on the ATC-13 and ATC-38 damage states and the ATC-38 database of building damage. Damage factors are assigned to each damage state to quantify structural damage as a percentage of structural replacement cost. To account for the inherent uncertainties, these factors are expressed as random variables with a Beta distribution. A set of fragility curves, quantifying the structural vulnerability of a building, is mapped onto the developed methodology to determine the expected structural damage. The total structural damage factor for a given seismic intensity is then calculated using a probabilistic approach. Prediction and confidence bands are also constructed to account for the prevailing uncertainties. The expected seismic structural damage is assessed for a typical building structure in the Mid-America region using the developed methodology. The developed methodology provides a transparent procedure, where the structural damage factors can be updated as additional seismic damage data becomes available

Development of probabilistic assessment methodology for geologic disposal of radioactive wastes

International Nuclear Information System (INIS)

Kimura, H.; Takahashi, T.

1998-01-01

The probabilistic assessment methodology is essential to evaluate uncertainties of long-term radiological consequences associated with geologic disposal of radioactive wastes. We have developed a probabilistic assessment methodology to estimate the influences of parameter uncertainties/variabilities. An exposure scenario considered here is based on a groundwater migration scenario. A computer code system GSRW-PSA thus developed is based on a non site-specific model, and consists of a set of sub modules for sampling of model parameters, calculating the release of radionuclides from engineered barriers, calculating the transport of radionuclides through the geosphere, calculating radiation exposures of the public, and calculating the statistical values relating the uncertainties and sensitivities. The results of uncertainty analyses for α-nuclides quantitatively indicate that natural uranium ( 238 U) concentration is suitable for an alternative safety indicator of long-lived radioactive waste disposal, because the estimated range of individual dose equivalent due to 238 U decay chain is narrower that that due to other decay chain ( 237 Np decay chain). It is internationally necessary to have detailed discussion on the PDF of model parameters and the PSA methodology to evaluated the uncertainties due to conceptual models and scenarios. (author)

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

International Nuclear Information System (INIS)

2002-01-01

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

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)

Seismic Probabilistic Risk Assessment (SPRA), approach and results

International Nuclear Information System (INIS)

Campbell, R.D.

1995-01-01

During the past 15 years there have been over 30 Seismic Probabilistic Risk Assessments (SPRAs) and Seismic Probabilistic Safety Assessments (SPSAs) conducted of Western Nuclear Power Plants, principally of US design. In this paper PRA and PSA are used interchangeably as the overall process is essentially the same. Some similar assessments have been done for reactors in Taiwan, Korea, Japan, Switzerland and Slovenia. These plants were also principally US supplied or built under US license. Since the restructuring of the governments in former Soviet Bloc countries, there has been grave concern regarding the safety of the reactors in these countries. To date there has been considerable activity in conducting partial seismic upgrades but the overall quantification of risk has not been pursued to the depth that it has in Western countries. This paper summarizes the methodology for Seismic PRA/PSA and compares results of two partially completed and two completed PRAs of soviet designed reactors to results from earlier PRAs on US Reactors. A WWER 440 and a WWER 1000 located in low seismic activity regions have completed PRAs and results show the seismic risk to be very low for both designs. For more active regions, partially completed PRAs of a WWER 440 and WWER 1000 located at the same site show the WWER 440 to have much greater seismic risk than the WWER 1000 plant. The seismic risk from the 1000 MW plant compares with the high end of seismic risk for earlier seismic PRAs in the US. Just as for most US plants, the seismic risk appears to be less than the risk from internal events if risk is measured is terms of mean core damage frequency. However, due to the lack of containment for the earlier WWER 440s, the risk to the public may be significantly greater due to the more probable scenario of an early release. The studies reported have not taken the accident sequences beyond the stage of core damage hence the public heath risk ratios are speculative. (author)

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

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

TOPAS 1 - construction and test of a scintillation counter hodoscope for the tagging of bremsstrahlung photons for the SAPHIR detector

International Nuclear Information System (INIS)

Merkel, R.

1989-09-01

The development of a tagging-hodoscope for the SAPHIR-detector at the stretcher ring ELSA in Bonn is described. The hodoscope covers the energy range 2.175 GeV γ 0 =3.500 GeV. 24 scintillation counters are used for the determination of the photon energy, giving a resolution of ΔE γ =25 MeV. The tagging method requires a good coincidence timing resoluting τ between the tagging hodoscope and the detector for the photon-induced reactions in order to keep the accidental coincidences low. The timing information is given by 8 fast timing counters (40 mm thick), covering 5 up to 7 energy channels each. Fluctuations of the timing signal which result from different impact-locations on the timing counter, due to different light travelling distances, are corrected by the energy defining counters. The timing-component (8 timing counters) is commpleted and tested. The results of first mesurements show an upper limit of σ=250 psec for the resolution of 7 coincidences out of 45 possible channels in the tagging hodscope. These results are obtained with a preliminary adjustment of the SAPHIR beam-line and with a not yet optimized signal to noize ratio in the extracted beam. We hope to obtain a σ<200 psec under optimized conditions. (orig.)

The importance of long range atmospheric transport in probabilistic accident consequence assessment

International Nuclear Information System (INIS)

ApSimon, H.M.; Goddard, A.J.H.; Wilson, J.J.N.

1988-01-01

The disaster at the Chernobyl-4 reactor has demonstrated that severe nuclear accidents can give rise to significant radiological consequences several thousand kilometres from the source. The subsequent dispersion of the release over much of Western Europe further demonstrated the importance of synoptic scale weather patterns in determining the magnitude of the consequences of such accidents. A version of the MESOS-II European scale trajectory model, which is able to simulate large scale variations in weather conditions through the use of spatially and temporally variable meteorological input data, has been used to simulate the pattern of dispersion from Chernobyl with some success. This paper presents the results of probabilistic consequence assessments for a number of West European sites, made using the MESOS-II model. The results illustrate the effects, on probabilistic assessments, of using a more realistic treatment of long range atmospheric transport than the Gaussian plume model and also the spatial variation in the distributions of consequences arising from the variation in synoptic scale weather conditions across Western Europe

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)

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.

Assessing Probabilistic Risk Assessment Approaches for Insect Biological Control Introductions.

Science.gov (United States)

Kaufman, Leyla V; Wright, Mark G

2017-07-07

The introduction of biological control agents to new environments requires host specificity tests to estimate potential non-target impacts of a prospective agent. Currently, the approach is conservative, and is based on physiological host ranges determined under captive rearing conditions, without consideration for ecological factors that may influence realized host range. We use historical data and current field data from introduced parasitoids that attack an endemic Lepidoptera species in Hawaii to validate a probabilistic risk assessment (PRA) procedure for non-target impacts. We use data on known host range and habitat use in the place of origin of the parasitoids to determine whether contemporary levels of non-target parasitism could have been predicted using PRA. Our results show that reasonable predictions of potential non-target impacts may be made if comprehensive data are available from places of origin of biological control agents, but scant data produce poor predictions. Using apparent mortality data rather than marginal attack rate estimates in PRA resulted in over-estimates of predicted non-target impact. Incorporating ecological data into PRA models improved the predictive power of the risk assessments.

Probabilistic disaggregation of a spatial portfolio of exposure for natural hazard risk assessment

DEFF Research Database (Denmark)

Custer, Rocco; Nishijima, Kazuyoshi

2018-01-01

In natural hazard risk assessment situations are encountered where information on the portfolio of exposure is only available in a spatially aggregated form, hindering a precise risk assessment. Recourse might be found in the spatial disaggregation of the portfolio of exposure to the resolution...... of a portfolio of buildings in two communes in Switzerland and the results are compared to sample observations. The relevance of probabilistic disaggregation uncertainty in natural hazard risk assessment is illustrated with the example of a simple flood risk assessment....

2009 Space Shuttle Probabilistic Risk Assessment Overview

Science.gov (United States)

Hamlin, Teri L.; Canga, Michael A.; Boyer, Roger L.; Thigpen, Eric B.

2010-01-01

Loss of a Space Shuttle during flight has severe consequences, including loss of a significant national asset; loss of national confidence and pride; and, most importantly, loss of human life. The Shuttle Probabilistic Risk Assessment (SPRA) is used to identify risk contributors and their significance; thus, assisting management in determining how to reduce risk. In 2006, an overview of the SPRA Iteration 2.1 was presented at PSAM 8 [1]. Like all successful PRAs, the SPRA is a living PRA and has undergone revisions since PSAM 8. The latest revision to the SPRA is Iteration 3. 1, and it will not be the last as the Shuttle program progresses and more is learned. This paper discusses the SPRA scope, overall methodology, and results, as well as provides risk insights. The scope, assumptions, uncertainties, and limitations of this assessment provide risk-informed perspective to aid management s decision-making process. In addition, this paper compares the Iteration 3.1 analysis and results to the Iteration 2.1 analysis and results presented at PSAM 8.

Probabilistic versus deterministic hazard assessment in liquefaction susceptible zones

Science.gov (United States)

Daminelli, Rosastella; Gerosa, Daniele; Marcellini, Alberto; Tento, Alberto

2015-04-01

Probabilistic seismic hazard assessment (PSHA), usually adopted in the framework of seismic codes redaction, is based on Poissonian description of the temporal occurrence, negative exponential distribution of magnitude and attenuation relationship with log-normal distribution of PGA or response spectrum. The main positive aspect of this approach stems into the fact that is presently a standard for the majority of countries, but there are weak points in particular regarding the physical description of the earthquake phenomenon. Factors like site effects, source characteristics like duration of the strong motion and directivity that could significantly influence the expected motion at the site are not taken into account by PSHA. Deterministic models can better evaluate the ground motion at a site from a physical point of view, but its prediction reliability depends on the degree of knowledge of the source, wave propagation and soil parameters. We compare these two approaches in selected sites affected by the May 2012 Emilia-Romagna and Lombardia earthquake, that caused widespread liquefaction phenomena unusually for magnitude less than 6. We focus on sites liquefiable because of their soil mechanical parameters and water table level. Our analysis shows that the choice between deterministic and probabilistic hazard analysis is strongly dependent on site conditions. The looser the soil and the higher the liquefaction potential, the more suitable is the deterministic approach. Source characteristics, in particular the duration of strong ground motion, have long since recognized as relevant to induce liquefaction; unfortunately a quantitative prediction of these parameters appears very unlikely, dramatically reducing the possibility of their adoption in hazard assessment. Last but not least, the economic factors are relevant in the choice of the approach. The case history of 2012 Emilia-Romagna and Lombardia earthquake, with an officially estimated cost of 6 billions

EVALUATION OF MILITARY ACTIVITY IMPACT ON HUMANS THROUGH A PROBABILISTIC ECOLOGICAL RISK ASSESSMENT. EXAMPLE OF A FORMER MISSILE BASE.

Directory of Open Access Journals (Sweden)

Sergiy ОREL

2015-10-01

Full Text Available The current article provides a methodology focused on the assessment of environmental factors after termination of military activity and uses a former missile base as an example. The assessment of environmental conditions is performed through an evaluation of the risks posed by the hazardous chemicals contained by underground and surface water sources and soil to human health . Moreover, by conducting deterministic and probabilistic risk assessments, the article determines that the probabilistic assessment provides more accurate and qualitative information for decision-making on the use of environmental protection measures, which often saves financial and material resources needed for their implementation.

Probabilistic Criterion for the Economical Assessment of Nuclear Reactors

International Nuclear Information System (INIS)

Juanico, L; Florido, Pablo; Bergallo, Juan

2000-01-01

In this paper a MonteCarlo probabilistic model for the economic evaluation of nuclear power plants is presented.The probabilistic results have shown a wide spread on the economic performance due to the schedule complexity and coupling if tasks.This spread increasing to the discount rate, end hence, it becomes more important for developing countries

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin [Power Engineering Research Institute, KEPCO Engineering and Construction, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2014-08-15

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor.

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

International Nuclear Information System (INIS)

Oh, Young-Jin; Chang, Yoon-Suk

2014-01-01

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor

Development of a Probabilistic Flood Hazard Assessment (PFHA) for the nuclear safety

Science.gov (United States)

Ben Daoued, Amine; Guimier, Laurent; Hamdi, Yasser; Duluc, Claire-Marie; Rebour, Vincent

2016-04-01

The purpose of this study is to lay the basis for a probabilistic evaluation of flood hazard (PFHA). Probabilistic assessment of external floods has become a current topic of interest to the nuclear scientific community. Probabilistic approaches complement deterministic approaches by exploring a set of scenarios and associating a probability to each of them. These approaches aim to identify all possible failure scenarios, combining their probability, in order to cover all possible sources of risk. They are based on the distributions of initiators and/or the variables caracterizing these initiators. The PFHA can characterize the water level for example at defined point of interest in the nuclear site. This probabilistic flood hazard characterization takes into account all the phenomena that can contribute to the flooding of the site. The main steps of the PFHA are: i) identification of flooding phenomena (rains, sea water level, etc.) and screening of relevant phenomena to the nuclear site, ii) identification and probabilization of parameters associated to selected flooding phenomena, iii) spreading of the probabilized parameters from the source to the point of interest in the site, v) obtaining hazard curves and aggregation of flooding phenomena contributions at the point of interest taking into account the uncertainties. Within this framework, the methodology of the PFHA has been developed for several flooding phenomena (rain and/or sea water level, etc.) and then implemented and tested with a simplified case study. In the same logic, our study is still in progress to take into account other flooding phenomena and to carry out more case studies.

A perspective of PC-based probabilistic risk assessment

International Nuclear Information System (INIS)

Sattison, M.B.; Rasmuson, D.M.; Robinson, R.C.; Russell, K.D.; Van Siclen, V.S.

1987-01-01

Probabilistic risk assessment (PRA) information has been under-utilized in the past due to the large effort required to input the PRA data and the large expense of the computers needed to run PRA codes. The microcomputer-based Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) System, under development at the Idaho National Engineering Laboratory, have greatly enhanced the ability of managers to use PRA techniques in their decision-making. IRRAS is a tool that allows an analyst to create, modify, update, and reanalyze a plant PRA to keep the risk assessment current with the plant's configuration and operation. The SARA system is used to perform sensitivity studies on the results of a PRA. This type of analysis can be used to evaluate proposed changes to a plant or its operation. The success of these two software projects demonstrate that risk information can be made readily available to those that need it. This is the first step in the development of a true risk management capability

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

International Nuclear Information System (INIS)

Nilsson, Fred

1997-01-01

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

Applications of the EBR-II Probabilistic Risk Assessment

International Nuclear Information System (INIS)

Roglans, J.: Ragland, W.A.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor 11 (EBR-11), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL), and has been performed with close collaboration between PRA analysts and engineering and operations staff. A product of this Involvement of plant personnel has been a excellent acceptance of the PRA as a tool, which has already resulted In a variety of applications of the EBR-11 PRA. The EBR-11 has been used in support of plant hardware and procedure modifications and In new system design work. A new application in support of the refueling safety analysis will be completed in the near future

14th International Probabilistic Workshop

CERN Document Server

Taerwe, Luc; Proske, Dirk

2017-01-01

This book presents the proceedings of the 14th International Probabilistic Workshop that was held in Ghent, Belgium in December 2016. Probabilistic methods are currently of crucial importance for research and developments in the field of engineering, which face challenges presented by new materials and technologies and rapidly changing societal needs and values. Contemporary needs related to, for example, performance-based design, service-life design, life-cycle analysis, product optimization, assessment of existing structures and structural robustness give rise to new developments as well as accurate and practically applicable probabilistic and statistical engineering methods to support these developments. These proceedings are a valuable resource for anyone interested in contemporary developments in the field of probabilistic engineering applications.

Neo-Deterministic and Probabilistic Seismic Hazard Assessments: a Comparative Analysis

Science.gov (United States)

Peresan, Antonella; Magrin, Andrea; Nekrasova, Anastasia; Kossobokov, Vladimir; Panza, Giuliano F.

2016-04-01

Objective testing is the key issue towards any reliable seismic hazard assessment (SHA). Different earthquake hazard maps must demonstrate their capability in anticipating ground shaking from future strong earthquakes before an appropriate use for different purposes - such as engineering design, insurance, and emergency management. Quantitative assessment of maps performances is an essential step also in scientific process of their revision and possible improvement. Cross-checking of probabilistic models with available observations and independent physics based models is recognized as major validation procedure. The existing maps from the classical probabilistic seismic hazard analysis (PSHA), as well as those from the neo-deterministic analysis (NDSHA), which have been already developed for several regions worldwide (including Italy, India and North Africa), are considered to exemplify the possibilities of the cross-comparative analysis in spotting out limits and advantages of different methods. Where the data permit, a comparative analysis versus the documented seismic activity observed in reality is carried out, showing how available observations about past earthquakes can contribute to assess performances of the different methods. Neo-deterministic refers to a scenario-based approach, which allows for consideration of a wide range of possible earthquake sources as the starting point for scenarios constructed via full waveforms modeling. The method does not make use of empirical attenuation models (i.e. Ground Motion Prediction Equations, GMPE) and naturally supplies realistic time series of ground shaking (i.e. complete synthetic seismograms), readily applicable to complete engineering analysis and other mitigation actions. The standard NDSHA maps provide reliable envelope estimates of maximum seismic ground motion from a wide set of possible scenario earthquakes, including the largest deterministically or historically defined credible earthquake. In addition

Probabilistic assessment of roadway departure risk in a curve

Science.gov (United States)

Rey, G.; Clair, D.; Fogli, M.; Bernardin, F.

2011-10-01

Roadway departure while cornering constitutes a major part of car accidents and casualties in France. Even though drastic policy about overspeeding contributes to reduce accidents, there obviously exist other factors. This article presents the construction of a probabilistic strategy for the roadway departure risk assessment. A specific vehicle dynamic model is developed in which some parameters are modelled by random variables. These parameters are deduced from a sensitivity analysis to ensure an efficient representation of the inherent uncertainties of the system. Then, structural reliability methods are employed to assess the roadway departure risk in function of the initial conditions measured at the entrance of the curve. This study is conducted within the French national road safety project SARI that aims to implement a warning systems alerting the driver in case of dangerous situation.

Probabilistic Risk Assessment (PRA): A Practical and Cost Effective Approach

Science.gov (United States)

Lee, Lydia L.; Ingegneri, Antonino J.; Djam, Melody

2006-01-01

The Lunar Reconnaissance Orbiter (LRO) is the first mission of the Robotic Lunar Exploration Program (RLEP), a space exploration venture to the Moon, Mars and beyond. The LRO mission includes spacecraft developed by NASA Goddard Space Flight Center (GSFC) and seven instruments built by GSFC, Russia, and contractors across the nation. LRO is defined as a measurement mission, not a science mission. It emphasizes the overall objectives of obtaining data to facilitate returning mankind safely to the Moon in preparation for an eventual manned mission to Mars. As the first mission in response to the President's commitment of the journey of exploring the solar system and beyond: returning to the Moon in the next decade, then venturing further into the solar system, ultimately sending humans to Mars and beyond, LRO has high-visibility to the public but limited resources and a tight schedule. This paper demonstrates how NASA's Lunar Reconnaissance Orbiter Mission project office incorporated reliability analyses in assessing risks and performing design tradeoffs to ensure mission success. Risk assessment is performed using NASA Procedural Requirements (NPR) 8705.5 - Probabilistic Risk Assessment (PRA) Procedures for NASA Programs and Projects to formulate probabilistic risk assessment (PRA). As required, a limited scope PRA is being performed for the LRO project. The PRA is used to optimize the mission design within mandated budget, manpower, and schedule constraints. The technique that LRO project office uses to perform PRA relies on the application of a component failure database to quantify the potential mission success risks. To ensure mission success in an efficient manner, low cost and tight schedule, the traditional reliability analyses, such as reliability predictions, Failure Modes and Effects Analysis (FMEA), and Fault Tree Analysis (FTA), are used to perform PRA for the large system of LRO with more than 14,000 piece parts and over 120 purchased or contractor

Probabilistic Seismic Hazard Assessment Method for Nonlinear Soil Sites based on the Hazard Spectrum of Bedrock Sites

International Nuclear Information System (INIS)

Hahm, Dae Gi; Seo, Jeong Moon; Choi, In Kil

2011-01-01

For the probabilistic safety assessment of the nuclear power plants (NPP) under seismic events, the rational probabilistic seismic hazard estimation should be performed. Generally, the probabilistic seismic hazard of NPP site is represented by the uniform hazard spectrum (UHS) for the specific annual frequency. In most case, since that the attenuation equations were defined for the bedrock sites, the standard attenuation laws cannot be applied to the general soft soil sites. Hence, for the probabilistic estimation of the seismic hazard of soft soil sites, a methodology of probabilistic seismic hazard analysis (PSHA) coupled with nonlinear dynamic analyses of the soil column are required. Two methods are commonly used for the site response analysis considering the nonlinearity of sites. The one is the deterministic method and another is the probabilistic method. In the analysis of site response, there exist many uncertainty factors such as the variation of the magnitude and frequency contents of input ground motion, and material properties of soil deposits. Hence, nowadays, it is recommended that the adoption of the probabilistic method for the PSHA of soft soil deposits considering such uncertainty factors. In this study, we estimated the amplification factor of the surface of the soft soil deposits with considering the uncertainties of the input ground motions and the soil material properties. Then, we proposed the probabilistic methodology to evaluate the UHS of the soft soil site by multiplying the amplification factor to that of the bedrock site. The proposed method was applied to four typical target sites of KNGR and APR1400 NPP site categories

Biomechanical rupture risk assessment of abdominal aortic aneurysms based on a novel probabilistic rupture risk index.

Science.gov (United States)

Polzer, Stanislav; Gasser, T Christian

2015-12-06

A rupture risk assessment is critical to the clinical treatment of abdominal aortic aneurysm (AAA) patients. The biomechanical AAA rupture risk assessment quantitatively integrates many known AAA rupture risk factors but the variability of risk predictions due to model input uncertainties remains a challenging limitation. This study derives a probabilistic rupture risk index (PRRI). Specifically, the uncertainties in AAA wall thickness and wall strength were considered, and wall stress was predicted with a state-of-the-art deterministic biomechanical model. The discriminative power of PRRI was tested in a diameter-matched cohort of ruptured (n = 7) and intact (n = 7) AAAs and compared to alternative risk assessment methods. Computed PRRI at 1.5 mean arterial pressure was significantly (p = 0.041) higher in ruptured AAAs (20.21(s.d. 14.15%)) than in intact AAAs (3.71(s.d. 5.77)%). PRRI showed a high sensitivity and specificity (discriminative power of 0.837) to discriminate between ruptured and intact AAA cases. The underlying statistical representation of stochastic data of wall thickness, wall strength and peak wall stress had only negligible effects on PRRI computations. Uncertainties in AAA wall stress predictions, the wide range of reported wall strength and the stochastic nature of failure motivate a probabilistic rupture risk assessment. Advanced AAA biomechanical modelling paired with a probabilistic rupture index definition as known from engineering risk assessment seems to be superior to a purely deterministic approach. © 2015 The Author(s).

A framework to integrate software behavior into dynamic probabilistic risk assessment

International Nuclear Information System (INIS)

Zhu Dongfeng; Mosleh, Ali; Smidts, Carol

2007-01-01

Software plays an increasingly important role in modern safety-critical systems. Although, research has been done to integrate software into the classical probabilistic risk assessment (PRA) framework, current PRA practice overwhelmingly neglects the contribution of software to system risk. Dynamic probabilistic risk assessment (DPRA) is considered to be the next generation of PRA techniques. DPRA is a set of methods and techniques in which simulation models that represent the behavior of the elements of a system are exercised in order to identify risks and vulnerabilities of the system. The fact remains, however, that modeling software for use in the DPRA framework is also quite complex and very little has been done to address the question directly and comprehensively. This paper develops a methodology to integrate software contributions in the DPRA environment. The framework includes a software representation, and an approach to incorporate the software representation into the DPRA environment SimPRA. The software representation is based on multi-level objects and the paper also proposes a framework to simulate the multi-level objects in the simulation-based DPRA environment. This is a new methodology to address the state explosion problem in the DPRA environment. This study is the first systematic effort to integrate software risk contributions into DPRA environments

Issues related to structural aging in probabilistic risk assessment of nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, Bruce R.

1998-01-01

Structural components and systems have an important safety function in nuclear power plants. Although they are essentially passive under normal operating conditions, they play a key role in mitigating the impact of extreme environmental events such as earthquakes, winds, fire and floods on plant safety. Moreover, the importance of structural components and systems in accident mitigation is amplified by common-cause effects. Reinforced concrete structural components and systems in NPPs are subject to a phenomenon known as aging, leading to time-dependent changes in strength and stiffness that may impact their ability to withstand various challenges during their service lives from operation, the environment and accidents. Time-dependent changes in structural properties as well as challenges to the system are random in nature. Accordingly, condition assessment of existing structures should be performed within a probabilistic framework. The mathematical formalism of a probabilistic risk assessment (PRA) provides a means for identifying aging structural components that may play a significant role in mitigating plant risk. Structural condition assessments supporting a decision regarding continued service can be rendered more efficient if guided by the logic of a PRA

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 simulation applications to reliability assessments

International Nuclear Information System (INIS)

Miller, Ian; Nutt, Mark W.; Hill, Ralph S. III

2003-01-01

Probabilistic risk/reliability (PRA) analyses for engineered systems are conventionally based on fault-tree methods. These methods are mature and efficient, and are well suited to systems consisting of interacting components with known, low probabilities of failure. Even complex systems, such as nuclear power plants or aircraft, are modeled by the careful application of these approaches. However, for systems that may evolve in complex and nonlinear ways, and where the performance of components may be a sensitive function of the history of their working environments, fault-tree methods can be very demanding. This paper proposes an alternative method of evaluating such systems, based on probabilistic simulation using intelligent software objects to represent the components of such systems. Using a Monte Carlo approach, simulation models can be constructed from relatively simple interacting objects that capture the essential behavior of the components that they represent. Such models are capable of reflecting the complex behaviors of the systems that they represent in a natural and realistic way. (author)

The Terrestrial Investigation Model: A probabilistic risk assessment model for birds exposed to pesticides

Science.gov (United States)

One of the major recommendations of the National Academy of Science to the USEPA, NMFS and USFWS was to utilize probabilistic methods when assessing the risks of pesticides to federally listed endangered and threatened species. The Terrestrial Investigation Model (TIM, version 3....

APROBA-Plus: A probabilistic tool to evaluate and express uncertainty in hazard characterization and exposure assessment of substances.

Science.gov (United States)

Bokkers, Bas G H; Mengelers, Marcel J; Bakker, Martine I; Chiu, Weihsueh A; Slob, Wout

2017-12-01

To facilitate the application of probabilistic risk assessment, the WHO released the APROBA tool. This tool applies lognormal uncertainty distributions to the different aspects of the hazard characterization, resulting in a probabilistic health-based guidance value. The current paper describes an extension, APROBA-Plus, which combines the output from the probabilistic hazard characterization with the probabilistic exposure to rapidly characterize risk and its uncertainty. The uncertainty in exposure is graphically compared with the uncertainty in the target human dose, i.e. the dose that complies with the specified protection goals. APROBA-Plus is applied to several case studies, resulting in distinct outcomes and illustrating that APROBA-Plus could serve as a standard extension of routine risk assessments. By visualizing the uncertainties, APROBA-Plus provides a more transparent and informative outcome than the more usual deterministic approaches, so that risk managers can make better informed decisions. For example, APROBA-Plus can help in deciding whether risk-reducing measures are warranted or that a refined risk assessment would first be needed. If the latter, the tool can be used to prioritize possible refinements. APROBA-Plus may also be used to rank substances into different risk categories, based on potential health risks without being compromised by different levels of conservatism that may be associated with point estimates of risk. Copyright © 2017 Elsevier Ltd. All rights reserved.

Probabilistic risk assessment for the Sandia National Laboratories Technical Area V Liquid Waste Disposal System surface impoundments

International Nuclear Information System (INIS)

Dawson, L.A.; Eidson, A.F.

1996-01-01

A probabilistic risk assessment was completed for a former radioactive waste disposal site. The site, two unlined surface impoundment, was designed as part of the Liquid Waste Disposal System (LWDS) to receive radioactive effluent from nuclear reactors in Technical Area-V (TA-V) at Sandia National Laboratories/New Mexico (SNL/NM). First, a statistical comparison of site sampling results to natural background, using EPA methods, and a spatial distribution analysis were performed. Risk assessment was conducted with SNL/NM's Probabilistic Risk Evaluation and Characterization Investigation System model. The risk assessment indicated that contamination from several constituents might have been high enough to require remediation. However, further analysis based on expected site closure activities and recent EPA guidance indicated that No Further Action was acceptable

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

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)

Probabilistic exposure assessment to face and oral care cosmetic products by the French population.

Science.gov (United States)

Bernard, A; Dornic, N; Roudot, Ac; Ficheux, As

2018-01-01

Cosmetic exposure data for face and mouth are limited in Europe. The aim of the study was to assess the exposure to face cosmetics using recent French consumption data (Ficheux et al., 2016b, 2015). Exposure was assessed using a probabilistic method for thirty one face products from four lines of products: cleanser, care, make-up and make-up remover products and two oral care products. Probabilistic exposure was assessed for different subpopulation according to sex and age in adults and children. Pregnant women were also studied. The levels of exposure to moisturizing cream, lip balm, mascara, eyeliner, cream foundation, toothpaste and mouthwash were higher than the values currently used by the Scientific Committee on Consumer Safety (SCCS). Exposure values found for eye shadow, lipstick, lotion and milk (make-up remover) were lower than SCCS values. These new French exposure values will be useful for safety assessors and for safety agencies in order to protect the general population and the at risk populations. Copyright © 2017. Published by Elsevier Ltd.

Probabilistic Design of Coastal Flood Defences in Vietnam

NARCIS (Netherlands)

Mai Van, C.

2010-01-01

This study further develops the method of probabilistic design and to address a knowledge gap in its application regarding safety and reliability, risk assessment and risk evaluation to the fields of flood defences. The thesis discusses: - a generic probabilistic design framework for assessing flood

The Gain-Loss Model: A Probabilistic Skill Multimap Model for Assessing Learning Processes

Science.gov (United States)

Robusto, Egidio; Stefanutti, Luca; Anselmi, Pasquale

2010-01-01

Within the theoretical framework of knowledge space theory, a probabilistic skill multimap model for assessing learning processes is proposed. The learning process of a student is modeled as a function of the student's knowledge and of an educational intervention on the attainment of specific skills required to solve problems in a knowledge…

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.

Probabilistic risk assessment of abalone Haliotis diversicolor supertexta exposed to waterborne zinc

International Nuclear Information System (INIS)

Liao Chungmin; Ling Minpei

2004-01-01

This paper describes a risk assessment approach that integrates predicted tissue concentrations of zinc (Zn) with a concentration-response relationship and leads to predictions of survival risk for pond abalone Haliotis diversicolor supertexta as well as to the uncertainties associated with these predictions. The models implemented include a probabilistic bioaccumulation model, which linking biokinetic and consumer-resource models, accounts for Zn exposure profile and a modified Hill model for reconstructing a dose-response profile for abalone exposed to waterborne Zn. The growth risk is assessed by hazard quotients characterized by measured water level and chronic no-observed effect concentration. Our risk analyses for H. diversicolor supertexta reared near Toucheng, Kouhu, and Anping, respectively, in north, central, and south Taiwan region indicate a relatively low likelihood that survival is being affected by waterborne Zn. Expected risks of mortality for abalone were estimated as 0.46 (Toucheng), 0.36 (Kouhu), and 0.29 (Anping). The predicted 90th-percentiles of hazard quotient for potential growth risk were estimated as 1.94 (Toucheng), 0.47 (Kouhu), and 0.51 (Anping). These findings indicate that waterborne Zn exposure poses no significant risk to pond abalone in Kouhu and Anping, yet a relative high growth risk in Toucheng is alarming. Because of a scarcity of toxicity and exposure data, the probabilistic risk assessment was based on very conservative assumptions. - A novel risk assessment method was developed for abalone

Probabilistic risk assessment of abalone Haliotis diversicolor supertexta exposed to waterborne zinc

Energy Technology Data Exchange (ETDEWEB)

Liao Chungmin; Ling Minpei

2004-01-01

This paper describes a risk assessment approach that integrates predicted tissue concentrations of zinc (Zn) with a concentration-response relationship and leads to predictions of survival risk for pond abalone Haliotis diversicolor supertexta as well as to the uncertainties associated with these predictions. The models implemented include a probabilistic bioaccumulation model, which linking biokinetic and consumer-resource models, accounts for Zn exposure profile and a modified Hill model for reconstructing a dose-response profile for abalone exposed to waterborne Zn. The growth risk is assessed by hazard quotients characterized by measured water level and chronic no-observed effect concentration. Our risk analyses for H. diversicolor supertexta reared near Toucheng, Kouhu, and Anping, respectively, in north, central, and south Taiwan region indicate a relatively low likelihood that survival is being affected by waterborne Zn. Expected risks of mortality for abalone were estimated as 0.46 (Toucheng), 0.36 (Kouhu), and 0.29 (Anping). The predicted 90th-percentiles of hazard quotient for potential growth risk were estimated as 1.94 (Toucheng), 0.47 (Kouhu), and 0.51 (Anping). These findings indicate that waterborne Zn exposure poses no significant risk to pond abalone in Kouhu and Anping, yet a relative high growth risk in Toucheng is alarming. Because of a scarcity of toxicity and exposure data, the probabilistic risk assessment was based on very conservative assumptions. - A novel risk assessment method was developed for abalone.

Potential for the adoption of probabilistic risk assessments by end-users and decision-makers

DEFF Research Database (Denmark)

Frewer, Lynn J.; Fischer, Arnout R. H.; van den Brink, Paul J.

2008-01-01

-user and regulatory uptake has not been, to date, extensive. A case study, utilizing the Theory of Planned Behavior, was conducted in order to identify potential determinants of end-user adoption of probabilistic risk assessments associated with the ecotoxicological impact of pesticides. Seventy potential end...

Assessing Probabilistic Risk Assessment Approaches for Insect Biological Control Introductions

Directory of Open Access Journals (Sweden)

Leyla V. Kaufman

2017-07-01

Full Text Available The introduction of biological control agents to new environments requires host specificity tests to estimate potential non-target impacts of a prospective agent. Currently, the approach is conservative, and is based on physiological host ranges determined under captive rearing conditions, without consideration for ecological factors that may influence realized host range. We use historical data and current field data from introduced parasitoids that attack an endemic Lepidoptera species in Hawaii to validate a probabilistic risk assessment (PRA procedure for non-target impacts. We use data on known host range and habitat use in the place of origin of the parasitoids to determine whether contemporary levels of non-target parasitism could have been predicted using PRA. Our results show that reasonable predictions of potential non-target impacts may be made if comprehensive data are available from places of origin of biological control agents, but scant data produce poor predictions. Using apparent mortality data rather than marginal attack rate estimates in PRA resulted in over-estimates of predicted non-target impact. Incorporating ecological data into PRA models improved the predictive power of the risk assessments.

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

Probabilistic Capacity Assessment of Lattice Transmission Towers under Strong Wind

Directory of Open Access Journals (Sweden)

Wei eZhang

2015-10-01

Full Text Available Serving as one key component of the most important lifeline infrastructure system, transmission towers are vulnerable to multiple nature hazards including strong wind and could pose severe threats to the power system security with possible blackouts under extreme weather conditions, such as hurricanes, derechoes, or winter storms. For the security and resiliency of the power system, it is important to ensure the structural safety with enough capacity for all possible failure modes, such as structural stability. The study is to develop a probabilistic capacity assessment approach for transmission towers under strong wind loads. Due to the complicated structural details of lattice transmission towers, wind tunnel experiments are carried out to understand the complex interactions of wind and the lattice sections of transmission tower and drag coefficients and the dynamic amplification factor for different panels of the transmission tower are obtained. The wind profile is generated and the wind time histories are simulated as a summation of time-varying mean and fluctuating components. The capacity curve for the transmission towers is obtained from the incremental dynamic analysis (IDA method. To consider the stochastic nature of wind field, probabilistic capacity curves are generated by implementing IDA analysis for different wind yaw angles and different randomly generated wind speed time histories. After building the limit state functions based on the maximum allowable drift to height ratio, the probabilities of failure are obtained based on the meteorological data at a given site. As the transmission tower serves as the key nodes for the power network, the probabilistic capacity curves can be incorporated into the performance based design of the power transmission network.

HTGR containment design options: an application of probabilistic risk assessment

International Nuclear Information System (INIS)

1977-08-01

Through the use of probabilistic risk assessment (PRA), it is possible to quantitatively evaluate the radiological risk associated with a given reactor design and to place such risk into perspective with alternative designs. The merits are discussed for several containment alternatives for the HTGR from the viewpoints of economics and licensability, as well as public risk. The quantification of cost savings and public risk indicates that presently acceptable public risk can be maintained and cost savings of $40 million can result from use of a vented confinement for the HTGR

Simplified probabilistic risk assessment in fuel reprocessing

International Nuclear Information System (INIS)

Solbrig, C.W.

1993-01-01

An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000

Innovative probabilistic risk assessment applications: barrier impairments and fracture toughness. Panel Discussion

International Nuclear Information System (INIS)

Osterman, Michael; Root, Steven; Li, F.; Modarres, Mohammad; Reinhart, F. Mark; Bradley, Biff; Calhoun, David J.

2001-01-01

Full text of publication follows: New probabilistic risk assessment (PRA) applications promise to improve the overall safety and efficiency of nuclear plant operations. This discussion will explore the use of PRA in evaluating barrier integrity with respect to the consequences of natural phenomena such as tornadoes, floods, and harsh environments. Additionally, the session will explore proposals to improve fracture toughness techniques using PRA. (authors)

Probabilistic Seismic Hazard Assessment for Northeast India Region

Science.gov (United States)

Das, Ranjit; Sharma, M. L.; Wason, H. R.

2016-08-01

Northeast India bounded by latitudes 20°-30°N and longitudes 87°-98°E is one of the most seismically active areas in the world. This region has experienced several moderate-to-large-sized earthquakes, including the 12 June, 1897 Shillong earthquake ( M w 8.1) and the 15 August, 1950 Assam earthquake ( M w 8.7) which caused loss of human lives and significant damages to buildings highlighting the importance of seismic hazard assessment for the region. Probabilistic seismic hazard assessment of the region has been carried out using a unified moment magnitude catalog prepared by an improved General Orthogonal Regression methodology (Geophys J Int, 190:1091-1096, 2012; Probabilistic seismic hazard assessment of Northeast India region, Ph.D. Thesis, Department of Earthquake Engineering, IIT Roorkee, Roorkee, 2013) with events compiled from various databases (ISC, NEIC,GCMT, IMD) and other available catalogs. The study area has been subdivided into nine seismogenic source zones to account for local variation in tectonics and seismicity characteristics. The seismicity parameters are estimated for each of these source zones, which are input variables into seismic hazard estimation of a region. The seismic hazard analysis of the study region has been performed by dividing the area into grids of size 0.1° × 0.1°. Peak ground acceleration (PGA) and spectral acceleration ( S a) values (for periods of 0.2 and 1 s) have been evaluated at bedrock level corresponding to probability of exceedance (PE) of 50, 20, 10, 2 and 0.5 % in 50 years. These exceedance values correspond to return periods of 100, 225, 475, 2475, and 10,000 years, respectively. The seismic hazard maps have been prepared at the bedrock level, and it is observed that the seismic hazard estimates show a significant local variation in contrast to the uniform hazard value suggested by the Indian standard seismic code [Indian standard, criteria for earthquake-resistant design of structures, fifth edition, Part

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

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)

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

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

Comparison of plant-specific probabilistic safety assessments and lessons learned

Energy Technology Data Exchange (ETDEWEB)

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

2001-11-01

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

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)

Significance of earthquake risk in nuclear power plant probabilistic risk assessments

International Nuclear Information System (INIS)

Sues, R.H.; Amico, P.J.; Campbell, R.D.

1990-01-01

During the last eight years, approximately 25 utility-sponsored probabilistic risk assessments (PRAs) have been conducted for US nuclear reactors. Of these, ten have been published, seven of which have included complete seismic risk assessment. The results of the seven published PRAs are reviewed here in order to ascertain the significance of the risk due to earthquake initiating events. While PRA methodology has been in a state of development over the past seven years, and the results are subject to interpretation (as discussed in the paper), from the review conducted it is clear that earthquake-induced initiating events are important risk contributors. It is concluded that earthquake initiating events should not be dismissed, a priori, in any nuclear plant risk assessment. (orig.)

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)

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

Illustration of probabilistic approach in consequence assessment of accidental radioactive releases

International Nuclear Information System (INIS)

Pecha, P.; Hofman, R.; Kuca, P.

2009-01-01

We are describing a certain application of uncertainty analysis of environmental model HARP applied on atmospheric and deposition sub-model. Simulation of uncertainties propagation through the model is basic inevitable task bringing data for advanced techniques of probabilistic consequence assessment and further improvement of reliability of model predictions based on statistical procedures of assimilation with measured data. The activities are investigated in the institute IITA AV CR within the grant project supported by GACR (2007-2009). The problem is solved in close cooperation with section of information systems in institute NRPI. The subject of investigation concerns evaluation of consequences of radioactivity propagation after an accidental radioactivity release from nuclear facility.Transport of activity is studied from initial atmospheric propagation, deposition of radionuclides on terrain and spreading through food chains towards human body .Subsequent deposition processes of admixtures and food chain activity transport are modeled. In the final step a hazard estimation based on doses on population is integrated into the software system HARP. Extension to probabilistic approach has increased the complexity substantially, but offers much more informative background for modem methods of estimation accounting for inherent stochastic nature of the problem. Example of probabilistic assessment illustrated here is based on uncertainty analysis of input parameters of SGPM model. Predicted background field of Cs-137 deposition are labelled with index p. as P X SGPM . Final goal is estimation of a certain unknown true background vector χ true , which accounts also for deficiencies of the SGPM formulation in itself insisting in insufficient description of reality. We must have on mind, that even if we know true values of all input parameters θ m true (m= 1 ,..., M) of SGPM model, the χ true still remain uncertain. One possibility how to approach reality insists

Illustration of probabilistic approach in consequence assessment of accidental radioactive releases

International Nuclear Information System (INIS)

Pecha, P.; Hofman, R.; Kuca, P.

2008-01-01

We are describing a certain application of uncertainty analysis of environmental model HARP applied on atmospheric and deposition sub-model. Simulation of uncertainties propagation through the model is basic inevitable task bringing data for advanced techniques of probabilistic consequence assessment and further improvement of reliability of model predictions based on statistical procedures of assimilation with measured data. The activities are investigated in the institute IITA AV CR within the grant project supported by GACR (2007-2009). The problem is solved in close cooperation with section of information systems in institute NRPI. The subject of investigation concerns evaluation of consequences of radioactivity propagation after an accidental radioactivity release from nuclear facility.Transport of activity is studied from initial atmospheric propagation, deposition of radionuclides on terrain and spreading through food chains towards human body .Subsequent deposition processes of admixtures and food chain activity transport are modeled. In the final step a hazard estimation based on doses on population is integrated into the software system HARP. Extension to probabilistic approach has increased the complexity substantially, but offers much more informative background for modem methods of estimation accounting for inherent stochastic nature of the problem. Example of probabilistic assessment illustrated here is based on uncertainty analysis of input parameters of SGPM model. Predicted background field of Cs-137 deposition are labelled with index p. as P X SGPM . Final goal is estimation of a certain unknown true background vector χ true , which accounts also for deficiencies of the SGPM formulation in itself insisting in insufficient description of reality. We must have on mind, that even if we know true values of all input parameters θ m true (m= 1 ,..., M) of SGPM model, the χ true still remain uncertain. One possibility how to approach reality insists

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-08-15

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

International Nuclear Information System (INIS)

Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

1982-01-01

One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

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

Energy Technology Data Exchange (ETDEWEB)

Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

2004-06-01

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

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

International Nuclear Information System (INIS)

Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

2004-06-01

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

Probabilistic Harmonic Modeling of Wind Power Plants

DEFF Research Database (Denmark)

Guest, Emerson; Jensen, Kim H.; Rasmussen, Tonny Wederberg

2017-01-01

A probabilistic sequence domain (SD) harmonic model of a grid-connected voltage-source converter is used to estimate harmonic emissions in a wind power plant (WPP) comprised of Type-IV wind turbines. The SD representation naturally partitioned converter generated voltage harmonics into those...... with deterministic phase and those with probabilistic phase. A case study performed on a string of ten 3MW, Type-IV wind turbines implemented in PSCAD was used to verify the probabilistic SD harmonic model. The probabilistic SD harmonic model can be employed in the planning phase of WPP projects to assess harmonic...

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 performance assessment of complex energy process systems - The case of a self-sustained sanitation system.

Science.gov (United States)

Kolios, Athanasios; Jiang, Ying; Somorin, Tosin; Sowale, Ayodeji; Anastasopoulou, Aikaterini; Anthony, Edward J; Fidalgo, Beatriz; Parker, Alison; McAdam, Ewan; Williams, Leon; Collins, Matt; Tyrrel, Sean

2018-05-01

A probabilistic modelling approach was developed and applied to investigate the energy and environmental performance of an innovative sanitation system, the "Nano-membrane Toilet" (NMT). The system treats human excreta via an advanced energy and water recovery island with the aim of addressing current and future sanitation demands. Due to the complex design and inherent characteristics of the system's input material, there are a number of stochastic variables which may significantly affect the system's performance. The non-intrusive probabilistic approach adopted in this study combines a finite number of deterministic thermodynamic process simulations with an artificial neural network (ANN) approximation model and Monte Carlo simulations (MCS) to assess the effect of system uncertainties on the predicted performance of the NMT system. The joint probability distributions of the process performance indicators suggest a Stirling Engine (SE) power output in the range of 61.5-73 W with a high confidence interval (CI) of 95%. In addition, there is high probability (with 95% CI) that the NMT system can achieve positive net power output between 15.8 and 35 W. A sensitivity study reveals the system power performance is mostly affected by SE heater temperature. Investigation into the environmental performance of the NMT design, including water recovery and CO 2 /NO x emissions, suggests significant environmental benefits compared to conventional systems. Results of the probabilistic analysis can better inform future improvements on the system design and operational strategy and this probabilistic assessment framework can also be applied to similar complex engineering systems.

Probabilistic assessment of the radiological consequences of radioactive waste disposal

International Nuclear Information System (INIS)

Smith, C.F.; Cohen, J.J.

1989-01-01

Conventional methods for prediction of radiological dose consequence of low level radioactive waste (LLW) disposal generally involve application of deterministic calculational modeling. Since the selection of parametric input values for such analyses is made on a conservative ('worst case') basis, the results can be subject to criticism as being unrealistically high. To address this problem, a method for probabilistic assessment has been developed in which input parameters are expressed as probability distribution functions. An example calculation is presented for the impacts from migration of Carbon-14 to a close-in well. (author). 4 refs.; 1 tab

A probabilistic approach to quantify the uncertainties in internal dose assessment using response surface and neural network

International Nuclear Information System (INIS)

Baek, M.; Lee, S.K.; Lee, U.C.; Kang, C.S.

1996-01-01

A probabilistic approach is formulated to assess the internal radiation exposure following the intake of radioisotopes. This probabilistic approach consists of 4 steps as follows: (1) screening, (2) quantification of uncertainties, (3) propagation of uncertainties, and (4) analysis of output. The approach has been applied for Pu-induced internal dose assessment and a multi-compartment dosimetric model is used for internal transport. In this approach, surrogate models of original system are constructed using response and neural network. And the results of these surrogate models are compared with those of original model. Each surrogate model well approximates the original model. The uncertainty and sensitivity analysis of the model parameters are evaluated in this process. Dominant contributors to each organ are identified and the results show that this approach could serve a good tool of assessing the internal radiation exposure

A global probabilistic tsunami hazard assessment from earthquake sources

Science.gov (United States)

Davies, Gareth; Griffin, Jonathan; Lovholt, Finn; Glimsdal, Sylfest; Harbitz, Carl; Thio, Hong Kie; Lorito, Stefano; Basili, Roberto; Selva, Jacopo; Geist, Eric L.; Baptista, Maria Ana

2017-01-01

Large tsunamis occur infrequently but have the capacity to cause enormous numbers of casualties, damage to the built environment and critical infrastructure, and economic losses. A sound understanding of tsunami hazard is required to underpin management of these risks, and while tsunami hazard assessments are typically conducted at regional or local scales, globally consistent assessments are required to support international disaster risk reduction efforts, and can serve as a reference for local and regional studies. This study presents a global-scale probabilistic tsunami hazard assessment (PTHA), extending previous global-scale assessments based largely on scenario analysis. Only earthquake sources are considered, as they represent about 80% of the recorded damaging tsunami events. Globally extensive estimates of tsunami run-up height are derived at various exceedance rates, and the associated uncertainties are quantified. Epistemic uncertainties in the exceedance rates of large earthquakes often lead to large uncertainties in tsunami run-up. Deviations between modelled tsunami run-up and event observations are quantified, and found to be larger than suggested in previous studies. Accounting for these deviations in PTHA is important, as it leads to a pronounced increase in predicted tsunami run-up for a given exceedance rate.

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

International Nuclear Information System (INIS)

Brewer, H. Duncan; Canady, Ken S.

1999-01-01

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

Approach to modeling of human performance for purposes of probabilistic risk assessment

International Nuclear Information System (INIS)

Swain, A.D.

1983-01-01

This paper describes the general approach taken in NUREG/CR-1278 to model human performance in sufficienct detail to permit probabilistic risk assessments of nuclear power plant operations. To show the basis for the more specific models in the above NUREG, a simplified model of the human component in man-machine systems is presented, the role of performance shaping factors is discussed, and special problems in modeling the cognitive aspect of behavior are described

Dynamical systems probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ames, Arlo Leroy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-03-01

Probabilistic Risk Assessment (PRA) is the primary tool used to risk-inform nuclear power regulatory and licensing activities. Risk-informed regulations are intended to reduce inherent conservatism in regulatory metrics (e.g., allowable operating conditions and technical specifications) which are built into the regulatory framework by quantifying both the total risk profile as well as the change in the risk profile caused by an event or action (e.g., in-service inspection procedures or power uprates). Dynamical Systems (DS) analysis has been used to understand unintended time-dependent feedbacks in both industrial and organizational settings. In dynamical systems analysis, feedback loops can be characterized and studied as a function of time to describe the changes to the reliability of plant Structures, Systems and Components (SSCs). While DS has been used in many subject areas, some even within the PRA community, it has not been applied toward creating long-time horizon, dynamic PRAs (with time scales ranging between days and decades depending upon the analysis). Understanding slowly developing dynamic effects, such as wear-out, on SSC reliabilities may be instrumental in ensuring a safely and reliably operating nuclear fleet. Improving the estimation of a plant's continuously changing risk profile will allow for more meaningful risk insights, greater stakeholder confidence in risk insights, and increased operational flexibility.

Uncertainty analysis in the applications of nuclear probabilistic risk assessment

International Nuclear Information System (INIS)

Le Duy, T.D.

2011-01-01

The aim of this thesis is to propose an approach to model parameter and model uncertainties affecting the results of risk indicators used in the applications of nuclear Probabilistic Risk assessment (PRA). After studying the limitations of the traditional probabilistic approach to represent uncertainty in PRA model, a new approach based on the Dempster-Shafer theory has been proposed. The uncertainty analysis process of the proposed approach consists in five main steps. The first step aims to model input parameter uncertainties by belief and plausibility functions according to the data PRA model. The second step involves the propagation of parameter uncertainties through the risk model to lay out the uncertainties associated with output risk indicators. The model uncertainty is then taken into account in the third step by considering possible alternative risk models. The fourth step is intended firstly to provide decision makers with information needed for decision making under uncertainty (parametric and model) and secondly to identify the input parameters that have significant uncertainty contributions on the result. The final step allows the process to be continued in loop by studying the updating of beliefs functions given new data. The proposed methodology was implemented on a real but simplified application of PRA model. (author)

A probabilistic risk assessment for field radiography based on expert judgment and opinion

International Nuclear Information System (INIS)

Jang, Han-Ki; Ryu, Hyung-Joon; Kim, Ji-Young; Lee, Jai-Ki; Cho, Kun-Woo

2011-01-01

A probabilistic approach was applied to assess radiation risk associated with the field radiography using gamma sources. The Delphi method based on the expert judgments and opinions was used in the process of characterization of parameters affecting risk, which are inevitably subject to large uncertainties. A mathematical approach applying the Bayesian inferences was employed for data processing to improve the Delphi results. This process consists of three phases: (1) setting prior distributions, (2) constructing the likelihood functions and (3) deriving the posterior distributions based on the likelihood functions. The approach for characterizing input parameters using the Bayesian inference is provided for improved risk estimates without intentional rejection of part of the data, which demonstrated utility of Bayesian updating of distributions of uncertain input parameters in PRA (Probabilistic Risk Assessment). The data analysis portion for PRA in field radiography is addressed for estimates of the parameters used to determine the frequencies and consequences of the various events modeled. In this study, radiological risks for the worker and the public member in the vicinity of the work place are estimated for field radiography system in Korea based on two-dimensional Monte Carlo Analysis (2D MCA). (author)

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.)

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)

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

Exploring probabilistic tools for the development of a platform for Quantitative Risk Assessment (QRA) of hydro-meteorological hazards in Europe

Science.gov (United States)

Zumpano, V.; Hussin, H. Y.; Breinl, K.

2012-04-01

Mass-movements and floods are hydro-meteorological hazards that can have catastrophic effects on communities living in mountainous areas prone to these disastrous events. Environmental, climate and socio-economic changes are expected to affect the tempo-spatial patterns of hydro-meteorological hazards and associated risks in Europe. These changes and their effects on the occurrence of future hazards need to be analyzed and modeled using probabilistic hazard and risk assessment methods in order to assist stakeholders in disaster management strategies and policy making. Quantitative Risk Assessment (QRA) using probabilistic methods can further calculate damage and losses to multi-hazards and determine the uncertainties related to all the probabilistic components of the hazard and the vulnerability of the elements at risk. Therefore, in order to develop an effective platform that can quantitatively calculate the risk of mass-movements and floods in several European test sites, an extensive inventory and analysis has been carried out of the available tools and software related to the probabilistic risk assessment of single and multi-hazards. The tools have been reviewed based on whether they are open source and freely available, their required input data, the availability and type of hazard and vulnerability modules, transparency of methods used, their validation and calibration techniques, the inclusion of uncertainties and their state of the art. The analysis also specially focused on the applicability of the tools to European study areas. The findings showed that assumptions and simplifications are made when assessing and quantifying the hazards. The interaction between multiple hazards, like cascading effects are not assessed in most tools and some consider the hazard and vulnerability as qualitative components, rather than quantitative ones. This analysis of hazard and risk assessment tools and software will give future developers and experts a better overview of

Advanced Seismic Probabilistic Risk Assessment Demonstration Project Plan

Energy Technology Data Exchange (ETDEWEB)

Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

Idaho National Laboratories (INL) has an ongoing research and development (R&D) project to remove excess conservatism from seismic probabilistic risk assessments (SPRA) calculations. These risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. This report presents a plan for improving our current traditional SPRA process using a seismic event recorded at a nuclear power plant site, with known outcomes, to improve the decision making process. SPRAs are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in general this approach has been conservative, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility).

Probabilistic Simulation of Multi-Scale Composite Behavior

Science.gov (United States)

Chamis, Christos C.

2012-01-01

A methodology is developed to computationally assess the non-deterministic composite response at all composite scales (from micro to structural) due to the uncertainties in the constituent (fiber and matrix) properties, in the fabrication process and in structural variables (primitive variables). The methodology is computationally efficient for simulating the probability distributions of composite behavior, such as material properties, laminate and structural responses. Bi-products of the methodology are probabilistic sensitivities of the composite primitive variables. The methodology has been implemented into the computer codes PICAN (Probabilistic Integrated Composite ANalyzer) and IPACS (Integrated Probabilistic Assessment of Composite Structures). The accuracy and efficiency of this methodology are demonstrated by simulating the uncertainties in composite typical laminates and comparing the results with the Monte Carlo simulation method. Available experimental data of composite laminate behavior at all scales fall within the scatters predicted by PICAN. Multi-scaling is extended to simulate probabilistic thermo-mechanical fatigue and to simulate the probabilistic design of a composite redome in order to illustrate its versatility. Results show that probabilistic fatigue can be simulated for different temperature amplitudes and for different cyclic stress magnitudes. Results also show that laminate configurations can be selected to increase the redome reliability by several orders of magnitude without increasing the laminate thickness--a unique feature of structural composites. The old reference denotes that nothing fundamental has been done since that time.

The european flood alert system EFAS – Part 2: Statistical skill assessment of probabilistic and deterministic operational forecasts

Directory of Open Access Journals (Sweden)

J. C. Bartholmes

2009-02-01

Full Text Available Since 2005 the European Flood Alert System (EFAS has been producing probabilistic hydrological forecasts in pre-operational mode at the Joint Research Centre (JRC of the European Commission. EFAS aims at increasing preparedness for floods in trans-national European river basins by providing medium-range deterministic and probabilistic flood forecasting information, from 3 to 10 days in advance, to national hydro-meteorological services.

This paper is Part 2 of a study presenting the development and skill assessment of EFAS. In Part 1, the scientific approach adopted in the development of the system has been presented, as well as its basic principles and forecast products. In the present article, two years of existing operational EFAS forecasts are statistically assessed and the skill of EFAS forecasts is analysed with several skill scores. The analysis is based on the comparison of threshold exceedances between proxy-observed and forecasted discharges. Skill is assessed both with and without taking into account the persistence of the forecasted signal during consecutive forecasts.

Skill assessment approaches are mostly adopted from meteorology and the analysis also compares probabilistic and deterministic aspects of EFAS. Furthermore, the utility of different skill scores is discussed and their strengths and shortcomings illustrated. The analysis shows the benefit of incorporating past forecasts in the probability analysis, for medium-range forecasts, which effectively increases the skill of the forecasts.

A probabilistic assessment of health risks associated with short-term exposure to tropospheric ozone

Energy Technology Data Exchange (ETDEWEB)

Whitfield, R.G; Biller, W.F.; Jusko, M.J.; Keisler, J.M.

1996-06-01

The work described in this report is part of a larger risk assessment sponsored by the U.S. Environmental Protection Agency. Earlier efforts developed exposure-response relationships for acute health effects among populations engaged in heavy exertion. Those efforts also developed a probabilistic national ambient air quality standards exposure model and a general methodology for integrating probabilistic exposure-response relation- ships and exposure estimates to calculate overall risk results. Recently published data make it possible to model additional health endpoints (for exposure at moderate exertion), including hospital admissions. New air quality and exposure estimates for alternative national ambient air quality standards for ozone are combined with exposure-response models to produce the risk results for hospital admissions and acute health effects. Sample results explain the methodology and introduce risk output formats.

Probabilistic fire risk assessment for Koeberg Nuclear Power Station Unit 1

International Nuclear Information System (INIS)

Grobbelaar, J.F.; Foster, N.A.S.; Luesse, L.J.

1995-01-01

A probabilistic fire risk assessment was done for Koeberg Nuclear Power Station Unit 1. Areas where fires are likely to start were identified. Equipment important to safety, as well as their power and/or control cable routes were identified in each fire confinement sector. Fire confinement sectors where internal initiating events could be caused by fire were identified. Detection failure and suppression failure fault trees and event trees were constructed. The core damage frequency associated with each fire confinement sector was calculated, and important fire confinement sectors were identified. (author)

A toolkit for integrated deterministic and probabilistic assessment for hydrogen infrastructure.

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina M.; Tchouvelev, Andrei V.

2014-03-01

There has been increasing interest in using Quantitative Risk Assessment [QRA] to help improve the safety of hydrogen infrastructure and applications. Hydrogen infrastructure for transportation (e.g. fueling fuel cell vehicles) or stationary (e.g. back-up power) applications is a relatively new area for application of QRA vs. traditional industrial production and use, and as a result there are few tools designed to enable QRA for this emerging sector. There are few existing QRA tools containing models that have been developed and validated for use in small-scale hydrogen applications. However, in the past several years, there has been significant progress in developing and validating deterministic physical and engineering models for hydrogen dispersion, ignition, and flame behavior. In parallel, there has been progress in developing defensible probabilistic models for the occurrence of events such as hydrogen release and ignition. While models and data are available, using this information is difficult due to a lack of readily available tools for integrating deterministic and probabilistic components into a single analysis framework. This paper discusses the first steps in building an integrated toolkit for performing QRA on hydrogen transportation technologies and suggests directions for extending the toolkit.

Risk management on nuclear power plant. Application of probabilistic risk assessment

International Nuclear Information System (INIS)

Kojima, Shigeo

2003-01-01

In U.S.A., nuclear safety regulation is moving to risk-informed regulation (RIR), so necessity of a standard to provide contents of probabilistic risk assessment (PRA) constructing its roots has been discussed for a long time. In 1998, the Committee on Nuclear Risk Management (CNRM) of the American Society of Mechanical Engineers (ASME) began to investigate the standard, of which last edition was published as the Standard for Probabilistic Risk Management for Nuclear Power Plant Applications: RA-S-2002 (PRMA) on April, 2002. As in the Committee, the Nuclear Regulatory Commission (NRC), electric power companies, national institutes, PRA specialists, and so on took parts to carry out many discussions with full energies of participants on risk management in U.S.A., the standard was finished after about four years' efforts. In U.S.A., risk management having already used PRA is successfully practiced, U.S.A. is at a stage with more advancing steps of the risk management than Japan is. Here was described on the standard of PRA and a concrete method of the risk management carried out at nuclear power stations. (G.K.)

77 FR 29391 - An Approach for Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific...

Science.gov (United States)

2012-05-17

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0110] An Approach for Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment. SUMMARY: The U.S. Nuclear Regulatory...

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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

Measurement of the reaction {gamma}p{yields}K{sup 0}{sigma}{sup +} for photon energies up to 2.65 GeV with the SAPHIR detector at ELSA; Messung der Reaktion {gamma}p {yields} K{sup 0}{sigma}{sup +} fuer Photonenergien bis 2.65 GeV mit dem SAPHIR-Detektor an ELSA

Energy Technology Data Exchange (ETDEWEB)

Lawall, R.

2004-01-01

The reaction {gamma}p {yields} K{sup 0}{sigma}{sup +} was measured with the SAPHIR-detector at ELSA during the run periods 1997 and 1998. Results were obtained for cross sections in the photon energy range from threshold up to 2.65 GeV for all production angles and for the {sigma}{sup +}-polarization. Emphasis has been put on the determination and reduction of the contributions of background reactions and the comparison with other measurements and theoretical predictions. (orig.)

Evaluation of structural fragilities for an IPEEE seismic probabilistic risk assessment study

International Nuclear Information System (INIS)

Ghiocel, D.M.; Wilson, P.R.; Stevenson, J.D.

1995-01-01

The paper presents the main issues and results of a structural fragility analysis for a Seismic Probabilistic Risk Assessment (SPRA) study of a nuclear power plant (NPP) in the Eastern US. The fragility evaluations were performed for the Reactor Building, Auxiliary Building, Intake Structure and Diesel Generator Building. The random seismic input is defined in terms of the Uniform Hazard Spectrum (UHS) earthquake on the NPP site anchored to a reference level of 0.40 g Zero Period Ground Acceleration (ZPGA). Because of the soft soil conditions new Soil-Structure Interaction (SSI) analyses were performed using the original finite element (stick) structural models and the complex frequency approach. The soil deposit randomness was described by the variations in both the low strain soil shear modules and in its dependence with the shear strain. The probabilistic SSI analyses were performed using digital simulation techniques. The critical failure modes for each structure are investigated and the fragility evaluations are discussed. Concluding remarks and recommendations for improving the quality of the structural fragility analyses are included

A Framework for Probabilistic Multi-Hazard Assessment of Rain-Triggered Lahars Using Bayesian Belief Networks

Directory of Open Access Journals (Sweden)

Pablo Tierz

2017-09-01

Full Text Available Volcanic water-sediment flows, commonly known as lahars, can often pose a higher threat to population and infrastructure than primary volcanic hazardous processes such as tephra fallout and Pyroclastic Density Currents (PDCs. Lahars are volcaniclastic flows of water, volcanic debris and entrained sediments that can travel long distances from their source, causing severe damage by impact and burial. Lahars are frequently triggered by intense or prolonged rainfall occurring after explosive eruptions, and their occurrence depends on numerous factors including the spatio-temporal rainfall characteristics, the spatial distribution and hydraulic properties of the tephra deposit, and the pre- and post-eruption topography. Modeling (and forecasting such a complex system requires the quantification of aleatory variability in the lahar triggering and propagation. To fulfill this goal, we develop a novel framework for probabilistic hazard assessment of lahars within a multi-hazard environment, based on coupling a versatile probabilistic model for lahar triggering (a Bayesian Belief Network: Multihaz with a dynamic physical model for lahar propagation (LaharFlow. Multihaz allows us to estimate the probability of lahars of different volumes occurring by merging varied information about regional rainfall, scientific knowledge on lahar triggering mechanisms and, crucially, probabilistic assessment of available pyroclastic material from tephra fallout and PDCs. LaharFlow propagates the aleatory variability modeled by Multihaz into hazard footprints of lahars. We apply our framework to Somma-Vesuvius (Italy because: (1 the volcano is strongly lahar-prone based on its previous activity, (2 there are many possible source areas for lahars, and (3 there is high density of population nearby. Our results indicate that the size of the eruption preceding the lahar occurrence and the spatial distribution of tephra accumulation have a paramount role in the lahar

Development of a Risk-Based Probabilistic Performance-Assessment Method for Long-Term Cover Systems - 2nd Edition

International Nuclear Information System (INIS)

HO, CLIFFORD K.; ARNOLD, BILL W.; COCHRAN, JOHN R.; TAIRA, RANDAL Y.

2002-01-01

A probabilistic, risk-based performance-assessment methodology has been developed to assist designers, regulators, and stakeholders in the selection, design, and monitoring of long-term covers for contaminated subsurface sites. This report describes the method, the software tools that were developed, and an example that illustrates the probabilistic performance-assessment method using a repository site in Monticello, Utah. At the Monticello site, a long-term cover system is being used to isolate long-lived uranium mill tailings from the biosphere. Computer models were developed to simulate relevant features, events, and processes that include water flux through the cover, source-term release, vadose-zone transport, saturated-zone transport, gas transport, and exposure pathways. The component models were then integrated into a total-system performance-assessment model, and uncertainty distributions of important input parameters were constructed and sampled in a stochastic Monte Carlo analysis. Multiple realizations were simulated using the integrated model to produce cumulative distribution functions of the performance metrics, which were used to assess cover performance for both present- and long-term future conditions. Performance metrics for this study included the water percolation reaching the uranium mill tailings, radon gas flux at the surface, groundwater concentrations, and dose. Results from uncertainty analyses, sensitivity analyses, and alternative design comparisons are presented for each of the performance metrics. The benefits from this methodology include a quantification of uncertainty, the identification of parameters most important to performance (to prioritize site characterization and monitoring activities), and the ability to compare alternative designs using probabilistic evaluations of performance (for cost savings)

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

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 Decision Graphs - Combining Verification and AI Techniques for Probabilistic Inference

DEFF Research Database (Denmark)

Jaeger, Manfred

2004-01-01

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

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.

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

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

Learning Probabilistic Logic Models from Probabilistic Examples.

Science.gov (United States)

Chen, Jianzhong; Muggleton, Stephen; Santos, José

2008-10-01

We revisit an application developed originally using abductive Inductive Logic Programming (ILP) for modeling inhibition in metabolic networks. The example data was derived from studies of the effects of toxins on rats using Nuclear Magnetic Resonance (NMR) time-trace analysis of their biofluids together with background knowledge representing a subset of the Kyoto Encyclopedia of Genes and Genomes (KEGG). We now apply two Probabilistic ILP (PILP) approaches - abductive Stochastic Logic Programs (SLPs) and PRogramming In Statistical modeling (PRISM) to the application. Both approaches support abductive learning and probability predictions. Abductive SLPs are a PILP framework that provides possible worlds semantics to SLPs through abduction. Instead of learning logic models from non-probabilistic examples as done in ILP, the PILP approach applied in this paper is based on a general technique for introducing probability labels within a standard scientific experimental setting involving control and treated data. Our results demonstrate that the PILP approach provides a way of learning probabilistic logic models from probabilistic examples, and the PILP models learned from probabilistic examples lead to a significant decrease in error accompanied by improved insight from the learned results compared with the PILP models learned from non-probabilistic examples.

Probabilistic safety analysis procedures guide

International Nuclear Information System (INIS)

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

1984-01-01

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

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.

Innovative real time simulation training and nuclear probabilistic risk assessment

International Nuclear Information System (INIS)

Reisinger, M.F.

1991-01-01

Operator errors have been an area of public concern for the safe operation of nuclear power plants since the TMI2 incident. Simply stated, nuclear plants are very complex systems and the public is skeptical of the operators' ability to comprehend and deal with the vast indications and complexities of potential nuclear power plant events. Prior to the TMI2 incident, operator errors and human factors were not included as contributing factors in the Probabilistic Risk Assessment (PRA) studies of nuclear power plant accidents. More recent efforts in nuclear risk assessment have addressed some of the human factors affecting safe nuclear plant operations. One study found four major factors having significant impact on operator effectiveness. This paper discusses human factor PRAs, new applications in simulation training and the specific potential benefits from simulation in promoting safer operation of future power plants as well as current operating power plants

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

Science.gov (United States)

Guarro, Sergio B.

2010-01-01

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

A shortened version of the THERP/Handbook approach to human reliability analysis for probabilistic risk assessment

International Nuclear Information System (INIS)

Swain, A.D.

1986-01-01

The approach to human reliability analysis (HRA) known as THERP/Handbook has been applied to several probabilistic risk assessments (PRAs) of nuclear power plants (NPPs) and other complex systems. The approach is based on a thorough task analysis of the man-machine interfaces, including the interactions among the people, involved in the operations being assessed. The idea is to assess fully the underlying performance shaping factors (PSFs) and dependence effects which result either in reliable or unreliable human performance

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)

PROBABILISTIC SEISMIC ASSESSMENT OF BASE-ISOLATED NPPS SUBJECTED TO STRONG GROUND MOTIONS OF TOHOKU EARTHQUAKE

Directory of Open Access Journals (Sweden)

AHMER ALI

2014-10-01

Full Text Available The probabilistic seismic performance of a standard Korean nuclear power plant (NPP with an idealized isolation is investigated in the present work. A probabilistic seismic hazard analysis (PSHA of the Wolsong site on the Korean peninsula is performed by considering peak ground acceleration (PGA as an earthquake intensity measure. A procedure is reported on the categorization and selection of two sets of ground motions of the Tohoku earthquake, i.e. long-period and common as Set A and Set B respectively, for the nonlinear time history response analysis of the base-isolated NPP. Limit state values as multiples of the displacement responses of the NPP base isolation are considered for the fragility estimation. The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.

Probabilistic seismic assessment of base-isolated NPPs subjected to strong ground motions of Tohoku earthquake

Energy Technology Data Exchange (ETDEWEB)

Ali, Ahmer; Hayah, Nadin Abu; Kim, Doo Kie [Dept. of Civil and Environmental Engineering, Kunsan National University, Kunsan (Korea, Republic of); Cho, Sung Gook [R and D Center, JACE KOREA Company, Gyeonggido (Korea, Republic of)

2014-10-15

The probabilistic seismic performance of a standard Korean nuclear power plant (NPP) with an idealized isolation is investigated in the present work. A probabilistic seismic hazard analysis (PSHA) of the Wolsong site on the Korean peninsula is performed by considering peak ground acceleration (PGA) as an earthquake intensity measure. A procedure is reported on the categorization and selection of two sets of ground motions of the Tohoku earthquake, i.e. long-period and common as Set A and Set B respectively, for the nonlinear time history response analysis of the base-isolated NPP. Limit state values as multiples of the displacement responses of the NPP base isolation are considered for the fragility estimation. The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.

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

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.

Modelling fog in probabilistic consequence assessment

International Nuclear Information System (INIS)

Underwood, B.Y.

1993-02-01

Earlier work examined the potential influence of foggy weather conditions on the probabilistic assessment of the consequences of accidental releases of radioactive material to the atmosphere (PCA), in particular the impact of a fraction of the released aerosol becoming incorporated into droplets. A major uncertainty emerging from the initial scoping study concerned estimation of the fraction of the released material that would be taken up into droplets. An objective is to construct a method for handling in a PCA context the effect of fog on deposition, basing the method on the experience gained from prior investigations. There are two aspects to explicitly including the effect of fog in PCA: estimating the probability of occurrence of various types of foggy condition and calculating the impact on the conventional end-points of consequence assessment. For the first, a brief outline is given of the use of meteorological data by PCA computer codes, followed by a discussion of some routinely-recorded meteorological parameters that are pertinent to fog, such as the presentweather code and horizontal visibility. Four stylized scenarios are defined to cover a wide range of situations in which particle growth by uptake of water may have an important impact on deposition. A description is then given of the way in which routine meteorological data could be used to flag the presence of each of these conditions in the meteorological data file used by the PCA code. The approach developed to calculate the impact on deposition is pitched at a level of complexity appropriate to the PCA context and reflects the physical constraints of the system and accounts for the specific characteristics of the released aerosol. (Author)

Application of probabilistic risk assessment methodology to fusion

International Nuclear Information System (INIS)

Piet, S.J.

1985-07-01

Probabilistic Risk Assessment (PRA) tools are applied to general fusion issues in a systematic way, generally qualitatively. The potential value of PRA to general fusion safety and economic issues is discussed. Several important design insights result: possible fault interactions must be minimized (decouple fault conditions), inherently safe designs must include provision for passively handling loss of site power and loss of coolant conditions, the reliability of the vacuum boundary appears vital to maximizing facility availabilty and minimizing safety risk, and economic analyses appear to be incomplete without consideration of potential availability loss from forced outrages. A modification to PRA formalism is introduced, called the fault interaction matrix. The fault interaction matrix contains information concerning what initial fault condition could lead to other fault conditions and with what frequency. Thus, the fault interaction matrix represents a way to present and measure the degree to which a designer has decoupled possible fault conditions in his design

A Probabilistic Assessment of the Next Geomagnetic Reversal

Science.gov (United States)

Buffett, Bruce; Davis, William

2018-02-01

Deterministic forecasts for the next geomagnetic reversal are not feasible due to large uncertainties in the present-day state of the Earth's core. A more practical approach relies on probabilistic assessments using paleomagnetic observations to characterize the amplitude of fluctuations in the geomagnetic dipole. We use paleomagnetic observations for the past 2 Myr to construct a stochastic model for the axial dipole field and apply well-established methods to evaluate the probability of the next geomagnetic reversal as a function of time. For a present-day axial dipole moment of 7.6 × 1022 A m2, the probability of the dipole entering a reversed state is less than 2% after 20 kyr. This probability rises to 11% after 50 kyr. An imminent geomagnetic reversal is not supported by paleomagnetic observations. The current rate of decline in the dipole moment is unusual but within the natural variability predicted by the stochastic model.

Advanced neutron source reactor probabilistic flow blockage assessment

International Nuclear Information System (INIS)

Ramsey, C.T.

1995-08-01

The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

Probabilistic health risk assessment for arsenic intake through drinking groundwater in Taiwan's Pingtung Plain

Science.gov (United States)

Liang, C. P.; Chen, J. S.

2017-12-01

An abundant and inexpensive supply of groundwater is used to meet drinking, agriculture and aquaculture requirements of the residents in the Pingtung Plain. Long-term groundwater quality monitoring data indicate that the As content in groundwater in the Pingtung Plain exceeds the maximum level of 10 g/L recommended by the World Health Organization (WHO). The situation is further complicated by the fact that only 46.89% of population in the Pingtung Plain has been served with tap water, far below the national average of 92.93%. Considering there is a considerable variation in the measured concentrations, from below the detection limit (consumption rate and body weight of the individual, the conventional approach to conducting a human health risk assessment may be insufficient for health risk management. This study presents a probabilistic risk assessment for inorganic As intake through the consumption of the drinking groundwater by local residents in the Pingtung Plain. The probabilistic risk assessment for inorganic As intake through the consumption of the drinking groundwater is achieved using Monte Carlo simulation technique based on the hazard quotient (HQ) and target cancer risk (TR) established by the U.S. Environmental Protection Agency. This study demonstrates the importance of the individual variability of inorganic As intake through drinking groundwater consumption when evaluating a high exposure sub-group of the population who drink high As content groundwater.

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

Using Probabilistic Methods in Water Scarcity Assessments: A First Step Towards a Water Scarcity Risk Assessment Framework

Science.gov (United States)

Veldkamp, Ted; Wada, Yoshihide; Aerts, Jeroen; Ward, Phillip

2016-01-01

Water scarcity -driven by climate change, climate variability, and socioeconomic developments- is recognized as one of the most important global risks, both in terms of likelihood and impact. Whilst a wide range of studies have assessed the role of long term climate change and socioeconomic trends on global water scarcity, the impact of variability is less well understood. Moreover, the interactions between different forcing mechanisms, and their combined effect on changes in water scarcity conditions, are often neglected. Therefore, we provide a first step towards a framework for global water scarcity risk assessments, applying probabilistic methods to estimate water scarcity risks for different return periods under current and future conditions while using multiple climate and socioeconomic scenarios.

A probabilistic assessment of calcium carbonate export and dissolution in the modern ocean

OpenAIRE

Battaglia Gianna; Steinacher Marco; Joos Fortunat

2016-01-01

The marine cycle of calcium carbonate (CaCO3) is an important element of the carbon cycle and co-governs the distribution of carbon and alkalinity within the ocean. However, CaCO3 export fluxes and mechanisms governing CaCO3 dissolution are highly uncertain. We present an observationally constrained, probabilistic assessment of the global and regional CaCO3 budgets. Parameters governing pelagic CaCO3 export fluxes and dissolution rates are sampled using a Monte Carlo sche...

Overview of seismic probabilistic risk assessment for structural analysis in nuclear facilities

International Nuclear Information System (INIS)

Reed, J.W.

1989-01-01

Probabilistic Risk Assessment (PRA) for seismic events is currently being performed for nuclear and DOE facilities. The background on seismic PRA is presented along with a basic description of the method. The seismic PRA technique is applicable to other critical facilities besides nuclear plants. The different approaches for obtained structure fragility curves are discussed and their applications to structures and equipment, in general, are addressed. It is concluded that seismic PRA is a useful technique for conducting probability analysis for a wide range of classes of structures and equipment

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

Assessing Probabilistic Reasoning in Verbal-Numerical and Graphical-Pictorial Formats: An Evaluation of the Psychometric Properties of an Instrument

Science.gov (United States)

Agus, Mirian; Penna, Maria Pietronilla; Peró-Cebollero, Maribel; Guàrdia-Olmos, Joan

2016-01-01

Research on the graphical facilitation of probabilistic reasoning has been characterised by the effort expended to identify valid assessment tools. The authors developed an assessment instrument to compare reasoning performances when problems were presented in verbal-numerical and graphical-pictorial formats. A sample of undergraduate psychology…

PRA (Probabilistic Risk Assessments) Participation versus Validation

Science.gov (United States)

DeMott, Diana; Banke, Richard

2013-01-01

Probabilistic Risk Assessments (PRAs) are performed for projects or programs where the consequences of failure are highly undesirable. PRAs primarily address the level of risk those projects or programs posed during operations. PRAs are often developed after the design has been completed. Design and operational details used to develop models include approved and accepted design information regarding equipment, components, systems and failure data. This methodology basically validates the risk parameters of the project or system design. For high risk or high dollar projects, using PRA methodologies during the design process provides new opportunities to influence the design early in the project life cycle to identify, eliminate or mitigate potential risks. Identifying risk drivers before the design has been set allows the design engineers to understand the inherent risk of their current design and consider potential risk mitigation changes. This can become an iterative process where the PRA model can be used to determine if the mitigation technique is effective in reducing risk. This can result in more efficient and cost effective design changes. PRA methodology can be used to assess the risk of design alternatives and can demonstrate how major design changes or program modifications impact the overall program or project risk. PRA has been used for the last two decades to validate risk predictions and acceptability. Providing risk information which can positively influence final system and equipment design the PRA tool can also participate in design development, providing a safe and cost effective product.

International Space Station End-of-Life Probabilistic Risk Assessment

Science.gov (United States)

Duncan, Gary W.

2014-01-01

The International Space Station (ISS) end-of-life (EOL) cycle is currently scheduled for 2020, although there are ongoing efforts to extend ISS life cycle through 2028. The EOL for the ISS will require deorbiting the ISS. This will be the largest manmade object ever to be de-orbited therefore safely deorbiting the station will be a very complex problem. This process is being planned by NASA and its international partners. Numerous factors will need to be considered to accomplish this such as target corridors, orbits, altitude, drag, maneuvering capabilities etc. The ISS EOL Probabilistic Risk Assessment (PRA) will play a part in this process by estimating the reliability of the hardware supplying the maneuvering capabilities. The PRA will model the probability of failure of the systems supplying and controlling the thrust needed to aid in the de-orbit maneuvering.

Development and Application of a Probabilistic Risk-Benefit Assessment Model for Infant Feeding Integrating Microbiological, Nutritional, and Chemical Components.

Science.gov (United States)

Boué, Géraldine; Cummins, Enda; Guillou, Sandrine; Antignac, Jean-Philippe; Le Bizec, Bruno; Membré, Jeanne-Marie

2017-12-01

A probabilistic and interdisciplinary risk-benefit assessment (RBA) model integrating microbiological, nutritional, and chemical components was developed for infant milk, with the objective of predicting the health impact of different scenarios of consumption. Infant feeding is a particular concern of interest in RBA as breast milk and powder infant formula have both been associated with risks and benefits related to chemicals, bacteria, and nutrients, hence the model considers these three facets. Cronobacter sakazakii, dioxin-like polychlorinated biphenyls (dl-PCB), and docosahexaenoic acid (DHA) were three risk/benefit factors selected as key issues in microbiology, chemistry, and nutrition, respectively. The present model was probabilistic with variability and uncertainty separated using a second-order Monte Carlo simulation process. In this study, advantages and limitations of undertaking probabilistic and interdisciplinary RBA are discussed. In particular, the probabilistic technique was found to be powerful in dealing with missing data and to translate assumptions into quantitative inputs while taking uncertainty into account. In addition, separation of variability and uncertainty strengthened the interpretation of the model outputs by enabling better consideration and distinction of natural heterogeneity from lack of knowledge. Interdisciplinary RBA is necessary to give more structured conclusions and avoid contradictory messages to policymakers and also to consumers, leading to more decisive food recommendations. This assessment provides a conceptual development of the RBA methodology and is a robust basis on which to build upon. © 2017 Society for Risk Analysis.

Scenario for a Short-Term Probabilistic Seismic Hazard Assessment (PSHA in Chiayi, Taiwan

Directory of Open Access Journals (Sweden)

Chung-Han Chan

2013-01-01

Full Text Available Using seismic activity and the Meishan earthquake sequence that occurred from 1904 to 1906, a scenario for short-term probabilistic seismic hazards in the Chiayi region of Taiwan is assessed. The long-term earthquake occurrence rate in Taiwan was evaluated using a smoothing kernel. The highest seismicity rate was calculated around the Chiayi region. To consider earthquake interactions, the rate-and-state friction model was introduced to estimate the seismicity rate evolution due to the Coulomb stress change. As imparted by the 1904 Touliu earthquake, stress changes near the 1906 Meishan and Yangshuigang epicenters was higher than the magnitude of tidal triggering. With regard to the impact of the Meishan earthquake, the region close to the Yangshuigang earthquake epicenter had a +0.75 bar stress increase. The results indicated significant interaction between the three damage events. Considering the path and site effect using ground motion prediction equations, a probabilistic seismic hazard in the form of a hazard evolution and a hazard map was assessed. A significant elevation in hazards following the three earthquakes in the sequence was determined. The results illustrate a possible scenario for seismic hazards in the Chiayi region which may take place repeatly in the future. Such scenario provides essential information on earthquake preparation, devastation estimations, emergency sheltering, utility restoration, and structure reconstruction.

Implementation of condition-dependent probabilistic risk assessment using surveillance data on passive components

International Nuclear Information System (INIS)

Lewandowski, Radoslaw; Denning, Richard; Aldemir, Tunc; Zhang, Jinsuo

2016-01-01

Highlights: • Condition-dependent probabilistic risk assessment (PRA). • Time-dependent characterization of plant-specific risk. • Containment bypass involving in secondary system piping and SCC in SG tubes. - Abstract: A great deal of surveillance data are collected for a nuclear power plant that reflect the changing condition of the plant as it ages. Although surveillance data are used to determine failure probabilities of active components for the plant’s probabilistic risk assessment (PRA) and to indicate the need for maintenance activities, they are not used in a structured manner to characterize the evolving risk of the plant. The present study explores the feasibility of using a condition-dependent PRA framework that takes a first principles approach to modeling the progression of degradation mechanisms to characterize evolving risk, periodically adapting the model to account for surveillance results. A case study is described involving a potential containment bypass accident sequence due to the progression of flow-accelerated corrosion in secondary system piping and stress corrosion cracking of steam generator tubes. In this sequence, a steam line break accompanied by failure to close of a main steam isolation valve results in depressurization of the steam generator and induces the rupture of one or more faulted steam generator tubes. The case study indicates that a condition-dependent PRA framework might be capable of providing early identification of degradation mechanisms important to plant risk.

Probabilistic risk assessment: A look at the role of artificial intelligence

International Nuclear Information System (INIS)

Wang, J.; Modarres, M.; Hunt, R.N.M.

1988-01-01

A review of traditional Probabilistic Risk Assessment (PRA) methods used in the nuclear power industry is presented. The shortcomings of the current PRA methods are pointed out. A method of performing a PRA is proposed and is computerized. The role of artificial intelligence in developing and performing the proposed PRA approach is discussed. The proposed PRA approach is verified by comparing the results to previously performed PRAs. The comparisons have supported the adequacy and completeness of the results of the proposed model. A discussion of how the proposed method can be used as an expert system to verify plant status following loss of plant hardware is also presented. (orig.)

Probabilistic risk assessment in the CPI

International Nuclear Information System (INIS)

Guymer, P.; Kaiser, G.D.; Mc Kelvey, T.C.; Hannaman, G.W.

1987-01-01

Probabilistic Risk Assessment (PRA) is a method of quantifying the frequency of occurrence and magnitude of the consequences of accidents in systems that contain hazardous materials such as toxic, flammable or explosive chemicals. The frequency and magnitude of the consequences are the basic elements in the definition of risk, often simply expressed as the product of frequency and magnitude, summed over all accident sequences. PRA is a mature technique that has been used to estimate risk for a number of industrial facilities: for example, the Canvey Island Petrochemical complex; the Port of Rotterdam; the Reactor Safety Study, the first study to put the risks associated with nuclear power into perspective; and the transportation of chlorine. PRA has been developed to a greater level of sophistication in the nuclear industry than in the chemical industry. In the nuclear area, its usefulness has been demonstrated by increased plant safety, engineering insights, and cost-saving recommendations. Data and methods have been developed to increase the level of realism of the treatment of operator actions in PRA studies. It can be stated generally that the same methods can be applied with equal success in the chemical industry. However, there are pitfalls into which the unwary nuclear-oriented PRA analyst may stumble if he does not bear in mind that there are significant differences between nuclear plants and chemical plants

Probabilistic methodology for assessing steam generator tube inspection - Phase II: CANTIA - a probabilistic method for assessing steam generator tube inspections

International Nuclear Information System (INIS)

Harris, J.E.; Gorman, J.A.; Turner, A.P.L.

1997-03-01

The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the documentation and verification of the code is provided in this volume. The user's manual for CANTIA is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

Expert elicitation for deriving input data for probabilistic risk assessment of shipwrecks.

Science.gov (United States)

Landquist, H; Norrman, J; Lindhe, A; Norberg, T; Hassellöv, I-M; Lindgren, J F; Rosén, L

2017-12-15

The necessity of having a process in place for adequate risk assessment of shipwrecks that pose a threat to the marine environment is today internationally acknowledged. However, retrieving the desired data for such a risk assessment can prove challenging. One means of addressing this problem is to make use of experts' knowledge and experience. The purpose of this paper is therefore to present and analyse data for risk assessment of shipwrecks derived by expert elicitation. The main outcome is the experts' estimations of (i) the generic probability of an opening in a shipwreck due to the occurrence of a number of activities and (ii) estimations of the degree to which site-specific and wreck-specific indicators affect the probability of opening. Results show that the derived information is applicable in probabilistic shipwreck risk assessment and that the VRAKA framework now contains needed information for integrating generic and site-specific information using Bayesian updating. Copyright © 2017 Elsevier Ltd. All rights reserved.

Review insights on the probabilistic risk assessment for the Limerick Generating Station

International Nuclear Information System (INIS)

1984-08-01

In recognition of the high population density around the Limerick Generating Station site and the proposed power level, the Philadelphia Electric Company, in response to NRC staff requests, conducted and submitted between March 1981 and November 1983 a probabilistic risk assessment (PRA) on internal event contributors and a severe accident risk assessment on external event contributors to assess risks posed by operation of the plant. The applicant has developed perspectives using PRA models on the safety profile of the Limerick plant and has altered the plant design to reduce accident vulnerabilities identified in these PRAs. The staff's review of the Limerick PRA has particularly emphasized the dominant accident sequences and the resulting insights into demonstration of compliance with regulatory requirments, unique design features and major plant vulnerabilities to assess the need for any additional measures to further improve the safety of the LGS. The staff's review insights and PRA safety review conclusions are presented in this report

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

International Nuclear Information System (INIS)

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

1995-06-01

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

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

International Nuclear Information System (INIS)

1991-01-01

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

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

Probabilistic tsunami hazard assessment considering time-lag of seismic event on Nankai trough

International Nuclear Information System (INIS)

Sugino, Hideharu; Sakagami, Masaharu; Ebisawa, Katsumi; Korenaga, Mariko

2011-01-01

In the area in front of Nankai trough, tsunami wave height may increase if tsunamis attacking from some wave sources overlap because of time-lag of seismic event on Nankai trough. To evaluation tsunami risk of the important facilities located in front of Nankai trough, we proposed the probabilistic tsunami hazard assessment considering uncertainty on time-lag of seismic event on Nankai trough and we evaluated the influence that the time-lag gave to tsunami hazard at the some representative points. (author)

Overview of Future of Probabilistic Methods and RMSL Technology and the Probabilistic Methods Education Initiative for the US Army at the SAE G-11 Meeting

Science.gov (United States)

Singhal, Surendra N.

2003-01-01

The SAE G-11 RMSL Division and Probabilistic Methods Committee meeting sponsored by the Picatinny Arsenal during March 1-3, 2004 at Westin Morristown, will report progress on projects for probabilistic assessment of Army system and launch an initiative for probabilistic education. The meeting features several Army and industry Senior executives and Ivy League Professor to provide an industry/government/academia forum to review RMSL technology; reliability and probabilistic technology; reliability-based design methods; software reliability; and maintainability standards. With over 100 members including members with national/international standing, the mission of the G-11s Probabilistic Methods Committee is to enable/facilitate rapid deployment of probabilistic technology to enhance the competitiveness of our industries by better, faster, greener, smarter, affordable and reliable product development.

Flood Risk and Probabilistic Benefit Assessment to Support Management of Flood-Prone Lands: Evidence From Candaba Floodplains, Philippines

Science.gov (United States)

Juarez, A. M.; Kibler, K. M.; Sayama, T.; Ohara, M.

2016-12-01

Flood management decision-making is often supported by risk assessment, which may overlook the role of coping capacity and the potential benefits derived from direct use of flood-prone land. Alternatively, risk-benefit analysis can support floodplain management to yield maximum socio-ecological benefits for the minimum flood risk. We evaluate flood risk-probabilistic benefit tradeoffs of livelihood practices compatible with direct human use of flood-prone land (agriculture/wild fisheries) and nature conservation (wild fisheries only) in Candaba, Philippines. Located north-west to Metro Manila, Candaba area is a multi-functional landscape that provides a temporally-variable mix of possible land uses, benefits and ecosystem services of local and regional value. To characterize inundation from 1.3- to 100-year recurrence intervals we couple frequency analysis with rainfall-runoff-inundation modelling and remotely-sensed data. By combining simulated probabilistic floods with both damage and benefit functions (e.g. fish capture and rice yield with flood intensity) we estimate potential damages and benefits over varying probabilistic flood hazards. We find that although direct human uses of flood-prone land are associated with damages, for all the investigated magnitudes of flood events with different frequencies, the probabilistic benefits ( 91 million) exceed risks by a large margin ( 33 million). Even considering risk, probabilistic livelihood benefits of direct human uses far exceed benefits provided by scenarios that exclude direct "risky" human uses (difference of 85 million). In addition, we find that individual coping strategies, such as adapting crop planting periods to the flood pulse or fishing rather than cultivating rice in the wet season, minimize flood losses ( 6 million) while allowing for valuable livelihood benefits ($ 125 million) in flood-prone land. Analysis of societal benefits and local capacities to cope with regular floods demonstrate the

Probabilistic Remaining Useful Life Prediction of Composite Aircraft Components, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — A Probabilistic Fatigue Damage Assessment Network (PFDAN) toolkit for Abaqus will be developed for probabilistic life management of a laminated composite structure...

Applying probabilistic well-performance parameters to assessments of shale-gas resources

Science.gov (United States)

Charpentier, Ronald R.; Cook, Troy

2010-01-01

In assessing continuous oil and gas resources, such as shale gas, it is important to describe not only the ultimately producible volumes, but also the expected well performance. This description is critical to any cost analysis or production scheduling. A probabilistic approach facilitates (1) the inclusion of variability in well performance within a continuous accumulation, and (2) the use of data from developed accumulations as analogs for the assessment of undeveloped accumulations. In assessing continuous oil and gas resources of the United States, the U.S. Geological Survey analyzed production data from many shale-gas accumulations. Analyses of four of these accumulations (the Barnett, Woodford, Fayetteville, and Haynesville shales) are presented here as examples of the variability of well performance. For example, the distribution of initial monthly production rates for Barnett vertical wells shows a noticeable change with time, first increasing because of improved completion practices, then decreasing from a combination of decreased reservoir pressure (in infill wells) and drilling in less productive areas. Within a partially developed accumulation, historical production data from that accumulation can be used to estimate production characteristics of undrilled areas. An understanding of the probabilistic relations between variables, such as between initial production and decline rates, can improve estimates of ultimate production. Time trends or spatial trends in production data can be clarified by plots and maps. The data can also be divided into subsets depending on well-drilling or well-completion techniques, such as vertical in relation to horizontal wells. For hypothetical or lightly developed accumulations, one can either make comparisons to a specific well-developed accumulation or to the entire range of available developed accumulations. Comparison of the distributions of initial monthly production rates of the four shale-gas accumulations that were

Improvement of the drift chamber system in the SAPHIR detector and first measurements of the Φ meson production at threshold

International Nuclear Information System (INIS)

Scholmann, J.N.

1996-09-01

The SAPHIR detector at ELSA enables the measurement of photon induced Φ meson production from threshold up to 3 GeV in the full kinematical range. A considerable improvement of the drift chamber system is a precondition of gaining the necessary data rate in an acceptable time. The research focuses attention on the choice of the chamber gas and on a different mechanical construction, so as to minimize the negative influences of the photon beam crossing the sensitive volume of the drift chamber system. In addition, first preliminary results of the total and the differential cross section for the Φ meson production close to threshold were evaluated. (orig.)

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

Probabilistic risk assessment framework for structural systems under multiple hazards using Bayesian statistics

Energy Technology Data Exchange (ETDEWEB)

Kwag, Shinyoung [North Carolina State University, Raleigh, NC 27695 (United States); Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Gupta, Abhinav, E-mail: agupta1@ncsu.edu [North Carolina State University, Raleigh, NC 27695 (United States)

2017-04-15

Highlights: • This study presents the development of Bayesian framework for probabilistic risk assessment (PRA) of structural systems under multiple hazards. • The concepts of Bayesian network and Bayesian inference are combined by mapping the traditionally used fault trees into a Bayesian network. • The proposed mapping allows for consideration of dependencies as well as correlations between events. • Incorporation of Bayesian inference permits a novel way for exploration of a scenario that is likely to result in a system level “vulnerability.” - Abstract: Conventional probabilistic risk assessment (PRA) methodologies (USNRC, 1983; IAEA, 1992; EPRI, 1994; Ellingwood, 2001) conduct risk assessment for different external hazards by considering each hazard separately and independent of each other. The risk metric for a specific hazard is evaluated by a convolution of the fragility and the hazard curves. The fragility curve for basic event is obtained by using empirical, experimental, and/or numerical simulation data for a particular hazard. Treating each hazard as an independently can be inappropriate in some cases as certain hazards are statistically correlated or dependent. Examples of such correlated events include but are not limited to flooding induced fire, seismically induced internal or external flooding, or even seismically induced fire. In the current practice, system level risk and consequence sequences are typically calculated using logic trees to express the causative relationship between events. In this paper, we present the results from a study on multi-hazard risk assessment that is conducted using a Bayesian network (BN) with Bayesian inference. The framework can consider statistical dependencies among risks from multiple hazards, allows updating by considering the newly available data/information at any level, and provide a novel way to explore alternative failure scenarios that may exist due to vulnerabilities.

Probabilistic risk assessment framework for structural systems under multiple hazards using Bayesian statistics

International Nuclear Information System (INIS)

Kwag, Shinyoung; Gupta, Abhinav

2017-01-01

Highlights: • This study presents the development of Bayesian framework for probabilistic risk assessment (PRA) of structural systems under multiple hazards. • The concepts of Bayesian network and Bayesian inference are combined by mapping the traditionally used fault trees into a Bayesian network. • The proposed mapping allows for consideration of dependencies as well as correlations between events. • Incorporation of Bayesian inference permits a novel way for exploration of a scenario that is likely to result in a system level “vulnerability.” - Abstract: Conventional probabilistic risk assessment (PRA) methodologies (USNRC, 1983; IAEA, 1992; EPRI, 1994; Ellingwood, 2001) conduct risk assessment for different external hazards by considering each hazard separately and independent of each other. The risk metric for a specific hazard is evaluated by a convolution of the fragility and the hazard curves. The fragility curve for basic event is obtained by using empirical, experimental, and/or numerical simulation data for a particular hazard. Treating each hazard as an independently can be inappropriate in some cases as certain hazards are statistically correlated or dependent. Examples of such correlated events include but are not limited to flooding induced fire, seismically induced internal or external flooding, or even seismically induced fire. In the current practice, system level risk and consequence sequences are typically calculated using logic trees to express the causative relationship between events. In this paper, we present the results from a study on multi-hazard risk assessment that is conducted using a Bayesian network (BN) with Bayesian inference. The framework can consider statistical dependencies among risks from multiple hazards, allows updating by considering the newly available data/information at any level, and provide a novel way to explore alternative failure scenarios that may exist due to vulnerabilities.

Probabilistic assessments of fuel performance

International Nuclear Information System (INIS)

Kelppe, S.; Ranta-Puska, K.

1998-01-01

The probabilistic Monte Carlo Method, coupled with quasi-random sampling, is applied for the fuel performance analyses. By using known distributions of fabrication parameters and real power histories with their randomly selected combinations, and by making a large number of ENIGMA code calculations, one expects to find out the state of the whole reactor fuel. Good statistics requires thousands of runs. A sample case representing VVER-440 reactor fuel indicates relatively low fuel temperatures and mainly athermal fission gas release if any. The rod internal pressure remains typically below 2.5 MPa, which leaves a large margin to the system pressure of 12 MPa Gap conductance, an essential parameter in the accident evaluations, shows no decrease from its start-of-life value. (orig.)

Probabilistic seismic vulnerability and risk assessment of stone masonry structures

Science.gov (United States)

Abo El Ezz, Ahmad

Earthquakes represent major natural hazards that regularly impact the built environment in seismic prone areas worldwide and cause considerable social and economic losses. The high losses incurred following the past destructive earthquakes promoted the need for assessment of the seismic vulnerability and risk of the existing buildings. Many historic buildings in the old urban centers in Eastern Canada such as Old Quebec City are built of stone masonry and represent un-measurable architectural and cultural heritage. These buildings were built to resist gravity loads only and generally offer poor resistance to lateral seismic loads. Seismic vulnerability assessment of stone masonry buildings is therefore the first necessary step in developing seismic retrofitting and pre-disaster mitigation plans. The objective of this study is to develop a set of probability-based analytical tools for efficient seismic vulnerability and uncertainty analysis of stone masonry buildings. A simplified probabilistic analytical methodology for vulnerability modelling of stone masonry building with systematic treatment of uncertainties throughout the modelling process is developed in the first part of this study. Building capacity curves are developed using a simplified mechanical model. A displacement based procedure is used to develop damage state fragility functions in terms of spectral displacement response based on drift thresholds of stone masonry walls. A simplified probabilistic seismic demand analysis is proposed to capture the combined uncertainty in capacity and demand on fragility functions. In the second part, a robust analytical procedure for the development of seismic hazard compatible fragility and vulnerability functions is proposed. The results are given by sets of seismic hazard compatible vulnerability functions in terms of structure-independent intensity measure (e.g. spectral acceleration) that can be used for seismic risk analysis. The procedure is very efficient for

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)

ISSUES ASSOCIATED WITH PROBABILISTIC FAILURE MODELING OF DIGITAL SYSTEMS

International Nuclear Information System (INIS)

CHU, T.L.; MARTINEZ-GURIDI, G.; LIHNER, J.; OVERLAND, D.

2004-01-01

The current U.S. Nuclear Regulatory Commission (NRC) licensing process of instrumentation and control (I and C) systems is based on deterministic requirements, e.g., single failure criteria, and defense in depth and diversity. Probabilistic considerations can be used as supplements to the deterministic process. The National Research Council has recommended development of methods for estimating failure probabilities of digital systems, including commercial off-the-shelf (COTS) equipment, for use in probabilistic risk assessment (PRA). NRC staff has developed informal qualitative and quantitative requirements for PRA modeling of digital systems. Brookhaven National Laboratory (BNL) has performed a review of the-state-of-the-art of the methods and tools that can potentially be used to model digital systems. The objectives of this paper are to summarize the review, discuss the issues associated with probabilistic modeling of digital systems, and identify potential areas of research that would enhance the state of the art toward a satisfactory modeling method that could be integrated with a typical probabilistic risk assessment

Use of probabilistic risk assessment in fuel cycle facilities

International Nuclear Information System (INIS)

Gonzalez, Felix; Gonzalez, Michelle; Wagner, Brian

2013-01-01

As expressed in its Policy Statement on the Use of Probabilistic Risk Assessment (PRA) Methods in Nuclear Regulatory Activities, the U.S Nuclear Regulatory Commission has been working for decades to increase the use of PRA technology in its regulatory activities. Since the policy statement was issued in 1995, PRA has become a core component of the nuclear power plant (NPP) licensing and oversight processes. In the last several years, interest has increased in PRA technologies and their possible application to other areas including, but not limited to, spent fuel handling, fuel cycle facilities, reprocessing facilities, and advanced reactors. This paper describes the application of PRA technology currently used in NPPs and its application in other areas such as fuel cycle facilities and advanced reactors. It describes major challenges that are being faced in the application of PRA into new technical areas and possible ways to resolve them. (authors)

MATILDA Version-2: Rough Earth TIALD Model for Laser Probabilistic Risk Assessment in Hilly Terrain - Part II

Science.gov (United States)

2017-07-28

risk assessment for “unsafe” scenarios. Recently, attention in the DoD has turned to Probabilistic Risk Assessment (PRA) models [5,6] as an...corresponding to the CRA undershoot boundary. The magenta- coloured line represents the portion of the C-RX(U) circle that would contribute to the...Tertiary Precaution Surface. Undershoot related laser firing restrictions within the green- coloured C-RX(U) can be ignored. Figure 34

Bayesian inference in probabilistic risk assessment-The current state of the art

International Nuclear Information System (INIS)

Kelly, Dana L.; Smith, Curtis L.

2009-01-01

Markov chain Monte Carlo (MCMC) approaches to sampling directly from the joint posterior distribution of aleatory model parameters have led to tremendous advances in Bayesian inference capability in a wide variety of fields, including probabilistic risk analysis. The advent of freely available software coupled with inexpensive computing power has catalyzed this advance. This paper examines where the risk assessment community is with respect to implementing modern computational-based Bayesian approaches to inference. Through a series of examples in different topical areas, it introduces salient concepts and illustrates the practical application of Bayesian inference via MCMC sampling to a variety of important problems

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)

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

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.

Application of probabilistic safety assessment to research reactors

International Nuclear Information System (INIS)

1989-07-01

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

Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

International Nuclear Information System (INIS)

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

2013-01-01

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

Risk assessment methods in radiotherapy: Probabilistic safety assessment (PSA); Los metodos de analisis de riesgo en radioterapia: Analisis Probabilistico de seguridad (APS)

Energy Technology Data Exchange (ETDEWEB)

Ramirez Vera, M. L.; Perez Mulas, A.; Delgado, J. M.; Barrientos Ontero, M.; Somoano, F.; Alvarez Garcia, C.; Rodriguez Marti, M.

2011-07-01

The understanding of accidents that have occurred in radiotherapy and the lessons learned from them are very useful to prevent repetition, but there are other risks that have not been detected to date. With a view to identifying and preventing such risks, proactive methods successfully applied in other fields, such as probabilistic safety assessment (PSA), have been developed. (Author)

A probabilistic risk assessment of Oconee Unit 3. Executive highlights 60

International Nuclear Information System (INIS)

1984-04-01

In 1980 the Nuclear Safety Analysis Center and Duke Power Co. joined in a project to provide the utility industry with a practical, useful example of the application of probabilistic risk assessment (PRA) methods. PRA is a structured analysis technique that accounts for all the failure possibilities that might conceivably lead to core damage. The technique uses probabilities as discriminators to determine which are most significant. The following were project objectives: to provide the host utility with an analytic model of the plant that describes and estimates the likelihood of failure combinations that could lead to core melt; to evaluate the risks to the plant and to the public; to improve utility capabilities in PRA methods and applications

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)

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)

Review of models for use in probabilistic assessments of the disposal of high level radioactive wastes on or beneath the seabed

International Nuclear Information System (INIS)

Hooper, A.G.; Gibson, A.E.

1985-08-01

The report describes a review of computer models applicable to probabilistic assessments of the ocean disposal of vitrified high level waste. Computer models were identified and each evaluated against a common set of criteria. The review covers each stage of the chain of radionuclide transport: canister deterioration and leaching, migration through the bottom sediments, dispersion in the ocean, suspended sediment effects, biological aspects and uptake by man. The models identified are generally at a relatively early stage of development and documentation was often only of a general nature. These overall systems models and sub-models considered to be most suitable for probabilistic assessments were identified and areas of further development identified. (author)

Probabilistic risk assessment model for allergens in food: sensitivity analysis of the minimum eliciting dose and food consumption

NARCIS (Netherlands)

Kruizinga, A.G.; Briggs, D.; Crevel, R.W.R.; Knulst, A.C.; Bosch, L.M.C.v.d.; Houben, G.F.

2008-01-01

Previously, TNO developed a probabilistic model to predict the likelihood of an allergic reaction, resulting in a quantitative assessment of the risk associated with unintended exposure to food allergens. The likelihood is estimated by including in the model the proportion of the population who is

The definition and treatment of scenarios in a probabilistic systems assessment

Energy Technology Data Exchange (ETDEWEB)

Andres, T.H.; Goodwin, B.W

1998-03-01

Concepts for the disposal of nuclear waste generally involve the use of multiple barriers, both man-made and natural, to isolate the waste from man's environment for very long periods of time. To assess the safety of these concepts, postclosure safety assessments are conducted to study the types, ranges and probabilities of potential impacts caused by wastes penetrating these barriers. Within the Canadian context, postclosure safety assessments must estimate radiological risk for comparison with regulatory criteria. The definition of radiological risk introduces a requirement to identify and then evaluate scenarios that describe quantitatively the possible behaviour of a disposal system and its surroundings for at least 10{sup 4} years. This report examines the connection between the identification and evaluation of scenarios. It shows how a large collection of factors can be combined to form simple scenarios, and how a large number of simple scenarios can be grouped into a manageably small number of compound scenarios. The compound scenarios, notably the 'central' scenario, can be evaluated using a probabilistic systems assessment code such as SYVAC3 (SYstems Variability Analysis Code, generation 3) to estimate the radiological risk and other environmental impacts. Examples are taken from a long-term performance assessment study to demonstrate a pragmatic application of the procedure. (author)

The definition and treatment of scenarios in a probabilistic systems assessment

International Nuclear Information System (INIS)

Andres, T.H.; Goodwin, B.W.

1998-03-01

Concepts for the disposal of nuclear waste generally involve the use of multiple barriers, both man-made and natural, to isolate the waste from man's environment for very long periods of time. To assess the safety of these concepts, postclosure safety assessments are conducted to study the types, ranges and probabilities of potential impacts caused by wastes penetrating these barriers. Within the Canadian context, postclosure safety assessments must estimate radiological risk for comparison with regulatory criteria. The definition of radiological risk introduces a requirement to identify and then evaluate scenarios that describe quantitatively the possible behaviour of a disposal system and its surroundings for at least 10 4 years. This report examines the connection between the identification and evaluation of scenarios. It shows how a large collection of factors can be combined to form simple scenarios, and how a large number of simple scenarios can be grouped into a manageably small number of compound scenarios. The compound scenarios, notably the 'central' scenario, can be evaluated using a probabilistic systems assessment code such as SYVAC3 (SYstems Variability Analysis Code, generation 3) to estimate the radiological risk and other environmental impacts. Examples are taken from a long-term performance assessment study to demonstrate a pragmatic application of the procedure. (author)

Probabilistic insurance

OpenAIRE

Wakker, P.P.; Thaler, R.H.; Tversky, A.

1997-01-01

textabstractProbabilistic insurance is an insurance policy involving a small probability that the consumer will not be reimbursed. Survey data suggest that people dislike probabilistic insurance and demand more than a 20% reduction in the premium to compensate for a 1% default risk. While these preferences are intuitively appealing they are difficult to reconcile with expected utility theory. Under highly plausible assumptions about the utility function, willingness to pay for probabilistic i...

Integration of Probabilistic and Multi-Hazard Risk Assessment Within Urban Development Planning and Emergency Preparedness and Response:Application to Manizales, Colombia

Institute of Scientific and Technical Information of China (English)

Gabriel A.Bernal; Mario A.Salgado-Gálvez; Daniela Zuloaga; Julián Tristancho; Diana González; Omar-Dar(i)o Cardona

2017-01-01

The details of a multi-hazard and probabilistic risk assessment,developed for urban planning and emergency response activities in Manizales,Colombia,are presented in this article.This risk assessment effort was developed under the framework of an integral disaster risk management project whose goal was to connect risk reduction activities by using open access and state-of-theart risk models.A probabilistic approach was used for the analysis of seismic,landslide,and volcanic hazards to obtain stochastic event sets suitable for probabilistic loss estimation and to generate risk results in different metrics after aggregating in a rigorous way the losses associated to the different hazards.Detailed and high resolution exposure databases were used for the building stock and infrastructure of the city together with a set of vulnerability functions for each of the perils considered.The urban and territorial ordering plan of the city was updated for socioeconomic development and land use using the hazard and risk inputs and determinants,which cover not only the current urban area but also those adjacent areas where the expansion of Manizales is expected to occur.The emergency response capabilities of the city were improved by taking into account risk scenarios and after updating an automatic and real-time post-earthquake damage assessment.

Probabilistic risk assessment and its role in plant modifications

International Nuclear Information System (INIS)

Diederich, A.R.; McElroy, W.F.

1986-01-01

Electric Utilities today have a tool available to improve management's ability to evaluate nuclear power plant modifications (MODS). Probabilistic Risk Assessment (PRA), is a tool of choice since it can be applied to a specific situation such as MOD request review, bringing the perspectives of reliability, financial risk and consequences to the public in addition to the more rigid requirements like those associated with Quality Assurance or licensing criteria. The techniques used in the PRA process revolve about the creation and manipulation of Fault Trees and Event Trees, which are used to quantify the event sequences and reliability of plant systems in a logical framework. It is through these methods that chains of sequences, or events, are understood. The degree to which plant systems are modelled in the PRA can vary depending on resources and purpose. Philadelphia Elecrtric Company's PRA modelled ten (10) major systems but this number may increase during the application and updating process

A Control Variate Method for Probabilistic Performance Assessment. Improved Estimates for Mean Performance Quantities of Interest

Energy Technology Data Exchange (ETDEWEB)

MacKinnon, Robert J.; Kuhlman, Kristopher L

2016-05-01

We present a method of control variates for calculating improved estimates for mean performance quantities of interest, E(PQI) , computed from Monte Carlo probabilistic simulations. An example of a PQI is the concentration of a contaminant at a particular location in a problem domain computed from simulations of transport in porous media. To simplify the presentation, the method is described in the setting of a one- dimensional elliptical model problem involving a single uncertain parameter represented by a probability distribution. The approach can be easily implemented for more complex problems involving multiple uncertain parameters and in particular for application to probabilistic performance assessment of deep geologic nuclear waste repository systems. Numerical results indicate the method can produce estimates of E(PQI)having superior accuracy on coarser meshes and reduce the required number of simulations needed to achieve an acceptable estimate.

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)

Undiscovered porphyry copper resources in the Urals—A probabilistic mineral resource assessment

Science.gov (United States)

Hammarstrom, Jane M.; Mihalasky, Mark J.; Ludington, Stephen; Phillips, Jeffrey; Berger, Byron R.; Denning, Paul; Dicken, Connie; Mars, John; Zientek, Michael L.; Herrington, Richard J.; Seltmann, Reimar

2017-01-01

A probabilistic mineral resource assessment of metal resources in undiscovered porphyry copper deposits of the Ural Mountains in Russia and Kazakhstan was done using a quantitative form of mineral resource assessment. Permissive tracts were delineated on the basis of mapped and inferred subsurface distributions of igneous rocks assigned to tectonic zones that include magmatic arcs where the occurrence of porphyry copper deposits within 1 km of the Earth's surface are possible. These permissive tracts outline four north-south trending volcano-plutonic belts in major structural zones of the Urals. From west to east, these include permissive lithologies for porphyry copper deposits associated with Paleozoic subduction-related island-arc complexes preserved in the Tagil and Magnitogorsk arcs, Paleozoic island-arc fragments and associated tonalite-granodiorite intrusions in the East Uralian zone, and Carboniferous continental-margin arcs developed on the Kazakh craton in the Transuralian zone. The tracts range from about 50,000 to 130,000 km2 in area. The Urals host 8 known porphyry copper deposits with total identified resources of about 6.4 million metric tons of copper, at least 20 additional porphyry copper prospect areas, and numerous copper-bearing skarns and copper occurrences.Probabilistic estimates predict a mean of 22 undiscovered porphyry copper deposits within the four permissive tracts delineated in the Urals. Combining estimates with established grade and tonnage models predicts a mean of 82 million metric tons of undiscovered copper. Application of an economic filter suggests that about half of that amount could be economically recoverable based on assumed depth distributions, availability of infrastructure, recovery rates, current metals prices, and investment environment.

Probabilistic Damage Stability Calculations for Ships

DEFF Research Database (Denmark)

Jensen, Jørgen Juncher

1996-01-01

The aim of these notes is to provide background material for the present probabilistic damage stability rules fro dry cargo ships.The formulas for the damage statistics are derived and shortcomings as well as possible improvements are discussed. The advantage of the definiton of fictitious...... compartments in the formulation of a computer-based general procedure for probabilistic damaged stability assessment is shown. Some comments are given on the current state of knowledge on the ship survivability in damaged conditions. Finally, problems regarding proper account of water ingress through openings...

Potential for accidents in a nuclear power plant: probabilistic risk assessment, applied statistical decision theory, and implications of such considerations to mathematics education

International Nuclear Information System (INIS)

Dios, R.A.

1984-01-01

This dissertation focuses upon the field of probabilistic risk assessment and its development. It investigates the development of probabilistic risk assessment in nuclear engineering. To provide background for its development, the related areas of population dynamics (demography), epidemiology and actuarial science are studied by presenting information upon how risk has been viewed in these areas over the years. A second major problem involves presenting an overview of the mathematical models related to risk analysis to mathematics educators and making recommendations for presenting this theory in classes of probability and statistics for mathematics and engineering majors at the undergraduate and graduate levels

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)

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

International Nuclear Information System (INIS)

Lin, Yufei; Chen, Maoyin; Zhou, Donghua

2013-01-01

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

Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

Directory of Open Access Journals (Sweden)

Sunarko

2017-01-01

Full Text Available Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2 PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

Review process and quality assurance in the EBR-II probabilistic risk assessment

International Nuclear Information System (INIS)

Roglans, J.; Hill, D.J.; Ragland, W.A.

1992-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A reactor, has recently been completed at Argonne National Laboratory (ANL). Within the scope of the ANL QA Programs, a QA Plan specifically for the EBR-II PRA was developed. The QA Plan covered all aspects of the PRA development, with emphasis on the procedures for document and software control, and the internal and external review process. The effort spent in the quality assurance tasks for the EBR-II PRA has reciprocated by providing acceptance of the work and confidence in the quality of the results

A probabilistic assessment of climate change impacts on yield and nitrogen leaching from winter wheat in Denmark

DEFF Research Database (Denmark)

Børgesen, Christen Duus; Olesen, Jørgen E

2011-01-01

Climate change will impact agricultural production both directly and indirectly, but uncertainties related to likely impacts constrain current political decision making on adaptation. This analysis focuses on a methodology for applying probabilistic climate change projections to assess modelled...... the importance of including soil information for regional studies of climate change impacts on cropping systems....

Probabilistic Risk Assessment of Cancer from Exposure Inorganic Arsenic in Duplicate Food by Villagers in Ronphibun, Thailand

OpenAIRE

Piyawat Saipan

2010-01-01

Ronphibun district is a district in Nakorn Si Thammarat province, within southern Thailand. This district is the site of several former tin mines that were in operation 100 years ago. Arsenic contamination caused by past mining activities remains in the area. The specific purpose of this study was conducted to assess cancer risk in people living within Ronphibun district from exposure to inorganic arsenic via duplicate food using probabilistic risk assessment. A hundred and fifty duplicate fo...

Probabilistic Risk Assessment of Cancer from Exposure Inorganic Arsenic in Duplicate Food by Villagers in Ronphibun, Thailand

Directory of Open Access Journals (Sweden)

Piyawat Saipan

2010-07-01

Full Text Available Ronphibun district is a district in Nakorn Si Thammarat province, within southern Thailand. This district is the site of several former tin mines that were in operation 100 years ago. Arsenic contamination caused by past mining activities remains in the area. The specific purpose of this study was conducted to assess cancer risk in people living within Ronphibun district from exposure to inorganic arsenic via duplicate food using probabilistic risk assessment. A hundred and fifty duplicate food samples were collected from participants. Inorganic arsenic concentrations are determined by hydride generation atomic absorption spectrometry. Inorganic arsenic concentrations in duplicate food ranged from 0.16 to 0.42 μg/g dry weight. The probabilistic carcinogenic risk levels were 6.76 x 10-4 and 1.74 x 10-3 based on the 50th and 95th percentile, respectively. Risk values for people in Ronphibun from exposure to inorganic arsenic remained higher than the acceptable target risk. Sensitivity analysis indicted that exposure duration and concentrations of arsenic in food were the two most influential of cancer risk estimates.

Review of the Oconee-3 probabilistic risk assessment: external events, core damage frequency. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Hanan, N.A.; Ilberg, D.; Xue, D.; Youngblood, R.; Reed, J.W.; McCann, M.; Talwani, T.; Wreathall, J.; Kurth, P.D.; Bandyopadhyay, K.

1986-03-01

A review of the Oconee-3 Probabilistic Risk Assessment (OPRA) was conducted with the broad objective of evaluating qualitatively and quantitatively (as much as possible) the OPRA assessment of the important sequences that are ''externally'' generated and lead to core damage. The review included a technical assessment of the assumptions and methods used in the OPRA within its stated objective and with the limited information available. Within this scope, BNL performed a detailed reevaluation of the accident sequences generated by internal floods and earthquakes and a less detailed review (in some cases a scoping review) for the accident sequences generated by fires, tornadoes, external floods, and aircraft impact. 12 refs., 24 figs., 31 tabs.

Review of the Oconee-3 probabilistic risk assessment: external events, core damage frequency. Volume 2

International Nuclear Information System (INIS)

Hanan, N.A.; Ilberg, D.; Xue, D.

1986-03-01

A review of the Oconee-3 Probabilistic Risk Assessment (OPRA) was conducted with the broad objective of evaluating qualitatively and quantitatively (as much as possible) the OPRA assessment of the important sequences that are ''externally'' generated and lead to core damage. The review included a technical assessment of the assumptions and methods used in the OPRA within its stated objective and with the limited information available. Within this scope, BNL performed a detailed reevaluation of the accident sequences generated by internal floods and earthquakes and a less detailed review (in some cases a scoping review) for the accident sequences generated by fires, tornadoes, external floods, and aircraft impact. 12 refs., 24 figs., 31 tabs

Safety evaluation for the LMFBR plant using probabilistic risk assessment techniques

International Nuclear Information System (INIS)

Kani, Y.; Aizawa, K.

1987-01-01

This paper presents an application of probabilistic risk assessment techniques to a typical loop-type liquid metal fast breeder reactor (LMFBR) plant in the detailed design stage. A comprehensive systems analysis has been performed to identify event sequences leading to core damage and provide insights into the importance of accident contributors. While traditional event tree/fault tree modeling was used for the analysis, this study involved a thorough investigation of initiating events and of support system faults. The qualification of accident sequences has been conducted by combining the fault trees based on the event trees and obtaining sequence cut sets with the use of the SETS code. This study also attempted to quantify the potential for operator recovery actions in the course of each accident sequence. (author)

Probabilistic risk assessment (PRA): status report and guidance for regulatory application. Draft report for comment

International Nuclear Information System (INIS)

1984-02-01

This document describes the current status of the methodologies used in probabilistic risk assessment (PRA) and provides guidance for the application of the results of PRAs to the nuclear reactor regulatory process. The PRA studies that have been completed or are underway are reviewed. The levels of maturity of the methodologies used in a PRA are discussed. Insights derived from PRAs are listed. The potential uses of PRA results for regulatory purposes are discussed

Spatial interactions database development for effective probabilistic risk assessment

International Nuclear Information System (INIS)

Liming, J. K.; Dunn, R. F.

2008-01-01

In preparation for a subsequent probabilistic risk assessment (PRA) fire risk analysis update, the STP Nuclear Operating Company (STPNOC) is updating its spatial interactions database (SID). This work is being performed to support updating the spatial interactions analysis (SIA) initially performed for the original South Texas Project Electric Generating Station (STPEGS) probabilistic safely assessment (PSA) and updated in the STPEGS Level 2 PSA and IPE Report. S/A is a large-scope screening analysis performed for nuclear power plant PRA that serves as a prerequisite basis for more detailed location-dependent, hazard-spec analyses in the PRA, such as fire risk analysis, flooding risk analysis, etc. SIA is required to support the 'completeness' argument for the PRA scope. The objectives of the current SID development effort are to update the spatial interactions analysis data, to the greatest degree practical, to be consistent with the following: the as-built plant as of December 31, 2007 the in-effect STPNOC STPEGS Units 1 and 2 PRA the current technology and intent of NUREG/CR-6850 guidance for lire risk analysis database support the requirements for PRA SIA, including fire and flooding risk analysis, established by NRC Regulatory Guide 1.200 and the ASME PRA Standard (ASME RA-S-2002 updated through ASME RA-Sc-2007,) This paper presents the approach and methodology for state-of-the-art SID development and applications, including an overview of the SIA process for nuclear power plant PRA. The paper shows how current relational database technology and existing, conventional station information sources can be employed to collect, process, and analyze spatial interactions data for the plant in an effective and efficient manner to meet the often challenging requirements of industry guidelines and standards such as NUREG/CR-6850, NRC Regulatory Guide 1.200, and ASME RA-S-2002 (updated through ASME RA-Sc 2007). This paper includes tables and figures illustrating how SIA

Application of Bayesian network to the probabilistic risk assessment of nuclear waste disposal

International Nuclear Information System (INIS)

Lee, Chang-Ju; Lee, Kun Jai

2006-01-01

The scenario in a risk analysis can be defined as the propagating feature of specific initiating event which can go to a wide range of undesirable consequences. If we take various scenarios into consideration, the risk analysis becomes more complex than do without them. A lot of risk analyses have been performed to actually estimate a risk profile under both uncertain future states of hazard sources and undesirable scenarios. Unfortunately, in case of considering specific systems such as a radioactive waste disposal facility, since the behaviour of future scenarios is hardly predicted without special reasoning process, we cannot estimate their risk only with a traditional risk analysis methodology. Moreover, we believe that the sources of uncertainty at future states can be reduced pertinently by setting up dependency relationships interrelating geological, hydrological, and ecological aspects of the site with all the scenarios. It is then required current methodology of uncertainty analysis of the waste disposal facility be revisited under this belief. In order to consider the effects predicting from an evolution of environmental conditions of waste disposal facilities, this paper proposes a quantitative assessment framework integrating the inference process of Bayesian network to the traditional probabilistic risk analysis. We developed and verified an approximate probabilistic inference program for the specific Bayesian network using a bounded-variance likelihood weighting algorithm. Ultimately, specific models, including a model for uncertainty propagation of relevant parameters were developed with a comparison of variable-specific effects due to the occurrence of diverse altered evolution scenarios (AESs). After providing supporting information to get a variety of quantitative expectations about the dependency relationship between domain variables and AESs, we could connect the results of probabilistic inference from the Bayesian network with the consequence

Probabilistic safety assessment for Balakovo 1000 MW NPP

International Nuclear Information System (INIS)

Foden, R.W.

1995-01-01

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

Global assessment of predictability of water availability: A bivariate probabilistic Budyko analysis

Science.gov (United States)

Wang, Weiguang; Fu, Jianyu

2018-02-01

Estimating continental water availability is of great importance for water resources management, in terms of maintaining ecosystem integrity and sustaining society development. To more accurately quantify the predictability of water availability, on the basis of univariate probabilistic Budyko framework, a bivariate probabilistic Budyko approach was developed using copula-based joint distribution model for considering the dependence between parameter ω of Wang-Tang's equation and the Normalized Difference Vegetation Index (NDVI), and was applied globally. The results indicate the predictive performance in global water availability is conditional on the climatic condition. In comparison with simple univariate distribution, the bivariate one produces the lower interquartile range under the same global dataset, especially in the regions with higher NDVI values, highlighting the importance of developing the joint distribution by taking into account the dependence structure of parameter ω and NDVI, which can provide more accurate probabilistic evaluation of water availability.

Toward uniform probabilistic seismic hazard assessments for Southeast Asia

Science.gov (United States)

Chan, C. H.; Wang, Y.; Shi, X.; Ornthammarath, T.; Warnitchai, P.; Kosuwan, S.; Thant, M.; Nguyen, P. H.; Nguyen, L. M.; Solidum, R., Jr.; Irsyam, M.; Hidayati, S.; Sieh, K.

2017-12-01

Although most Southeast Asian countries have seismic hazard maps, various methodologies and quality result in appreciable mismatches at national boundaries. We aim to conduct a uniform assessment across the region by through standardized earthquake and fault databases, ground-shaking scenarios, and regional hazard maps. Our earthquake database contains earthquake parameters obtained from global and national seismic networks, harmonized by removal of duplicate events and the use of moment magnitude. Our active-fault database includes fault parameters from previous studies and from the databases implemented for national seismic hazard maps. Another crucial input for seismic hazard assessment is proper evaluation of ground-shaking attenuation. Since few ground-motion prediction equations (GMPEs) have used local observations from this region, we evaluated attenuation by comparison of instrumental observations and felt intensities for recent earthquakes with predicted ground shaking from published GMPEs. We then utilize the best-fitting GMPEs and site conditions into our seismic hazard assessments. Based on the database and proper GMPEs, we have constructed regional probabilistic seismic hazard maps. The assessment shows highest seismic hazard levels near those faults with high slip rates, including the Sagaing Fault in central Myanmar, the Sumatran Fault in Sumatra, the Palu-Koro, Matano and Lawanopo Faults in Sulawesi, and the Philippine Fault across several islands of the Philippines. In addition, our assessment demonstrates the important fact that regions with low earthquake probability may well have a higher aggregate probability of future earthquakes, since they encompass much larger areas than the areas of high probability. The significant irony then is that in areas of low to moderate probability, where building codes are usually to provide less seismic resilience, seismic risk is likely to be greater. Infrastructural damage in East Malaysia during the 2015

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.

Efficient probabilistic inference in generic neural networks trained with non-probabilistic feedback.

Science.gov (United States)

Orhan, A Emin; Ma, Wei Ji

2017-07-26

Animals perform near-optimal probabilistic inference in a wide range of psychophysical tasks. Probabilistic inference requires trial-to-trial representation of the uncertainties associated with task variables and subsequent use of this representation. Previous work has implemented such computations using neural networks with hand-crafted and task-dependent operations. We show that generic neural networks trained with a simple error-based learning rule perform near-optimal probabilistic inference in nine common psychophysical tasks. In a probabilistic categorization task, error-based learning in a generic network simultaneously explains a monkey's learning curve and the evolution of qualitative aspects of its choice behavior. In all tasks, the number of neurons required for a given level of performance grows sublinearly with the input population size, a substantial improvement on previous implementations of probabilistic inference. The trained networks develop a novel sparsity-based probabilistic population code. Our results suggest that probabilistic inference emerges naturally in generic neural networks trained with error-based learning rules.Behavioural tasks often require probability distributions to be inferred about task specific variables. Here, the authors demonstrate that generic neural networks can be trained using a simple error-based learning rule to perform such probabilistic computations efficiently without any need for task specific operations.

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.

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

Probabilistic Design of Wind Turbines

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Toft, H.S.

2010-01-01

Probabilistic design of wind turbines requires definition of the structural elements to be included in the probabilistic basis: e.g., blades, tower, foundation; identification of important failure modes; careful stochastic modeling of the uncertain parameters; recommendations for target reliability....... It is described how uncertainties in wind turbine design related to computational models, statistical data from test specimens, results from a few full-scale tests and from prototype wind turbines can be accounted for using the Maximum Likelihood Method and a Bayesian approach. Assessment of the optimal...... reliability level by cost-benefit optimization is illustrated by an offshore wind turbine example. Uncertainty modeling is illustrated by an example where physical, statistical and model uncertainties are estimated....

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

Measurement of the reaction {gamma}d {yields}pn{pi}{sup +}{pi}{sup -} at SAPHIR and investigation of the decay angular distribution of the {Delta}{sup ++}(1232) resonance; Messung der Reaktion {gamma}d {yields}pn{pi}{sup +}{pi}{sup -} an SAPHIR und Untersuchung der Zerfallswinkelverteilung der {Delta}{sup ++}(1232)-Resonanz

Energy Technology Data Exchange (ETDEWEB)

Schuetz, P.

1993-03-01

SAPHIR, a new experiment at the Bonn electron stretcher ring ELSA, started taking data in spring 1992. It was set up for the investigation of photon induced reactions with multiparticle final states. In the first part of this paper the special design of the target is described. It can be operated with liquefied hydrogen or deuterium and is placed in the middle of the central drift chamber. To project the surrounding chamber in case of a fracture of the target cell as safety system is installed. In addition two independent methods of monitoring the cell are procided. The first measurement was performed with a deuterium target at a photon energy range of E{sub {gamma}} = 500-700 MeV. In the second part of this paper first results of an analysis of the decay angular distribution of the {Delta}{sup ++}(1232) in the reaction {gamma}d {yields} n{Delta}{sup ++}{pi}{sup -} are presented. They are compared to old data from a hydrogen bubble chamber experiment and are discussed on the basis of a spectator model. (orig.) [Deutsch] Im Rahmen dieser Arbeit ist der Aufbau eines Fluessiggas-Targets beschrieben worden, das speziell fuer den Einsatz im SAPHIR-Detektor entwickelt worden ist. Es wurden Funktionen zur Ueberwachung der Targetzelle vorgestellt und ein Sicherheitssystem zum Schutz der zentralen Driftkammer, die das Target unmittelbar umgibt. Weiterhin ist in Simulationsrechnungen untersucht worden, welchen Einfluss die Konstruktion des Targetstreutopfes auf die Messung unterschiedlicher Reaktionen haben kann. Dabei sind bei 50% bis 70% der Ereignisse Treffer in den Aluminiumstuetzen des Targetstreutopfes aufgetreten. Diese starke Beeintraechtigung kann durch eine Neukonstruktion des Streutopfes und der Verwendung von z.B. Rohazell als Streutopffenster vermieden werden. Rohazell zeichnet sich durch eine hohe Festigkeit und grosse Strahlungslaenge aus. An der Neukonstruktion des Streutopfes wird z.Z. gearbeitet. Der zweite Teil der Arbeit beschreibt eine der ersten

Probabilistic Modeling and Visualization for Bankruptcy Prediction

DEFF Research Database (Denmark)

Antunes, Francisco; Ribeiro, Bernardete; Pereira, Francisco Camara

2017-01-01

In accounting and finance domains, bankruptcy prediction is of great utility for all of the economic stakeholders. The challenge of accurate assessment of business failure prediction, specially under scenarios of financial crisis, is known to be complicated. Although there have been many successful...... studies on bankruptcy detection, seldom probabilistic approaches were carried out. In this paper we assume a probabilistic point-of-view by applying Gaussian Processes (GP) in the context of bankruptcy prediction, comparing it against the Support Vector Machines (SVM) and the Logistic Regression (LR......). Using real-world bankruptcy data, an in-depth analysis is conducted showing that, in addition to a probabilistic interpretation, the GP can effectively improve the bankruptcy prediction performance with high accuracy when compared to the other approaches. We additionally generate a complete graphical...

An assessment of dietary exposure to glyphosate using refined deterministic and probabilistic methods.

Science.gov (United States)

Stephenson, C L; Harris, C A

2016-09-01

Glyphosate is a herbicide used to control broad-leaved weeds. Some uses of glyphosate in crop production can lead to residues of the active substance and related metabolites in food. This paper uses data on residue levels, processing information and consumption patterns, to assess theoretical lifetime dietary exposure to glyphosate. Initial estimates were made assuming exposure to the highest permitted residue levels in foods. These intakes were then refined using median residue levels from trials, processing information, and monitoring data to achieve a more realistic estimate of exposure. Estimates were made using deterministic and probabilistic methods. Exposures were compared to the acceptable daily intake (ADI)-the amount of a substance that can be consumed daily without an appreciable health risk. Refined deterministic intakes for all consumers were at or below 2.1% of the ADI. Variations were due to cultural differences in consumption patterns and the level of aggregation of the dietary information in calculation models, which allows refinements for processing. Probabilistic exposure estimates ranged from 0.03% to 0.90% of the ADI, depending on whether optimistic or pessimistic assumptions were made in the calculations. Additional refinements would be possible if further data on processing and from residues monitoring programmes were available. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

The application of probabilistic risk assessment to a LLW incinerator

International Nuclear Information System (INIS)

Li, K.K.; Huang, F.T.

1993-01-01

The 100 Kg/hr low-level radioactive waste (LLW) incinerator and the 1,500 ton supercompactor are two main vehicles in the Taiwan Power Company's Volume Reduction Center. Since the hot test of the incinerator in mid 1990, various problems associated with the original design and operating procedures were encountered. During the early stages of putting an incinerator in service, the modification and fine-tuning of the system would help future reliable operations. The probabilistic risk assessment (PRA) method was introduced to evaluate the interaction between potential system failure and its environmental impact and further help diagnose the system defects initially. The draft Level 1 system analysis was completed and the event and fault trees were constructed. Qualitatively, this approach is useful for preventing the system failure from occurring. However, Levels 2 and 3 analysis can only be done when sufficient data become available in the future

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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.

Does a more sophisticated storm erosion model improve probabilistic erosion estimates?

NARCIS (Netherlands)

Ranasinghe, R.W.M.R.J.B.; Callaghan, D.; Roelvink, D.

2013-01-01

The dependency between the accuracy/uncertainty of storm erosion exceedance estimates obtained via a probabilistic model and the level of sophistication of the structural function (storm erosion model) embedded in the probabilistic model is assessed via the application of Callaghan et al.'s (2008)

Probabilistic assessment of exposure to hair cosmetic products by the French population.

Science.gov (United States)

Ficheux, A S; Bernard, A; Chevillotte, G; Dornic, N; Roudot, A C

2016-06-01

Cosmetic exposure data are limited in Europe and especially in France. The aim of this study was to assess the exposure to hair cosmetics using recent consumption data (percentage of users, frequency of use and amount per use) generated for the French population (Ficheux et al., 2015, 2016). Exposure was assessed using a probabilistic method for eleven hair products: liquid shampoo, dry shampoo, conditioner, hair mask, hair serum, hair oil, styling lacquer, styling gel, styling foam, styling wax and styling spray. Exposure was assessed by sex and by age classes in adults and children. Pregnant women were also studied. For liquid shampoo, conditioner and some styling products (gel, lacquer and foam), the levels of exposure were higher than the values currently used by the Scientific Committee on Consumer Safety (SCCS). Exposure values found for styling wax and styling spray were lower than SCCS values. Exposure was assessed for the first time for dry shampoo, hair mask, hair serum and hair oil products. These new French exposure values will be useful for safety assessors and for safety agencies in order to protect the general population and these at-risk populations. Copyright © 2016 Elsevier Ltd. All rights reserved.

Probabilistic safety analysis and interpretation thereof

International Nuclear Information System (INIS)

Steininger, U.; Sacher, H.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

On Probabilistic Alpha-Fuzzy Fixed Points and Related Convergence Results in Probabilistic Metric and Menger Spaces under Some Pompeiu-Hausdorff-Like Probabilistic Contractive Conditions

OpenAIRE

De la Sen, M.

2015-01-01

In the framework of complete probabilistic metric spaces and, in particular, in probabilistic Menger spaces, this paper investigates some relevant properties of convergence of sequences to probabilistic α-fuzzy fixed points under some types of probabilistic contractive conditions.

Use of probabilistic methods for analysis of cost and duration uncertainties in a decision analysis framework

International Nuclear Information System (INIS)

Boak, D.M.; Painton, L.

1995-01-01

Probabilistic forecasting techniques have been used in many risk assessment and performance assessment applications on radioactive waste disposal projects such as Yucca Mountain and the Waste Isolation Pilot Plant (WIPP). Probabilistic techniques such as Monte Carlo and Latin Hypercube sampling methods are routinely used to treat uncertainties in physical parameters important in simulating radionuclide transport in a coupled geohydrologic system and assessing the ability of that system to comply with regulatory release limits. However, the use of probabilistic techniques in the treatment of uncertainties in the cost and duration of programmatic alternatives on risk and performance assessment projects is less common. Where significant uncertainties exist and where programmatic decisions must be made despite existing uncertainties, probabilistic techniques may yield important insights into decision options, especially when used in a decision analysis framework and when properly balanced with deterministic analyses. For relatively simple evaluations, these types of probabilistic evaluations can be made using personal computer-based software