WorldWideScience

Sample records for accident sequence events

  1. Accident sequence precursor events with age-related contributors

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, G.A.; Kohn, W.E.

    1995-12-31

    The Accident Sequence Precursor (ASP) Program at ORNL analyzed about 14.000 Licensee Event Reports (LERs) filed by US nuclear power plants 1987--1993. There were 193 events identified as precursors to potential severe core accident sequences. These are reported in G/CR-4674. Volumes 7 through 20. Under the NRC Nuclear Plant Aging Research program, the authors evaluated these events to determine the extent to which component aging played a role. Events were selected that involved age-related equipment degradation that initiated an event or contributed to an event sequence. For the 7-year period, ORNL identified 36 events that involved aging degradation as a contributor to an ASP event. Except for 1992, the percentage of age-related events within the total number of ASP events over the 7-year period ({approximately}19%) appears fairly consistent up to 1991. No correlation between plant ape and number of precursor events was found. A summary list of the age-related events is presented in the report.

  2. Prediction of accident sequence probabilities in a nuclear power plant due to earthquake events

    International Nuclear Information System (INIS)

    Hudson, J.M.; Collins, J.D.

    1980-01-01

    This paper presents a methodology to predict accident probabilities in nuclear power plants subject to earthquakes. The resulting computer program accesses response data to compute component failure probabilities using fragility functions. Using logical failure definitions for systems, and the calculated component failure probabilities, initiating event and safety system failure probabilities are synthesized. The incorporation of accident sequence expressions allows the calculation of terminal event probabilities. Accident sequences, with their occurrence probabilities, are finally coupled to a specific release category. A unique aspect of the methodology is an analytical procedure for calculating top event probabilities based on the correlated failure of primary events

  3. Event sequence quantification for a loss of shutdown cooling accident in the GCFR

    International Nuclear Information System (INIS)

    Frank, M.; Reilly, J.

    1979-10-01

    A summary is presented of the core-wide sequence of events of a postulated total loss of forced and natural convection decay heat removal in a shutdown Gas-Cooled Fast Reactor (GCFR). It outlines the analytical methods and results for the progression of the accident sequence. This hypothetical accident proceeds in the distinct phases of cladding melting, assembly wall melting and molten steel relocation into the interassembly spacing, and fuel relocation. It identifies the key phenomena of the event sequence and the concerns and mechanisms of both recriticality and recriticality prevention

  4. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

    International Nuclear Information System (INIS)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

    1994-01-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  5. Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events

    Directory of Open Access Journals (Sweden)

    Robertas Alzbutas

    2015-01-01

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

  6. Accident sequence quantification with KIRAP

    International Nuclear Information System (INIS)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

    1997-01-01

    The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

  7. Accident sequence quantification with KIRAP

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

    1997-01-01

    The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

  8. Current understanding of the sequence of events. Overview of current understanding of accident progression at Fukushima Dai-ichi

    International Nuclear Information System (INIS)

    Gulliford, Jim

    2013-01-01

    An overview of the main sequence of events, particularly the evolution of the cores in Units 1-3 was given. The presentation is based on information provided by Dr Okajima of JAEA to the June 2012 Nuclear Science Committee meeting. During the accident, conditions at the plant were such that operators were initially unable to obtain instruments readouts from the control panel and hence could not know what condition the reactors were in. (Reactor Power, Pressure, Temperature, Water height and flow rate, etc.). Subsequently, as electrical power supplies were gradually restored more data became available. In addition to the reactor data, other information from off-site measurements and from measuring stations inside the site boundary is now available, particularly for radiation dose rates in air. These types of information, combined with detailed knowledge of the plant design and operations history up to the time of the accident are being used to construct detailed computer models which simulate the behaviour of the reactor core, pressure vessel and containment during the accident sequence. This combination of detailed design/operating data, limited measured data during the accident and computer modelling allows us to construct a fairly clear picture of the accident progression. The main sequence of events (common to Units 1, 2 and 3) is summarised. The OECD/NEA is currently coordinating an international benchmark study of the accident at Fukushima Daiichi known as the BSAF Project. The objectives of this activity are to analyse and evaluate the accident progression and improve severe accident (SA) analysis methods and models. The project provides valuable additional (and corrected) data from plant measurements as well as an improved understanding of the role played by the fuel and cladding design. Based on (limited) plant data and extensive modelling analysis, we have a detailed qualitative description of the Fukushima-Daiichi accident. Further analyses of the type

  9. Efficient method for simulation of BWR severe accident sequence events before core uncovery

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1984-01-01

    BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are modeled. Valuable insights have been gained into accident sequences. A Fortran version is under development and it will be modified for application to Mark II plants

  10. Treatment of Events Representing System Success in Accident Sequences in PSA Models with ET/FT Linking

    International Nuclear Information System (INIS)

    Vrbanic, I.; Spiler, J.; Mikulicic, V.; Simic, Z.

    2002-01-01

    Treatment of events that represent systems' successes in accident sequences is well known issue associated primarily with those PSA models that employ event tree / fault tree (ET / FT) linking technique. Even theoretically clear, practical implementation and usage creates for certain PSA models a number of difficulties regarding result correctness. Strict treatment of success-events would require consistent applying of de Morgan laws. However, there are several problems related to it. First, Boolean resolution of the overall model, such as the one representing occurrence of reactor core damage, becomes very challenging task if De Morgan rules are applied consistently at all levels. Even PSA tools of the newest generation have some problems with performing such a task in a reasonable time frame. The second potential issue is related to the presence of negated basic events in minimal cutsets. If all the basic events that result from strict applying of De Morgan rules are retained in presentation of minimal cutsets, their readability and interpretability may be impaired severely. It is also worth noting that the concept of a minimal cutset is tied to equipment failures, rather than to successes. For reasons like these, various simplifications are employed in PSA models and tools, when it comes to the treatment of success-events in the sequences. This paper provides a discussion of major concerns associated with the treatment of success-events in accident sequences of a typical PSA model. (author)

  11. Development on quantitative safety analysis method of accident scenario. The automatic scenario generator development for event sequence construction of accident

    International Nuclear Information System (INIS)

    Kojima, Shigeo; Onoue, Akira; Kawai, Katsunori

    1998-01-01

    This study intends to develop a more sophisticated tool that will advance the current event tree method used in all PSA, and to focus on non-catastrophic events, specifically a non-core melt sequence scenario not included in an ordinary PSA. In the non-catastrophic event PSA, it is necessary to consider various end states and failure combinations for the purpose of multiple scenario construction. Therefore it is anticipated that an analysis work should be reduced and automated method and tool is required. A scenario generator that can automatically handle scenario construction logic and generate the enormous size of sequences logically identified by state-of-the-art methodology was developed. To fulfill the scenario generation as a technical tool, a simulation model associated with AI technique and graphical interface, was introduced. The AI simulation model in this study was verified for the feasibility of its capability to evaluate actual systems. In this feasibility study, a spurious SI signal was selected to test the model's applicability. As a result, the basic capability of the scenario generator could be demonstrated and important scenarios were generated. The human interface with a system and its operation, as well as time dependent factors and their quantification in scenario modeling, was added utilizing human scenario generator concept. Then the feasibility of an improved scenario generator was tested for actual use. Automatic scenario generation with a certain level of credibility, was achieved by this study. (author)

  12. Domino effect in chemical accidents: main features and accident sequences.

    Science.gov (United States)

    Darbra, R M; Palacios, Adriana; Casal, Joaquim

    2010-11-15

    The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes are external events (31%) and mechanical failure (29%). Storage areas (35%) and process plants (28%) are by far the most common settings for domino accidents. Eighty-nine per cent of the accidents involved flammable materials, the most frequent of which was LPG. The domino effect sequences were analyzed using relative probability event trees. The most frequent sequences were explosion→fire (27.6%), fire→explosion (27.5%) and fire→fire (17.8%). Copyright © 2010 Elsevier B.V. All rights reserved.

  13. CNE (central nuclear en Embalse): probabilistic safety study. Loss-of-coolant accidents. Analysis through events sequence

    International Nuclear Information System (INIS)

    Layral, S.I.

    1987-01-01

    The aim of this study was to perform for the Embalse nuclear power plant, a probabilistic evaluation of loss-of-coolant accidents (LOCA) to identify the risks associated with them and to determine their acceptability in accordance with norms. This study includes all ruptures in the primary system that produce the automatic activation of 'emergency core cooling system'. Three starting events were selected for the probabilistic evaluation: 100% rupture of an input collector; 5% rupture of an input collector; 1.2% rupture of an input collector. At this stage the evaluation is focussed on the identification and quantization of the main failure sequences that follow a LOCA and lead to an uncontrolled reactor state or 'core meltdown'. The most important contribution to the core meltdown due to LOCA is the failure of supplies that are required for the emergency core cooling system. (Author)

  14. 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

  15. Domino effect in chemical accidents: main features and accident sequences

    OpenAIRE

    Casal Fàbrega, Joaquim; Darbra Roman, Rosa Maria

    2010-01-01

    The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes a...

  16. Quantitative risk trends deriving from PSA-based event analyses. Analysis of results from U.S.NRC's accident sequence precursor program

    International Nuclear Information System (INIS)

    Watanabe, Norio

    2004-01-01

    The United States Nuclear Regulatory Commission (U.S.NRC) has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the probabilistic safety assessment (PSA) technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U.S. nuclear power plants. Although the results from the ASP Program include valuable information that could be useful for obtaining and characterizing risk significant insights and for monitoring risk trends in nuclear power industry, there are only a few attempts to determine and develop the trends using the ASP results. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the results of the ASP analysis. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. As well, the present study demonstrates that two risk indicators used here can provide quantitative information useful for examining and monitoring the risk trends and/or risk characteristics in nuclear power industry. (author)

  17. Severe accident sequences simulated at the Grand Gulf Nuclear Station

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1999-01-01

    Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents

  18. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Cuantificacion de secuencias de accidente y resultados

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A; Aguilar T, O; Nunez C, A; Lopez M, R [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1994-07-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the {sup I}nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  19. Accident sequences simulated at the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1998-01-01

    Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida

  20. Risk assessment for long-term post-accident sequences

    International Nuclear Information System (INIS)

    Ellia-Hervy, A.; Ducamp, F.

    1987-11-01

    Probabilistic risk analysis, currently conducted by the CEA (French Atomic Energy Commission) for the French replicate series of 900 MWe power plants, has identified accident sequences requiring long-term operation of some systems after the initiating event. They have been named long-term sequences. Quantification of probabilities of such sequences cannot rely exclusively on equipment failure-on-demand data: it must also take into account operating failures, the probability of which increase with time. Specific studies have therefore been conducted for a number of plant systems actuated during these long-term sequences. This has required: - Definition of the most realistic equipment utilization strategies based on existing emergency procedures for 900 MWe French plants. - Evaluation of the potential to repair failed equipment, given accessibility, repair time, and specific radiation conditions for the given sequence. - Definition of the event bringing the long-term sequence to an end. - Establishment of an appropriate quantification method, capable of taking into account the evolution of assumptions concerning equipment utilization strategies or repair conditions over time. The accident sequence quantification method based on realistic scenarios has been used in the risk assessment of the initiating event loss of reactor coolant accident occurring at power and at shutdown. Compared with the results obtained from conventional methods, this method redistributes the relative weight of accident sequences and also demonstrates that the long term can be a significant contribution to the probability of core melt

  1. Accident sequence analysis of human-computer interface design

    International Nuclear Information System (INIS)

    Fan, C.-F.; Chen, W.-H.

    2000-01-01

    It is important to predict potential accident sequences of human-computer interaction in a safety-critical computing system so that vulnerable points can be disclosed and removed. We address this issue by proposing a Multi-Context human-computer interaction Model along with its analysis techniques, an Augmented Fault Tree Analysis, and a Concurrent Event Tree Analysis. The proposed augmented fault tree can identify the potential weak points in software design that may induce unintended software functions or erroneous human procedures. The concurrent event tree can enumerate possible accident sequences due to these weak points

  2. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe accident research results to enhance the capability of plant operating personnel to effectively manage severe accidents. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed as part of the research program designed to resolve this issue. The methodology identifies the information needs of the plant personnel during a wide range of accident conditions, the existing plant measurements capable of supplying these information needs and what, if any minor additions to instrument and display systems would enhance the capability to manage accidents, known limitations on the capability of these measurements to function properly under the conditions that will be present during a wide range of severe accidents, and areas in which the information systems could mislead plant personnel. This paper presents an application of this methodology to a severe accident sequence to demonstrate its use in identifying the information which is available for management of the event. The methodology has been applied to a severe accident sequence in a Pressurized Water Reactor with a large dry containment. An examination of the capability of the existing measurements was then performed to determine whether the information needs can be supplied

  3. Progress in methodology for probabilistic assessment of accidents: timing of accident sequences

    International Nuclear Information System (INIS)

    Lanore, J.M.; Villeroux, C.; Bouscatie, F.; Maigret, N.

    1981-09-01

    There is an important problem for probabilistic studies of accident sequences using the current event tree techniques. Indeed this method does not take into account the dependence in time of the real accident scenarios, involving the random behaviour of the systems (lack or delay in intervention, partial failures, repair, operator actions ...) and the correlated evolution of the physical parameters. A powerful method to perform the probabilistic treatment of these complex sequences (dynamic evolution of systems and associated physics) is Monte-Carlo simulation, very rare events being treated with the help of suitable weighting and biasing techniques. As a practical example the accident sequences related to the loss of the residual heat removal system in a fast breeder reactor has been treated with that method

  4. Probabilistic accident sequence recovery analysis

    International Nuclear Information System (INIS)

    Stutzke, Martin A.; Cooper, Susan E.

    2004-01-01

    Recovery analysis is a method that considers alternative strategies for preventing accidents in nuclear power plants during probabilistic risk assessment (PRA). Consideration of possible recovery actions in PRAs has been controversial, and there seems to be a widely held belief among PRA practitioners, utility staff, plant operators, and regulators that the results of recovery analysis should be skeptically viewed. This paper provides a framework for discussing recovery strategies, thus lending credibility to the process and enhancing regulatory acceptance of PRA results and conclusions. (author)

  5. Stressful life events and occupational accidents.

    Science.gov (United States)

    Cordeiro, Ricardo; Dias, Adriano

    2005-10-01

    The purpose of this study was to examine the association between stressful life events and occupational accidents. This was a population-based case-control study, carried out in the city of Botucatu, in southeast Brazil. The cases consisted of 108 workers who had recently experienced occupational accidents. Each case was matched with three controls. The cases and controls answered a questionnaire about recent exposure to stressful life events. Reporting of "environmental problems", "being a victim of assault", "not having enough food at home" and "nonoccupational fatigue" were found to be risk factors for work-related accidents with estimated incidence rate ratios of 1.4 [95% confidence interval (95% CI) 1.1-1.7], 1.3 (95% CI 1.1-1.7), 1.3 (95% CI 1.1-1.6), and 1.4 (95% CI 1.2-1.7) respectively. The findings of the study suggested that nonwork variables contribute to occupational accidents, thus broadening the understanding of these phenomena, which can support new approaches to the prevention of occupational accidents.

  6. PSA modeling of long-term accident sequences

    International Nuclear Information System (INIS)

    Georgescu, Gabriel; Corenwinder, Francois; Lanore, Jeanne-Marie

    2014-01-01

    In the context of the extension of PSA scope to include external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the accident sequences induced by initiators which affect a whole site containing several nuclear units (reactors, fuel pools,...). These methodological improvements represent an essential prerequisite for the development of external hazards PSA. However, it has to be noted that in French PSA, even before Fukushima, long term accident sequences were taken into account: many insight were therefore used, as complementary information, to enhance the safety level of the plants. IRSN proposed an external events PSA development program. One of the first steps of the program is the development of methods to model in the PSA the long term accident sequences, based on the experience gained. At short term IRSN intends to enhance the modeling of the 'long term' accident sequences induced by the loss of the heat sink or/and the loss of external power supply. The experience gained by IRSN and EDF from the development of several probabilistic studies treating long term accident sequences shows that the simple extension of the mission time of the mitigation systems from 24 hours to longer times is not sufficient to realistically quantify the risk and to obtain a correct ranking of the risk contributions and that treatment of recoveries is also necessary. IRSN intends to develop a generic study which can be used as a general methodology for the assessment of the long term accident sequences, mainly generated by external hazards and their combinations. This first attempt to develop this generic study allowed identifying some aspects, which may be hazard (or combinations of hazards) or related to initial boundary conditions, which should be taken into account for further developments. (authors)

  7. Accident Sequence Precursor Analysis for SGTR by Using Dynamic PSA Approach

    International Nuclear Information System (INIS)

    Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

    2016-01-01

    In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. Based on the results of a sampling analysis and data from the conventional PSA model, the CDF caused by Sequence no. 26 can be realistically estimated. For a TLOFW accident with LOCA, second accident timings were categorized according to plant condition. Indicators were selected as branch point using the flow chart and tables, and a corresponding sequence tree model was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model

  8. The Accident Sequence Precursor program: Methods improvements and current results

    International Nuclear Information System (INIS)

    Minarick, J.W.; Manning, F.M.; Harris, J.D.

    1987-01-01

    Changes in the US NRC Accident Sequence Precursor program methods since the initial program evaluations of 1969-81 operational events are described, along with insights from the review of 1984-85 events. For 1984-85, the number of significant precursors was consistent with the number observed in 1980-81, dominant sequences associated with significant events were reasonably consistent with PRA estimates for BWRs, but lacked the contribution due to small-break LOCAs previously observed and predicted in PWRs, and the frequency of initiating events and non-recoverable system failures exhibited some reduction compared to 1980-81. Operational events which provide information concerning additional PRA modeling needs are also described

  9. Categorization of PWR accident sequences and guidelines for fault trees: seismic initiators

    International Nuclear Information System (INIS)

    Kimura, C.Y.

    1984-09-01

    This study developed a set of dominant accident sequences that could be applied generically to domestic commercial PWRs as a standardized basis for a probabilistic seismic risk assessment. This was accomplished by ranking the Zion 1 accident sequences. The pertinent PWR safety systems were compared on a plant-by-plant basis to determine the applicability of the dominant accident sequences of Zion 1 to other PWR plants. The functional event trees were developed to describe the system functions that must work or not work in order for a certain accident sequence to happen, one for pipe breaks and one for transients

  10. Study of event sequence database for a nuclear power domain

    International Nuclear Information System (INIS)

    Kusumi, Yoshiaki

    1998-01-01

    A retrieval engine developed to extract event sequences from an accident information database using a time series retrieval formula expressed with ordered retrieval terms is explored. This engine outputs not only a sequence which completely matches with a time series retrieval formula, but also sequence which approximately matches the formula (fuzzy retrieval). An event sequence database in which records consist of three ordered parameters, namely the causal event, the process and result. Then the database is used to assess the feasibility of this engine and favorable results were obtained. (author)

  11. Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

    International Nuclear Information System (INIS)

    Kim, Bo Gyung; Kang Hyun Gook; Yoon, Ho Joon

    2016-01-01

    In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results

  12. Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bo Gyung; Kang Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Abu Dhabi (United Arab Emirates)

    2016-05-15

    In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results.

  13. Identification and evaluation of accident sequences in nuclear power reactors

    International Nuclear Information System (INIS)

    Amendola, A.; Capobianchi, S.; Mancini, G.; Olivi, L.; Volta, G.; Reina, G.

    1981-01-01

    Probabilistic analysis techniques are being more and more used for the evaluation of accident progression in nuclear power plants, especially after the issue of the Reactor Safety Study (Report WASH-1400). This study and subsequent discussions have indicated the necessity of better investigating some major items, namely: adequate data base for the probabilistic evaluations; completeness of the analysis with respect both to accident initiation and behaviour; adequate treatment of uncertainties on the physical and operational parameters governing the accident behaviour. Furthermore, recent occurrences have stressed the importance of the operational aspects of reactor safety, such as plant-specific identification of possible occurrences, their prompt recognition, on-line prediction of subsequent developments and actions to be taken. The paper reviews the contributions in progress at JRC-Ispra to all these aspects, and specifically reports on the following: (1) The set-up of a European Reliability Data System for the acquisition and organisation of operational data of LWRs in the European Community. (2) The development of more complete and realistic models of systems. This work includes multistate static models of components and systems with a view to automatic fault-tree construction and dynamic models for accident sequence identification. The dynamic modelling approach ESCS (Event Sequence and Consequences Spectrum), shown in detail with an example, represents a step forward with respect to event-tree technique and opens new possibilities in dealing with human factors and on-line diagnosis problems. (3) The development of RSM (Response Surface Methodology) for the analysis of uncertainty propagations in consequence and in probability of accident chains. (author)

  14. Development of an accident sequence precursor methodology and its application to significant accident precursors

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Seung Hyun; Park, Sung Hyun; Jae, Moo Sung [Dept. of of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-03-15

    The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

  15. Accident sequence precursor analysis level 2/3 model development

    International Nuclear Information System (INIS)

    Lui, C.H.; Galyean, W.J.; Brownson, D.A.

    1997-01-01

    The US Nuclear Regulatory Commission's Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models

  16. Fukushima. The accident sequence and important causes. Pt. 1/3

    International Nuclear Information System (INIS)

    Pistner, Christoph

    2013-01-01

    On March 11, 2011 a strong earthquake at the east coast of Japan and a subsequent tsunami caused severe damage at the NPP site of Fukushima Daiichi. The article covers the fundamental safety aspects of the accident progress according to the state of knowledge. The principles of nuclear technology and reactor safety are summarized in order to allow the understanding of the accidental sequence. Even two years after the disaster many questions on the sequence of accident events are still open.

  17. Review of the severe accident risk reduction program (SARRP) containment event trees

    International Nuclear Information System (INIS)

    1986-05-01

    A part of the Severe Accident Risk Reduction Program, researchers at Sandia National Laboratories have constructed a group of containment event trees to be used in the analysis of key accident sequences for light water reactors (LWR) during postulated severe accidents. The ultimate goal of the program is to provide to the NRC staff a current assessment of the risk from severe reactor accidents for a group of five light water reactors. This review specifically focuses on the development and construction of the containment event trees and the results for containment failure probability, modes and timing. The report first gives the background on the program, the review criteria, and a summary of the observations, findings and recommendations. secondly, the individual reviews of each committee member on the event trees is presented. Finally, a review is provided on the computer model used to construct and evaluate the event trees

  18. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    International Nuclear Information System (INIS)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul

    2015-01-01

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree

  19. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

  20. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  1. BWR severe accident sequence analyses at ORNL - some lessons learned

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1983-01-01

    Boiling water reactor severe accident sequence studies are being carried out using Browns Ferry Unit 1 as the model plant. Four accident studies were completed, resulting in recommendations for improvements in system design, emergency procedures, and operator training. Computer code improvements were an important by-product

  2. Event course analysis of core disruptive accidents

    International Nuclear Information System (INIS)

    Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

    1995-01-01

    The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

  3. A Quantitative Accident Sequence Analysis for a VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jintae; Lee, Joeun; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    In Korea, the basic design features of VHTR are currently discussed in the various design concepts. Probabilistic risk assessment (PRA) offers a logical and structured method to assess risks of a large and complex engineered system, such as a nuclear power plant. It will be introduced at an early stage in the design, and will be upgraded at various design and licensing stages as the design matures and the design details are defined. Risk insights to be developed from the PRA are viewed as essential to developing a design that is optimized in meeting safety objectives and in interpreting the applicability of the existing demands to the safety design approach of the VHTR. In this study, initiating events which may occur in VHTRs were selected through MLD method. The initiating events were then grouped into four categories for the accident sequence analysis. Initiating events frequency and safety systems failure rate were calculated by using reliability data obtained from the available sources and fault tree analysis. After quantification, uncertainty analysis was conducted. The SR and LR frequency are calculated respectively 7.52E- 10/RY and 7.91E-16/RY, which are relatively less than the core damage frequency of LWRs.

  4. CATEGORIZATION OF EVENT SEQUENCES FOR LICENSE APPLICATION

    Energy Technology Data Exchange (ETDEWEB)

    G.E. Ragan; P. Mecheret; D. Dexheimer

    2005-04-14

    The purposes of this analysis are: (1) Categorize (as Category 1, Category 2, or Beyond Category 2) internal event sequences that may occur before permanent closure of the repository at Yucca Mountain. (2) Categorize external event sequences that may occur before permanent closure of the repository at Yucca Mountain. This includes examining DBGM-1 seismic classifications and upgrading to DBGM-2, if appropriate, to ensure Beyond Category 2 categorization. (3) State the design and operational requirements that are invoked to make the categorization assignments valid. (4) Indicate the amount of material put at risk by Category 1 and Category 2 event sequences. (5) Estimate frequencies of Category 1 event sequences at the maximum capacity and receipt rate of the repository. (6) Distinguish occurrences associated with normal operations from event sequences. It is beyond the scope of the analysis to propose design requirements that may be required to control radiological exposure associated with normal operations. (7) Provide a convenient compilation of the results of the analysis in tabular form. The results of this analysis are used as inputs to the consequence analyses in an iterative design process that is depicted in Figure 1. Categorization of event sequences for permanent retrieval of waste from the repository is beyond the scope of this analysis. Cleanup activities that take place after an event sequence and other responses to abnormal events are also beyond the scope of the analysis.

  5. CATEGORIZATION OF EVENT SEQUENCES FOR LICENSE APPLICATION

    International Nuclear Information System (INIS)

    G.E. Ragan; P. Mecheret; D. Dexheimer

    2005-01-01

    The purposes of this analysis are: (1) Categorize (as Category 1, Category 2, or Beyond Category 2) internal event sequences that may occur before permanent closure of the repository at Yucca Mountain. (2) Categorize external event sequences that may occur before permanent closure of the repository at Yucca Mountain. This includes examining DBGM-1 seismic classifications and upgrading to DBGM-2, if appropriate, to ensure Beyond Category 2 categorization. (3) State the design and operational requirements that are invoked to make the categorization assignments valid. (4) Indicate the amount of material put at risk by Category 1 and Category 2 event sequences. (5) Estimate frequencies of Category 1 event sequences at the maximum capacity and receipt rate of the repository. (6) Distinguish occurrences associated with normal operations from event sequences. It is beyond the scope of the analysis to propose design requirements that may be required to control radiological exposure associated with normal operations. (7) Provide a convenient compilation of the results of the analysis in tabular form. The results of this analysis are used as inputs to the consequence analyses in an iterative design process that is depicted in Figure 1. Categorization of event sequences for permanent retrieval of waste from the repository is beyond the scope of this analysis. Cleanup activities that take place after an event sequence and other responses to abnormal events are also beyond the scope of the analysis

  6. ACCIDENTS AND UNSCHEDULED EVENTS ASSOCIATED WITH NON-NUCLEAR ENERGY RESOURCES AND TECHNOLOGY

    Science.gov (United States)

    Accidents and unscheduled events associated with non-nuclear energy resources and technology are identified for each step in the energy cycle. Both natural and anthropogenic causes of accidents or unscheduled events are considered. Data concerning these accidents are summarized. ...

  7. Accident sequences and causes analysis in a hydrogen production process

    Energy Technology Data Exchange (ETDEWEB)

    Jae, Moo Sung; Hwang, Seok Won; Kang, Kyong Min; Ryu, Jung Hyun; Kim, Min Soo; Cho, Nam Chul; Jeon, Ho Jun; Jung, Gun Hyo; Han, Kyu Min; Lee, Seng Woo [Hanyang Univ., Seoul (Korea, Republic of)

    2006-03-15

    Since hydrogen production facility using IS process requires high temperature of nuclear power plant, safety assessment should be performed to guarantee the safety of facility. First of all, accident cases of hydrogen production and utilization has been surveyed. Based on the results, risk factors which can be derived from hydrogen production facility were identified. Besides the correlation between risk factors are schematized using influence diagram. Also initiating events of hydrogen production facility were identified and accident scenario development and quantification were performed. PSA methodology was used for identification of initiating event and master logic diagram was used for selection method of initiating event. Event tree analysis was used for quantification of accident scenario. The sum of all the leakage frequencies is 1.22x10{sup -4} which is similar value (1.0x10{sup -4}) for core damage frequency that International Nuclear Safety Advisory Group of IAEA suggested as a criteria.

  8. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core

  9. Sequencing Events: Exploring Art and Art Jobs.

    Science.gov (United States)

    Stephens, Pamela Geiger; Shaddix, Robin K.

    2000-01-01

    Presents an activity for upper-elementary students that correlates the actions of archaeologists, patrons, and artists with the sequencing of events in a logical order. Features ancient Egyptian art images. Discusses the preparation of materials, motivation, a pre-writing activity, and writing a story in sequence. (CMK)

  10. Service Processes as a Sequence of Events

    NARCIS (Netherlands)

    P.C. Verhoef (Peter); G. Antonides (Gerrit); A.N. de Hoog

    2002-01-01

    textabstractIn this paper the service process is considered as a sequence of events. Using theory from economics and psychology a model is formulated that explains how the utility of each event affects the overall evaluation of the service process. In this model we especially account for the

  11. Automated Testing with Targeted Event Sequence Generation

    DEFF Research Database (Denmark)

    Jensen, Casper Svenning; Prasad, Mukul R.; Møller, Anders

    2013-01-01

    Automated software testing aims to detect errors by producing test inputs that cover as much of the application source code as possible. Applications for mobile devices are typically event-driven, which raises the challenge of automatically producing event sequences that result in high coverage...

  12. An analysis of LOCA sequences in the development of severe accident analysis DB

    International Nuclear Information System (INIS)

    Choi, Young; Park, Soo Yong; Ahn, Kwang-Il; Kim, D.H.

    2006-01-01

    Although a Level 2 PSA was performed for the Korean Standard Power Plants (KSNPs), and it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results. At present, KAERI is calculating the severe accident sequences intensively for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviours. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by knowledge-base technique, and the expected plant behaviour. The plant model used in this paper is oriented to the case of LOCAs related severe accident phenomena and thus can simulate the plant behaviours for a severe accident. Therefore the developed system may play a central role as an information source for decision-making for a severe accident management, and will be used as a training simulator for a severe accident management. (author)

  13. Accident precursors, near misses, and warning signs: Critical review and formal definitions within the framework of Discrete Event Systems

    International Nuclear Information System (INIS)

    Saleh, Joseph H.; Saltmarsh, Elizabeth A.; Favarò, Francesca M.; Brevault, Loïc

    2013-01-01

    An important consideration in safety analysis and accident prevention is the identification of and response to accident precursors. These off-nominal events are opportunities to recognize potential accident pathogens, identify overlooked accident sequences, and make technical and organizational decisions to address them before further escalation can occur. When handled properly, the identification of precursors provides an opportunity to interrupt an accident sequence from unfolding; when ignored or missed, precursors may only provide tragic proof after the fact that an accident was preventable. In this work, we first provide a critical review of the concept of precursor, and we highlight important features that ought to be distinguished whenever accident precursors are discussed. We address for example the notion of ex-ante and ex-post precursors, identified for postulated and instantiated (occurred) accident sequences respectively, and we discuss the feature of transferability of precursors. We then develop a formal (mathematical) definition of accident precursors as truncated accident sequences within the modeling framework of Discrete Event Systems. Additionally, we examine the related notions of “accident pathogens” as static or lurking adverse conditions that can contribute to or aggravate an accident, as well as “near misses”, “warning signs” and the novel concept of “accident pathway”. While these terms are within the same linguistic neighborhood as “accident precursors”, we argue that there are subtle but important differences between them and recommend that they not be used interchangeably for the sake of accuracy and clarity of communication within the risk and safety community. We also propose venues for developing quantitative importance measures for accident precursors, similar to component importance measures in reliability engineering. Our objective is to establish a common understanding and clear delineation of these terms, and

  14. Fukushima. The accident sequence and important causes. Pt. 1/3; Fukushima. Unfallablauf und wesentliche Ursachen. T. 1/3

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany). Bereich Nukleartechnik und Anlagensicherheit

    2013-07-01

    On March 11, 2011 a strong earthquake at the east coast of Japan and a subsequent tsunami caused severe damage at the NPP site of Fukushima Daiichi. The article covers the fundamental safety aspects of the accident progress according to the state of knowledge. The principles of nuclear technology and reactor safety are summarized in order to allow the understanding of the accidental sequence. Even two years after the disaster many questions on the sequence of accident events are still open.

  15. Human factors review for Severe Accident Sequence Analysis (SASA)

    International Nuclear Information System (INIS)

    Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

    1984-01-01

    The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure

  16. Core damage frequency estimation using accident sequence precursor data: 1990-1993

    International Nuclear Information System (INIS)

    Martz, H.F.

    1998-01-01

    The Nuclear Regulatory Commission's (NRC's) ongoing Accident Sequence Precursor (ASP) program uses probabilistic risk assessment (PRA) techniques to assess the potential for severe core damage (henceforth referred to simply as core damage) based on operating events. The types of operating events considered include accident sequence initiators, safety equipment failures, and degradation of plant conditions that could increase the probability that various postulated accident sequences occur. Such operating events potentially reduce the margin of safety available for prevention of core damage an thus can be considered as precursors to core damage. The current process for identifying, analyzing, and documenting ASP events is described in detail in Vanden Heuval et al. The significance of a Licensee Event Report (LER) event (or events) is measured by means of the conditional probability that the event leads to core damage, the so-called conditional core damage probability or, simply, CCDP. When the first ASP study results were published in 1982, it covered the period 1969--1979. In addition to identification and ranking of precursors, the original study attempted to estimate core damage frequency (CDF) based on the precursor events. The purpose of this paper is to compare the average annual CDF estimates calculated using the CCDP sum, Cooke-Goossens, Bier, and Abramson estimators for various reactor classes using the combined ASP data for the four years, 1990--1993. An important outcome of this comparison is an answer to the persistent question regarding the degree and effect of the positive bias of the CCDP sum method in practice. Note that this paper only compares the estimators with each other. Because the true average CDF is unknown, the estimation error is also unknown. Therefore, any observations or characterizations of bias are based on purely theoretical considerations

  17. Accident Sequence Evaluation Program: Human reliability analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  18. Accident Sequence Evaluation Program: Human reliability analysis procedure

    International Nuclear Information System (INIS)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs

  19. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

  20. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-01-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  1. Accident analyses in nuclear power plants following external initiating events and in the shutdown state. Final report

    International Nuclear Information System (INIS)

    Loeffler, Horst; Kowalik, Michael; Mildenberger, Oliver; Hage, Michael

    2016-06-01

    The work which is documented here provides the methodological basis for improvement of the state of knowledge for accident sequences after plant external initiating events and for accident sequences which begin in the shutdown state. The analyses have been done for a PWR and for a BWR reference plant. The work has been supported by the German federal ministry BMUB under the label 3612R01361. Top objectives of the work are: - Identify relevant event sequences in order to define characteristic initial and boundary conditions - Perform accident analysis of selected sequences - Evaluate the relevance of accident sequences in a qualitative way The accident analysis is performed with the code MELCOR 1.8.6. The applied input data set has been significantly improved compared to previous analyses. The event tree method which is established in PSA level 2 has been applied for creating a structure for a unified summarization and evaluation of the results from the accident analyses. The computer code EVNTRE has been applied for this purpose. In contrast to a PSA level 2, the branching probabilities of the event tree have not been determined with the usual accuracy, but they are given in an approximate way only. For the PWR, the analyses show a considerable protective effect of the containment also in the case of beyond design events. For the BWR, there is a rather high probability for containment failure under core melt impact, but nevertheless the release of radionuclides into the environment is very limited because of plant internal retention mechanisms. This report concludes with remarks about existing knowledge gaps and with regard to core melt sequences, and about possible improvements of the plant safety.

  2. Indemnification of damage in the event of a nuclear accident

    International Nuclear Information System (INIS)

    2003-01-01

    The Workshop on the Indemnification of Damage in the Event of a Nuclear Accident, organised by the OECD Nuclear Energy Agency in close co-operation with the French authorities, was held in Paris from 26 to 28 November 2001. This event was an integral part of the International Nuclear Emergency Exercise INEX 2000. It attracted wide participation from national nuclear authorities, regulators, operators of nuclear installations, nuclear insurers and international organisations. The objective was to test the capacity of the existing nuclear liability and compensation mechanisms in the 29 countries represented at the workshop to manage the consequences of a nuclear emergency. This workshop was based upon the scenario used for the INEX 2000 Exercise, i.e. an accident simulated at the Gravelines nuclear power plant in the north of France in May 2001. These proceedings contain a comparative analysis of legislative and regulatory provisions governing emergency response and nuclear third party liability, based upon country replies to a questionnaire. This publication also includes the full responses provided to that questionnaire, as well as the texts of presentations made by special guests from Germany and Japan describing the manner in which the public authorities in their respective countries responded to two nuclear accidents of a very different nature and scale. (authors)

  3. Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

    International Nuclear Information System (INIS)

    Park, S.Y.; Ahn, K.I.

    2014-01-01

    Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

  4. System risk evolution analysis and risk critical event identification based on event sequence diagram

    International Nuclear Information System (INIS)

    Luo, Pengcheng; Hu, Yang

    2013-01-01

    During system operation, the environmental, operational and usage conditions are time-varying, which causes the fluctuations of the system state variables (SSVs). These fluctuations change the accidents’ probabilities and then result in the system risk evolution (SRE). This inherent relation makes it feasible to realize risk control by monitoring the SSVs in real time, herein, the quantitative analysis of SRE is essential. Besides, some events in the process of SRE are critical to system risk, because they act like the “demarcative points” of safety and accident, and this characteristic makes each of them a key point of risk control. Therefore, analysis of SRE and identification of risk critical events (RCEs) are remarkably meaningful to ensure the system to operate safely. In this context, an event sequence diagram (ESD) based method of SRE analysis and the related Monte Carlo solution are presented; RCE and risk sensitive variable (RSV) are defined, and the corresponding identification methods are also proposed. Finally, the proposed approaches are exemplified with an accident scenario of an aircraft getting into the icing region

  5. Indemnification of Damage in the Event of a Nuclear Accident

    International Nuclear Information System (INIS)

    2006-01-01

    The Second International Workshop on the Indemnification of Nuclear Damage was held in Bratislava, Slovak Republic, from 18 to 20 May 2005. The workshop was co-organised by the OECD Nuclear Energy Agency and the Nuclear Regulatory Authority of the Slovak Republic. It attracted wide participation from national nuclear authorities, regulators, operators of nuclear installations, nuclear insurers and international organisations. The purpose of the workshop was to assess the third party liability and compensation mechanisms that would be implemented by participating countries in the event of a nuclear accident taking place within or near their borders. To accommodate this objective, two fictitious accident scenarios were developed: one involving a fire in a nuclear installation located in the Slovak Republic and resulting in the release of significant amounts of radioactive materials off-site, and the other a fire on board a ship transporting enriched uranium hexafluoride along the Danube River. The first scenario was designed to involve the greatest possible number of countries, with the second being restricted to countries with a geographical proximity to the Danube. These proceedings contain the papers presented at the workshop, as well as reports on the discussion sessions held. (author)

  6. Modeling framework for crew decisions during accident sequences

    International Nuclear Information System (INIS)

    Lukic, Y.D.; Worledge, D.H.; Hannaman, G.W.; Spurgin, A.J.

    1986-01-01

    The ability to model the average behavior of operating crews in the course of accident sequences is vital in learning on how to prevent damage to power plants and to maintain safety. This paper summarizes the work carried out in support of a Human Reliability Model framework. This work develops the mathematical framework of the model and identifies the parameters which could be measured in some way, e.g., through simulator experience and/or small scale tests. Selected illustrative examples are presented, of the numerical experiments carried out in order to understand the model sensitivity to parameter variation. These examples ar discussed with the objective of deriving insights of general nature regarding operating of the model which may lead to enhanced understanding of man/machine interactions

  7. A proposal for accident management optimization based on the study of accident sequence analysis for a BWR

    International Nuclear Information System (INIS)

    Sobajima, M.

    1998-01-01

    The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

  8. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  9. Methodology for time-dependent reliability analysis of accident sequences and complex reactor systems

    International Nuclear Information System (INIS)

    Paula, H.M.

    1984-01-01

    The work presented here is of direct use in probabilistic risk assessment (PRA) and is of value to utilities as well as the Nuclear Regulatory Commission (NRC). Specifically, this report presents a methodology and a computer program to calculate the expected number of occurrences for each accident sequence in an event tree. The methodology evaluates the time-dependent (instantaneous) and the average behavior of the accident sequence. The methodology accounts for standby safety system and component failures that occur (a) before they are demanded, (b) upon demand, and (c) during the mission (system operation). With respect to failures that occur during the mission, this methodology is unique in the sense that it models components that can be repaired during the mission. The expected number of system failures during the mission provides an upper bound for the probability of a system failure to run - the mission unreliability. The basic event modeling includes components that are continuously monitored, periodically tested, and those that are not tested or are otherwise nonrepairable. The computer program ASA allows practical applications of the method developed. This work represents a required extension of the presently available methodology and allows a more realistic PRA of nuclear power plants

  10. Cooperation in the Event of Nuclear Accidents; Cooperation en Matiere d'Accidents Nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Bresson, G. [CEA, Centre d' etudes nucleaires, de Fontenay-aux-Roses (France)

    1969-10-15

    This paper is concerned only with the action to be taken in respect of an individual directly affected by an accident and not with the more general measures relating to the population as a whole. Keeping the same sequence of ideas, the paper deals with nuclear establishments and cites criteria for classifying them; hence only the relationship between the establishment and the hospital, and between the radiation protection experts and medical personnel, is discussed. The complex organization of emergency measures, reception of the victim of the accident, and the treatment possibly required should be based on standard practice and published material, both national and international, allowance being made for the characteristics of each sector. A ''flexible'' plan of co-ordination is given as an illustration. Action must be taken in such cases at the site of the accident, inside and outside the establishment, and above all at the hospital. All categories of persons are involved in the process, i.e. fellow-workers, management, specialized services, and medical personnel, each with their own part to play. The manpower and equipment brought into service therefore vary, and depend upon the internal and external relations maintained by the establishment. The measures envisaged should provide for the transport, reception and treatment of those involved in the accident. An existing organization of this kind is described as an illustration. Finally, no action can be of value without full knowledge of the facts and thorough training of the personnel. Some clearly defined ideas on the.subject are considered under this heading. (author) [French] Le memoire ne traite que de la conduite a tenir envers un accidente et non du probleme, plus general, des mesures relatives a une population. Dans le meme ordre d'idees, l'etude porte sur les etablissements nucleaires et leurs criteres de classement; il ne s'agit donc que des liaisons entre retablissement et l'hopital et entre les

  11. A methodology for the quantitative risk assessment of major accidents triggered by seismic events

    International Nuclear Information System (INIS)

    Antonioni, Giacomo; Spadoni, Gigliola; Cozzani, Valerio

    2007-01-01

    A procedure for the quantitative risk assessment of accidents triggered by seismic events in industrial facilities was developed. The starting point of the procedure was the use of available historical data to assess the expected frequencies and the severity of seismic events. Available equipment-dependant failure probability models (vulnerability or fragility curves) were used to assess the damage probability of equipment items due to a seismic event. An analytic procedure was subsequently developed to identify, evaluate the credibility and finally assess the expected consequences of all the possible scenarios that may follow the seismic events. The procedure was implemented in a GIS-based software tool in order to manage the high number of event sequences that are likely to be generated in large industrial facilities. The developed methodology requires a limited amount of additional data with respect to those used in a conventional QRA, and yields with a limited effort a preliminary quantitative assessment of the contribution of the scenarios triggered by earthquakes to the individual and societal risk indexes. The application of the methodology to several case-studies evidenced that the scenarios initiated by seismic events may have a relevant influence on industrial risk, both raising the overall expected frequency of single scenarios and causing specific severe scenarios simultaneously involving several plant units

  12. Analysis on the nitrogen drilling accident of Well Qionglai 1 (I: Major inducement events of the accident

    Directory of Open Access Journals (Sweden)

    Yingfeng Meng

    2015-12-01

    Full Text Available Nitrogen drilling in poor tight gas sandstone should be safe because of very low gas production. But a serious accident of fire blowout occurred during nitrogen drilling of Well Qionglai 1. This is the first nitrogen drilling accident in China, which was beyond people's knowledge about the safety of nitrogen drilling and brought negative effects on the development of gas drilling technology still in start-up phase and resulted in dramatic reduction in application of gas drilling. In order to form a correct understanding, the accident was systematically analyzed, the major events resulting in this accident were inferred. It is discovered for the first time that violent ejection of rock clasts and natural gas occurred due to the sudden burst of downhole rock when the fractured tight gas zone was penetrated during nitrogen drilling, which has been named as “rock burst and blowout by gas bomb”, short for “rock burst”. Then all the induced events related to the rock burst are as following: upthrust force on drilling string from rock burst, bridging-off formed and destructed repeatedly at bit and centralizer, and so on. However, the most direct important event of the accident turns out to be the blockage in the blooie pipe from rock burst clasts and the resulted high pressure at the wellhead. The high pressure at the wellhead causes the blooie pipe to crack and trigged blowout and deflagration of natural gas, which is the direct presentation of the accident.

  13. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel

  14. Application of the accident management information needs methodology to a severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E. (Nuclear Regulatory Commission, Washington, DC (USA))

    1989-11-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel.

  15. Some implications of an event-based definition of exposure to the risk of road accident

    DEFF Research Database (Denmark)

    Elvik, Rune

    2015-01-01

    This paper proposes a new definition of exposure to the risk of road accident as any event, limited in space and time, representing a potential for an accident to occur by bringing road users close to each other in time or space of by requiring a road user to take action to avoid leaving the road......This paper proposes a new definition of exposure to the risk of road accident as any event, limited in space and time, representing a potential for an accident to occur by bringing road users close to each other in time or space of by requiring a road user to take action to avoid leaving...

  16. Heath Effects Sequence of Meet Halfa Radiological Accident After Twelve Years

    International Nuclear Information System (INIS)

    Shabon, M.H.

    2013-01-01

    The accident of Meet-Halfa developed consequent upon the loss of an industrial gamma radiography source. The source was found by a farmer resident of Meet-Halfa who took it to his house occupied by his family. The sequence of events developed over a period of seven weeks from the time the source was found on May 5, 2000, till the day of its retrieval from the house by the national authorities on June 26. The protracted exposure patterns of the family members during the period of source possession are not precisely known, however these exposures resulted in two fatalities, clinical forms of bone marrow depression, and several skin burns of different severities. The recent sequences of the accident is as follows:-The three survived siblings married and get good children. That mean there is no hereditary stochastic effects. The sister died at 2007 with 72 years old with senility and no specific disease. The youngest daughter amputate the left thumb and index fingers at 2001. The elder son amputate the terminal phalanx of the right thumb at 2009. The youngest daughter amputate the right index finger at 2009. The elder son graft the burn at the lower right quadrant of the abdomen for more than 20 times (3 of them were in the Mansheat Al-Bakry Millitary Hospital), but there is residual of burn untill now. Sever abdominal hernia in the elder son due to necroses in the right quadrant abdominal muscles. Grafting for these muscles occur but failed.

  17. Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1983-01-01

    Since its inception in October 1980, the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory (ORNL) has completed four studies including Station Blackout, Scram Discharge Volume Break, Loss of Decay Heat Removal, and Loss of Injection accident sequences for the Browns Ferry Nuclear Plant. The accident analyses incorporated in a SASA study provide much greater detail than that practically achievable in a Probabilistic Risk Assessment (PRA). When applied to the candidate dominant accident sequences identified by a PRA, the detailed SASA results determine if factors neglected by the PRA would have a significant effect on the order of dominant sequences. Ongoing SASA work at ORNL involves the analysis of Anticipated Transients Without Scram (ATWS) sequences for Browns Ferry

  18. The sequence coding and search system: an approach for constructing and analyzing event sequences at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Mays, G.T.

    1990-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has recognized the importance of the collection, assessment, and feedback of operating experience data from commercial nuclear power plants and has centralized these activities in the Office for Analysis and Evaluation of Operational Data (AEOD). Such data is essential for performing safety and reliability analyses, especially analyses of trends and patterns to identify undesirable changes in plant performance at the earliest opportunity to implement corrective measures to preclude the occurrence of a more serious event. One of NRC's principal tools for collecting and evaluating operating experience data is the Sequence Coding and Search System (SCSS). The SCSS consists of a methodology for structuring event sequences and the requisite computer system to store and search the data. The source information for SCSS is the Licensee Event Report (LER), which is a legally required document. This paper describes the objectives of SCSS, the information it contains, and the format and approach for constructing SCSS event sequences. Examples are presented demonstrating the use of SCSS to support the analysis of LER data. The SCSS contains over 30,000 LERs describing events from 1980 through the present. Insights gained from working with a complex data system from the initial developmental stage to the point of a mature operating system are highlighted. Considerable experience has been gained in the areas of evolving and changing data requirements, staffing requirements, and quality control and quality assurance procedures for addressing consistency, software/hardware considerations for developing and maintaining a complex system, documentation requirements, and end-user needs. Two other approaches for constructing and evaluating event sequences are examined including the Accident Precursor Program (ASP) where sequences having the potential for core damage are identified and analyzed, and the Significant Event Compilation Tree

  19. A sequence of events across the Cretaceous-Tertiary boundary

    NARCIS (Netherlands)

    Smit, J.; Romein, A.J.T.

    1985-01-01

    The lithological and biological sequence of events across the Cretaceous-Tertiary (K/T), as developed in thick and complete landbased sections and termed the standard K/T event sequence, is also found in many DSDP cores from all over the globe. Microtektite-like spherules have been found in

  20. Human factors review for nuclear power plant severe accident sequence analysis

    International Nuclear Information System (INIS)

    Krois, P.A.; Haas, P.M.

    1985-01-01

    The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release

  1. Development of Safety Significance Evaluation Program for Accidents and Events in NPPs

    International Nuclear Information System (INIS)

    Yang, Hui Chang; Hong, Seok Jin; Cho, Nam Chul; Chung, Dae Wook; Lee, Chang Joo

    2010-01-01

    To evaluate the significance in terms of safety for the accidents and events occurred in nuclear power plants using probabilistic safety assessment techniques can provide useful insights to the regulator. Based on the quantified risk information of accident or event occurred, regulators can decide which regulatory areas should be focused than the others. To support these regulatory analysis activities, KINS-ASP program was developed. KINS-ASP program can supports the risk increase due to the occurred accidents or events by providing the graphic interfaces and linked quantification engines for the PSA experts and non- PSA acquainted regulators both

  2. On the sequence and consequences of the Chernobyl reactor accident

    Energy Technology Data Exchange (ETDEWEB)

    Hennies, H H

    1986-01-01

    A serious reactor accident occurred on April 26, 1986 at Chernobyl near Kiev (Soviet Union) where, after melting of the core, there was a considerable release of radioactivity to the environment and to the atmosphere. The radioactivity release caused irradiation of the operating staff, which led to 24 deaths by June 1986. Hardly anything is known about the irradiation of the environment of the reactor plant, but the population within a radius of 30 km was evacuated. The radioactivity released into the atmosphere spread all over Europe, and Germany was affected a few days after the accident. The article gives a short description of the plant which suffered the accident, one tries to describe the course of the accident and to discuss the applicability to German plants.

  3. Life Change Events as a Predictor of Accident Incidence in a College Population.

    Science.gov (United States)

    Furney, Steven R.

    1983-01-01

    To test the relationship between stressful life-change events and accident incidence, researchers administered the College Schedule of Recent Experience to male students at a large midwestern university. The study's implications for identifying high-risk persons and for accident prevention are discussed. (PP)

  4. Accident progression event tree analysis for postulated severe accidents at N Reactor

    International Nuclear Information System (INIS)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied

  5. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

  6. Incident sequence analysis; event trees, methods and graphical symbols

    International Nuclear Information System (INIS)

    1980-11-01

    When analyzing incident sequences, unwanted events resulting from a certain cause are looked for. Graphical symbols and explanations of graphical representations are presented. The method applies to the analysis of incident sequences in all types of facilities. By means of the incident sequence diagram, incident sequences, i.e. the logical and chronological course of repercussions initiated by the failure of a component or by an operating error, can be presented and analyzed simply and clearly

  7. Accident analyses in nuclear power plants following external initiating events and in the shutdown state. Final report; Unfallanalysen in Kernkraftwerken nach anlagenexternen ausloesenden Ereignissen und im Nichtleistungsbetrieb. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Loeffler, Horst; Kowalik, Michael; Mildenberger, Oliver; Hage, Michael

    2016-06-15

    The work which is documented here provides the methodological basis for improvement of the state of knowledge for accident sequences after plant external initiating events and for accident sequences which begin in the shutdown state. The analyses have been done for a PWR and for a BWR reference plant. The work has been supported by the German federal ministry BMUB under the label 3612R01361. Top objectives of the work are: - Identify relevant event sequences in order to define characteristic initial and boundary conditions - Perform accident analysis of selected sequences - Evaluate the relevance of accident sequences in a qualitative way The accident analysis is performed with the code MELCOR 1.8.6. The applied input data set has been significantly improved compared to previous analyses. The event tree method which is established in PSA level 2 has been applied for creating a structure for a unified summarization and evaluation of the results from the accident analyses. The computer code EVNTRE has been applied for this purpose. In contrast to a PSA level 2, the branching probabilities of the event tree have not been determined with the usual accuracy, but they are given in an approximate way only. For the PWR, the analyses show a considerable protective effect of the containment also in the case of beyond design events. For the BWR, there is a rather high probability for containment failure under core melt impact, but nevertheless the release of radionuclides into the environment is very limited because of plant internal retention mechanisms. This report concludes with remarks about existing knowledge gaps and with regard to core melt sequences, and about possible improvements of the plant safety.

  8. The tsunami probabilistic risk assessment (PRA). Example of accident sequence analysis of tsunami PRA according to the standard for procedure of tsunami PRA for nuclear power plants

    International Nuclear Information System (INIS)

    Ohara, Norihiro; Hasegawa, Keiko; Kuroiwa, Katsuya

    2013-01-01

    After the Fukushima Daiichi nuclear power plant (NPP) accident, standard for procedure of tsunami PRA for NPP had been established by the Standardization Committee of AESJ. Industry group had been conducting analysis of Tsunami PRA for PWR based on the standard under the cooperation with electric utilities. This article introduced overview of the standard and examples of accident sequence analysis of Tsunami PRA studied by the industry group according to the standard. The standard consisted of (1) investigation of NPP's composition, characteristics and site information, (2) selection of relevant components for Tsunami PRA and initiating events and identification of accident sequence, (3) evaluation of Tsunami hazards, (4) fragility evaluation of building and components and (5) evaluation of accident sequence. Based on the evaluation, countermeasures for further improvement of safety against Tsunami could be identified by the sensitivity analysis. (T. Tanaka)

  9. Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

    1995-08-01

    The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

  10. Swallow Event Sequencing: Comparing Healthy Older and Younger Adults.

    Science.gov (United States)

    Herzberg, Erica G; Lazarus, Cathy L; Steele, Catriona M; Molfenter, Sonja M

    2018-04-23

    Previous research has established that a great deal of variation exists in the temporal sequence of swallowing events for healthy adults. Yet, the impact of aging on swallow event sequence is not well understood. Kendall et al. (Dysphagia 18(2):85-91, 2003) suggested there are 4 obligatory paired-event sequences in swallowing. We directly compared adherence to these sequences, as well as event latencies, and quantified the percentage of unique sequences in two samples of healthy adults: young ( 65). The 8 swallowing events that contribute to the sequences were reliably identified from videofluoroscopy in a sample of 23 healthy seniors (10 male, mean age 74.7) and 20 healthy young adults (10 male, mean age 31.5) with no evidence of penetration-aspiration or post-swallow residue. Chi-square analyses compared the proportions of obligatory pairs and unique sequences by age group. Compared to the older subjects, younger subjects had significantly lower adherence to two obligatory sequences: Upper Esophageal Sphincter (UES) opening occurs before (or simultaneous with) the bolus arriving at the UES and UES maximum distention occurs before maximum pharyngeal constriction. The associated latencies were significantly different between age groups as well. Further, significantly fewer unique swallow sequences were observed in the older group (61%) compared with the young (82%) (χ 2  = 31.8; p < 0.001). Our findings suggest that paired swallow event sequences may not be robust across the age continuum and that variation in swallow sequences appears to decrease with aging. These findings provide normative references for comparisons to older individuals with dysphagia.

  11. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  12. Containment event analysis for postulated severe accidents: Peach Bottom Atomic Power Station, Unit 2. Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Griesmeyer, J M [Sandia National Laboratories, Albuquerque, NM (United States); Kolaczkowski, A M [Science Applications International Corporation, Albuquerque, NM (United States)

    1987-05-01

    A study has been performed as part of the Severe Accident Risk Reduction Program (SARRP) to investigate the response of a particular boiling water reactor with a Mark I containment (Peach Bottom Unit 2) to postulated severe accidents. A detailed containment event tree for the Peach Bottom plant has been developed to describe the various possible accident pathways that can lead to radioactive releases from containment. Data and analyses from a large number of NRC and industry-sponsored programs have been reviewed and used as a basis for quantifying the event tree, i.e., determining the likelihood of the pathways at each branch point for a variety of accident sequence initiators. A generalized containment event tree code, called EVNTRE, has been developed to facilitate the quantification. The uncertainty in the results has been examined by performing the quantification three times, using a different set of input each time to represent the variation of opinion in the reactor safety community. In the so-called 'central' estimate, the likelihood of early containment failure (occurring before or within a short time after reactor vessel breach) was found to be significant because of the possible occurrence of the following phenomena that can threaten containment integrity: (1) meltthrough of the drywell shell caused by thermal attack from core debris, and (2) drywell overpressurization caused by rapid depressurization of the reactor vessel in combination with other events such as direct heating. However, uncertainties surrounding these issues could cause the early failure likelihood to be significantly lower than in the central estimate. This work supports NRC's assessment of severe accident risks to be published in NUREG-1150. (author)

  13. A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

    International Nuclear Information System (INIS)

    Svenson, Ola

    1998-01-01

    This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

  14. Contrasting safety assessments of a runway incursion scenario: Event sequence analysis versus multi-agent dynamic risk modelling

    International Nuclear Information System (INIS)

    Stroeve, Sybert H.; Blom, Henk A.P.; Bakker, G.J.

    2013-01-01

    In the safety literature it has been argued, that in a complex socio-technical system safety cannot be well analysed by event sequence based approaches, but requires to capture the complex interactions and performance variability of the socio-technical system. In order to evaluate the quantitative and practical consequences of these arguments, this study compares two approaches to assess accident risk of an example safety critical sociotechnical system. It contrasts an event sequence based assessment with a multi-agent dynamic risk model (MA-DRM) based assessment, both of which are performed for a particular runway incursion scenario. The event sequence analysis uses the well-known event tree modelling formalism and the MA-DRM based approach combines agent based modelling, hybrid Petri nets and rare event Monte Carlo simulation. The comparison addresses qualitative and quantitative differences in the methods, attained risk levels, and in the prime factors influencing the safety of the operation. The assessments show considerable differences in the accident risk implications of the performance of human operators and technical systems in the runway incursion scenario. In contrast with the event sequence based results, the MA-DRM based results show that the accident risk is not manifest from the performance of and relations between individual human operators and technical systems. Instead, the safety risk emerges from the totality of the performance and interactions in the agent based model of the safety critical operation considered, which coincides very well with the argumentation in the safety literature.

  15. Long-Term Recall of Event Sequences in Infancy.

    Science.gov (United States)

    Mandler, Jean M.; McDonough, Laraine

    1995-01-01

    Two experiments demonstrated that 11-month olds can encode novel causal events from a brief period of observational learning and recall much of the information after 24 hours and after 3 months. The infants remembered more individual actions than whole sequences, but reproduced many of the events in their entirety after the long delay. (MDM)

  16. Memory for sequences of events impaired in typical aging

    Science.gov (United States)

    Allen, Timothy A.; Morris, Andrea M.; Stark, Shauna M.; Fortin, Norbert J.

    2015-01-01

    Typical aging is associated with diminished episodic memory performance. To improve our understanding of the fundamental mechanisms underlying this age-related memory deficit, we previously developed an integrated, cross-species approach to link converging evidence from human and animal research. This novel approach focuses on the ability to remember sequences of events, an important feature of episodic memory. Unlike existing paradigms, this task is nonspatial, nonverbal, and can be used to isolate different cognitive processes that may be differentially affected in aging. Here, we used this task to make a comprehensive comparison of sequence memory performance between younger (18–22 yr) and older adults (62–86 yr). Specifically, participants viewed repeated sequences of six colored, fractal images and indicated whether each item was presented “in sequence” or “out of sequence.” Several out of sequence probe trials were used to provide a detailed assessment of sequence memory, including: (i) repeating an item from earlier in the sequence (“Repeats”; e.g., ABADEF), (ii) skipping ahead in the sequence (“Skips”; e.g., ABDDEF), and (iii) inserting an item from a different sequence into the same ordinal position (“Ordinal Transfers”; e.g., AB3DEF). We found that older adults performed as well as younger controls when tested on well-known and predictable sequences, but were severely impaired when tested using novel sequences. Importantly, overall sequence memory performance in older adults steadily declined with age, a decline not detected with other measures (RAVLT or BPS-O). We further characterized this deficit by showing that performance of older adults was severely impaired on specific probe trials that required detailed knowledge of the sequence (Skips and Ordinal Transfers), and was associated with a shift in their underlying mnemonic representation of the sequences. Collectively, these findings provide unambiguous evidence that the

  17. Sequence Synopsis: Optimize Visual Summary of Temporal Event Data.

    Science.gov (United States)

    Chen, Yuanzhe; Xu, Panpan; Ren, Liu

    2018-01-01

    Event sequences analysis plays an important role in many application domains such as customer behavior analysis, electronic health record analysis and vehicle fault diagnosis. Real-world event sequence data is often noisy and complex with high event cardinality, making it a challenging task to construct concise yet comprehensive overviews for such data. In this paper, we propose a novel visualization technique based on the minimum description length (MDL) principle to construct a coarse-level overview of event sequence data while balancing the information loss in it. The method addresses a fundamental trade-off in visualization design: reducing visual clutter vs. increasing the information content in a visualization. The method enables simultaneous sequence clustering and pattern extraction and is highly tolerant to noises such as missing or additional events in the data. Based on this approach we propose a visual analytics framework with multiple levels-of-detail to facilitate interactive data exploration. We demonstrate the usability and effectiveness of our approach through case studies with two real-world datasets. One dataset showcases a new application domain for event sequence visualization, i.e., fault development path analysis in vehicles for predictive maintenance. We also discuss the strengths and limitations of the proposed method based on user feedback.

  18. Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2015-01-01

    The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

  19. Selection of events at Ukrainian NPPs using the algorithm based on accident precursor method

    International Nuclear Information System (INIS)

    Vorontsov, D.V.; Lyigots'kij, O.Yi.; Serafin, R.Yi.; Tkachova, L.M.

    2012-01-01

    The paper describes a general approach to the first stage of research and development on analysis of Ukrainian NPP operation events from 1 January 2000 to 31 December 2010 using the accident precursor approach. Groups of potentially important events formed after their selection and classification are provided

  20. Some implications of an event-based definition of exposure to the risk of road accident.

    Science.gov (United States)

    Elvik, Rune

    2015-03-01

    This paper proposes a new definition of exposure to the risk of road accident as any event, limited in space and time, representing a potential for an accident to occur by bringing road users close to each other in time or space of by requiring a road user to take action to avoid leaving the roadway. A typology of events representing a potential for an accident is proposed. Each event can be interpreted as a trial as defined in probability theory. Risk is the proportion of events that result in an accident. Defining exposure as events demanding the attention of road users implies that road users will learn from repeated exposure to these events, which in turn implies that there will normally be a negative relationship between exposure and risk. Four hypotheses regarding the relationship between exposure and risk are proposed. Preliminary tests support these hypotheses. Advantages and disadvantages of defining exposure as specific events are discussed. It is argued that developments in vehicle technology are likely to make events both observable and countable, thus ensuring that exposure is an operational concept. Copyright © 2014 Elsevier Ltd. All rights reserved.

  1. Methodology to classify accident sequences of an Individual Plant Examination according to the severe releases for BWR type reactors

    International Nuclear Information System (INIS)

    Sandoval V, S.

    2001-01-01

    The Light Water Reactor (LWR) operation regulations require to every operating plant to perform of an Individual Plant Examination study (Ipe). One of the main purposes of an Ipe is t o gain a more quantitative understanding of the overall probabilities of core damage and fission product releases . Probabilistic Safety Analysis (PSA) methodologies and Severe Accident Analysis are used to perform Ipe studies. PSA methodologies are used to identify and analyse the set of event sequences that might originate the fission product release from a nuclear power plant; these methodologies are combinatorial in nature and generate thousands of sequences. Among other uses within an Ipe, severe accident simulations are used to determine the characteristics of the fission product release for the identified sequences and in this way, the releases can be understood and characterized. A vast amount of resources is required to simulate and analyse every Ipe sequence. This effort is unnecessary if similar sequences are grouped. The grouping scheme must achieve an efficient trade off between problem reduction and accuracy. The methodology presented in this work enables an accurate characterization and analysis of the Ipe fission product releases by using a reduced problem. The methodology encourages the use of specific plant simulations. (Author)

  2. CESAS: Computerized event sequence abstracting system outlines and applications

    International Nuclear Information System (INIS)

    Watanabe, N.; Kobayashi, K.; Fujiki, K.

    1990-01-01

    For the purpose of efficient utilization of the safety-related event information on the nuclear power plants, a new computer software package CESAS has been under development. CESAS is to systematically abstract the event sequence, that is a series of sequential and causal relationships between occurrences, from the event description written in natural language of English. This system is designed to be based on the knowledge engineering technique utilized in the field of natural language processing. The analytical process in this system consists of morphemic, syntactic, semantic, and syntagmatic analyses. At this moment, the first version of CESAS has been developed and applied to several real event descriptions for studying its feasibility. This paper describes the outlines of CESAS and one of analytical results in comparison with a manually-extracted event sequence

  3. Differentially Private Event Histogram Publication on Sequences over Graphs

    Institute of Scientific and Technical Information of China (English)

    Ning Wang; Yu Gu; Jia Xu; Fang-Fang Li; Ge Yu

    2017-01-01

    The big data era is coming with strong and ever-growing demands on analyzing personal information and footprints in the cyber world. To enable such analysis without privacy leak risk, differential privacy (DP) has been quickly rising in recent years, as the first practical privacy protection model with rigorous theoretical guarantee. This paper discusses how to publish differentially private histograms on events in time series domain, with sequences of personal events over graphs with events as edges. Such individual-generated sequences commonly appear in formalized industrial workflows, online game logs, and spatial-temporal trajectories. Directly publishing the statistics of sequences may compromise personal privacy. While existing DP mechanisms mainly target at normalized domains with fixed and aligned dimensions, our problem raises new challenges when the sequences could follow arbitrary paths on the graph. To tackle the problem, we reformulate the problem with a three-step framework, which 1) carefully truncates the original sequences, trading off errors introduced by the truncation with those introduced by the noise added to guarantee privacy, 2) decomposes the event graph into path sub-domains based on a group of event pivots, and 3) employs a deeply optimized tree-based histogram construction approach for each sub-domain to benefit with less noise addition. We present a careful analysis on our framework to support thorough optimizations over each step of the framework, and verify the huge improvements of our proposals over state-of-the-art solutions.

  4. A classification of event sequences in the influence network

    Science.gov (United States)

    Walsh, James Lyons; Knuth, Kevin H.

    2017-06-01

    We build on the classification in [1] of event sequences in the influence network as respecting collinearity or not, so as to determine in future work what phenomena arise in each case. Collinearity enables each observer to uniquely associate each particle event of influencing with one of the observer's own events, even in the case of events of influencing the other observer. We further classify events as to whether they are spacetime events that obey in the fine-grained case the coarse-grained conditions of [2], finding that Newton's First and Second Laws of motion are obeyed at spacetime events. A proof of Newton's Third Law under particular circumstances is also presented.

  5. Time Separation Between Events in a Sequence: a Regional Property?

    Science.gov (United States)

    Muirwood, R.; Fitzenz, D. D.

    2013-12-01

    Earthquake sequences are loosely defined as events occurring too closely in time and space to appear unrelated. Depending on the declustering method, several, all, or no event(s) after the first large event might be recognized as independent mainshocks. It can therefore be argued that a probabilistic seismic hazard assessment (PSHA, traditionally dealing with mainshocks only) might already include the ground shaking effects of such sequences. Alternatively all but the largest event could be classified as an ';aftershock' and removed from the earthquake catalog. While in PSHA the question is only whether to keep or remove the events from the catalog, for Risk Management purposes, the community response to the earthquakes, as well as insurance risk transfer mechanisms, can be profoundly affected by the actual timing of events in such a sequence. In particular the repetition of damaging earthquakes over a period of weeks to months can lead to businesses closing and families evacuating from the region (as happened in Christchurch, New Zealand in 2011). Buildings that are damaged in the first earthquake may go on to be damaged again, even while they are being repaired. Insurance also functions around a set of critical timeframes - including the definition of a single 'event loss' for reinsurance recoveries within the 192 hour ';hours clause', the 6-18 month pace at which insurance claims are settled, and the annual renewal of insurance and reinsurance contracts. We show how temporal aspects of earthquake sequences need to be taken into account within models for Risk Management, and what time separation between events are most sensitive, both in terms of the modeled disruptions to lifelines and business activity as well as in the losses to different parties (such as insureds, insurers and reinsurers). We also explore the time separation between all events and between loss causing events for a collection of sequences from across the world and we point to the need to

  6. Event tree analysis of accidents during transport of radioactive materials in Japan

    International Nuclear Information System (INIS)

    Watabe, N.; Shirai, K.; Noguchi, K.; Suzuki, H.; Kinehara, Y.

    1993-01-01

    The Event Tree Method is one of the Probabilistic Safety Assessment Method. It introduces the accident scenario and the results of countermeasures. Therefore, it is effective in determining latent accident scenarios in the transfer. In this report the Event Tree Method is used to study the tunnel fire and its effects are evaluated. And this is the first trail of our Probabilistic Safety Assessment. The Event Tree for determining the early conditions when a car engine catches fire in a tunnel is examined. There are fire extinguishers, tunnel equipments for fire-fighting, fire stations and the heat-resisting property of the container for protecting from the fire. The protection level against the over 800degC-30min. fire accident is 88.3 %. (J.P.N.)

  7. Accident and Off-Normal Response and Recovery from Multi-Canister Overpack (MCO) Processing Events

    International Nuclear Information System (INIS)

    ALDERMAN, C.A.

    2000-01-01

    In the process of removing spent nuclear fuel (SNF) from the K Basins through its subsequent packaging, drymg, transportation and storage steps, the SNF Project must be able to respond to all anticipated or foreseeable off-normal and accident events that may occur. Response procedures and recovery plans need to be in place, personnel training established and implemented to ensure the project will be capable of appropriate actions. To establish suitable project planning, these events must first be identified and analyzed for their expected impact to the project. This document assesses all off-normal and accident events for their potential cross-facility or Multi-Canister Overpack (MCO) process reversal impact. Table 1 provides the methodology for establishing the event planning level and these events are provided in Table 2 along with the general response and recovery planning. Accidents and off-normal events of the SNF Project have been evaluated and are identified in the appropriate facility Safety Analysis Report (SAR) or in the transportation Safety Analysis Report for Packaging (SARP). Hazards and accidents are summarized from these safety analyses and listed in separate tables for each facility and the transportation system in Appendix A, along with identified off-normal events. The tables identify the general response time required to ensure a stable state after the event, governing response documents, and the events with potential cross-facility or SNF process reversal impacts. The event closure is predicated on stable state response time, impact to operations and the mitigated annual occurrence frequency of the event as developed in the hazard analysis process

  8. An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

    International Nuclear Information System (INIS)

    Di Salvo, R.; Leonard, M.T.; Wreathall, J.

    1986-01-01

    An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

  9. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

    2017-03-15

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

  10. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    International Nuclear Information System (INIS)

    Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

    2017-01-01

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective

  11. Seismically induced accident sequence analysis of the advanced test reactor

    International Nuclear Information System (INIS)

    Khericha, S.T.; Henry, D.M.; Ravindra, M.K.; Hashimoto, P.S.; Griffin, M.J.; Tong, W.H.; Nafday, A.M.

    1991-01-01

    A seismic probabilistic risk assessment (PRA) was performed for the Department of Energy (DOE) Advanced Test Reactor (ATR) as part of the external events analysis. The risk from seismic events to the fuel in the core and in the fuel storage canal was evaluated. The key elements of this paper are the integration of seismically induced internal flood and internal fire, and the modeling of human error rates as a function of the magnitude of earthquake. The systems analysis was performed by EG ampersand G Idaho, Inc. and the fragility analysis and quantification were performed by EQE International, Inc. (EQE)

  12. Source terms associated with two severe accident sequences in a 900 MWe PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

    1983-12-01

    Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

  13. A methodology for analyzing precursors to earthquake-initiated and fire-initiated accident sequences

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Lambert, H.E.; Apostolakis, G.

    1998-04-01

    This report covers work to develop a methodology for analyzing precursors to both earthquake-initiated and fire-initiated accidents at commercial nuclear power plants. Currently, the U.S. Nuclear Regulatory Commission sponsors a large ongoing project, the Accident Sequence Precursor project, to analyze the safety significance of other types of accident precursors, such as those arising from internally-initiated transients and pipe breaks, but earthquakes and fires are not within the current scope. The results of this project are that: (1) an overall step-by-step methodology has been developed for precursors to both fire-initiated and seismic-initiated potential accidents; (2) some stylized case-study examples are provided to demonstrate how the fully-developed methodology works in practice, and (3) a generic seismic-fragility date base for equipment is provided for use in seismic-precursors analyses. 44 refs., 23 figs., 16 tabs

  14. Protection of the Population in the event of a Nuclear accident. A Basis for Intervention

    International Nuclear Information System (INIS)

    1990-01-01

    During the years following the Chernobyl accident in 1986, the NEA actively participated in the international effort towards the improvement and better harmonization of the international and national criteria for the protection of the public in the event of a nuclear accident. A first report on this matter, titled Nuclear Accidents: Intervention Levels for Protection of the Public was published by the NEA in 1989. Subsequently, the NEA Committee on Radiation Protection and Public Health set up a small Task Group to provide additional guidance, and to take into account recent developments in other international organizations. The report outlines the status of relevant international activities in the period following the preparation of the 1989 report, discusses the intervention principles and describes both the proposed accident management system and a general scheme for its application. It is to be noted that the principles and criteria for intervention discussed in this report, although developed with specific reference to reactor accidents, apply equally well to activities and possible accidents at other nuclear facilities. The report briefly describes the transition from an accident management situation back to a normal situation and the related problem of changing criteria for the protection of the public. In addition to the traditional exposure pathways -inhalation from the cloud, external irradiation from the cloud and the ground and ingestion of food - the report acknowledges the existence of special pathways, proposing criteria for protecting workers and the public and some examples of their application

  15. iROCS: Integrated accident management framework for coping with beyond-design-basis external events

    International Nuclear Information System (INIS)

    Kim, Jaewhan; Park, Soo-Yong; Ahn, Kwang-Il; Yang, Joon-Eon

    2016-01-01

    Highlights: • An integrated mitigating strategy to cope with extreme external events, iROCS, is proposed. • The strategy aims to preserve the integrity of the reactor vessel as well as core cooling. • A case study for an extreme damage state is performed to assess the effectiveness and feasibility of candidate mitigation strategies under an extreme event. - Abstract: The Fukushima Daiichi accident induced by the Great East Japan earthquake and tsunami on March 11, 2011, poses a new challenge to the nuclear society, especially from an accident management viewpoint. This paper presents a new accident management framework called an integrated, RObust Coping Strategy (iROCS) to cope with beyond-design-basis external events (BDBEEs). The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. From a case study for an extreme damage condition, it showed that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against accident scenarios and extreme damage conditions of the site, especially when employing mobile or portable equipment under BDBEEs within the limited time available to achieve desired goals such as prevention of core damage as well as a reactor vessel failure.

  16. Self-Exciting Point Process Modeling of Conversation Event Sequences

    Science.gov (United States)

    Masuda, Naoki; Takaguchi, Taro; Sato, Nobuo; Yano, Kazuo

    Self-exciting processes of Hawkes type have been used to model various phenomena including earthquakes, neural activities, and views of online videos. Studies of temporal networks have revealed that sequences of social interevent times for individuals are highly bursty. We examine some basic properties of event sequences generated by the Hawkes self-exciting process to show that it generates bursty interevent times for a wide parameter range. Then, we fit the model to the data of conversation sequences recorded in company offices in Japan. In this way, we can estimate relative magnitudes of the self excitement, its temporal decay, and the base event rate independent of the self excitation. These variables highly depend on individuals. We also point out that the Hawkes model has an important limitation that the correlation in the interevent times and the burstiness cannot be independently modulated.

  17. Fault trees based on past accidents. Factorial analysis of events

    International Nuclear Information System (INIS)

    Vaillant, M.

    1977-01-01

    The method of the fault tree is already useful in the qualitative step before any reliability calculation. The construction of the tree becomes even simpler when we just want to describe how the events happened. Differently from screenplays that introduce several possibilities by means of the conjunction OR, you only have here the conjunction AND, which will not be written at all. This method is presented by INRS (1) for the study of industrial injuries; it may also be applied to material damages. (orig.) [de

  18. ES-RBE Event sequence reliability Benchmark exercise

    International Nuclear Information System (INIS)

    Poucet, A.E.J.

    1991-01-01

    The event Sequence Reliability Benchmark Exercise (ES-RBE) can be considered as a logical extension of the other three Reliability Benchmark Exercices : the RBE on Systems Analysis, the RBE on Common Cause Failures and the RBE on Human Factors. The latter, constituting Activity No. 1, was concluded by the end of 1987. The ES-RBE covered the techniques that are currently used for analysing and quantifying sequences of events starting from an initiating event to various plant damage states, including analysis of various system failures and/or successes, human intervention failure and/or success and dependencies between systems. By this way, one of the scopes of the ES-RBE was to integrate the experiences gained in the previous exercises

  19. Harmonic spectral components in time sequences of Markov correlated events

    Science.gov (United States)

    Mazzetti, Piero; Carbone, Anna

    2017-07-01

    The paper concerns the analysis of the conditions allowing time sequences of Markov correlated events give rise to a line power spectrum having a relevant physical interest. It is found that by specializing the Markov matrix in order to represent closed loop sequences of events with arbitrary distribution, generated in a steady physical condition, a large set of line spectra, covering all possible frequency values, is obtained. The amplitude of the spectral lines is given by a matrix equation based on a generalized Markov matrix involving the Fourier transform of the distribution functions representing the time intervals between successive events of the sequence. The paper is a complement of a previous work where a general expression for the continuous power spectrum was given. In that case the Markov matrix was left in a more general form, thus preventing the possibility of finding line spectra of physical interest. The present extension is also suggested by the interest of explaining the emergence of a broad set of waves found in the electro and magneto-encephalograms, whose frequency ranges from 0.5 to about 40Hz, in terms of the effects produced by chains of firing neurons within the complex neural network of the brain. An original model based on synchronized closed loop sequences of firing neurons is proposed, and a few numerical simulations are reported as an application of the above cited equation.

  20. Below Regulatory Conern Owners Group: Radiologic impact of accidents and unexpected events from disposal of BRC waste

    International Nuclear Information System (INIS)

    Waite, D.A.; Dolan, M.M.; Rish, W.R.; Rossi, A.J.; McCourt, J.E.

    1989-07-01

    This report determines the radiological impact of accidents and unexpected events in the disposal of Below Regulatory Concern (BRC) waste. The accident analysis considers the transportation, incineration, and disposal of BRC waste as municipal solid waste. The potential greatest radiological impact for each type of accident is identified through the use of event trees. These accident events are described in terms of the generic waste property(ies) (e.g., flammability, dispersibility, leachability, and solubility) that cause the greatest radiological impact. 7 refs., 32 figs., 12 tabs

  1. Recovery operations in the event of a nuclear accident or radiological emergency

    International Nuclear Information System (INIS)

    1990-01-01

    Much progress has been made over the last decade in the field of emergency planning and preparedness, including the development of guidance, criteria, training programmes, regulations and comprehensive plans in the support of nuclear facilities. To provide a forum for international review and discussion of actual experiences gained and lessons learned from the different aspects of recovery techniques and operations in response to serious accidents at nuclear facilities and accidents associated with radioactive materials, the IAEA organized the International Symposium on Recovery Operations in the Event of a Nuclear Accident or Radiological Emergency. The symposium was held from 6 to 10 November 1989 in Vienna, Austria, and was attended by over 250 experts from 35 Member State and 7 international organizations. Although the prime focus was on on-site and off-site recovery from nuclear reactor accidents and on recovery from radiological accidents unrelated to nuclear power plants, development of emergency planning and preparedness resources was covered as well. From the experiences reported, lessons learned were identified. While further work remains to be done to improve concepts, plans, materials, communications and mechanisms to assemble quickly all the special resources needed in the event of an accident, there was general agreement that worldwide preparations to handle any possible future radiological emergencies had vastly improved. A special feature of the symposium programme was the inclusion of a full session on an accident involving a chemical explosion in a high level waste tank a a plutonium extraction plant in the Southern Urals in the USSR in 1957. Information was presented on the radioactive release, its dissemination and deposition, the resultant radiation situation, dose estimates, health effects follow-up, and the rehabilitation of contaminated land. This volume contains the full text of the 49 papers presented at the symposium together with a

  2. The next nuclear power station generation: Beyond-design accident concepts, methods, and action sequence

    International Nuclear Information System (INIS)

    Asmolov, V.G.; Khakh, O.Ya.; Shashkov, M.G.

    1993-01-01

    The problem of beyond-design accidents at nuclear stations will not be solved unless a safety culture becomes a basic characteristic of all lines of activity. Only then can the danger of accidents as an objective feature of nuclear stations be eliminated by purposive skilled and responsible activities of those implementing safety. Nuclear-station safety is provided by the following interacting and complementary lines of activity: (1) the design and construction of nuclear stations by properly qualified design and building organizations; (2) monitoring and supervision of safety by special state bodies; (3) control of the station by the exploiting organization; and (4) scientific examination of safety within the above framework and by independent organizations. The distribution of the responsibilities, powers, and right in these lines should be defined by a law on atomic energy, but there is not such law in Russian. The beyond-design accident problem is a key one in nuclear station safety, as it clear from the serious experience with accidents and numerous probabilistic studies. There are four features of the state of this topic in Russia that are of major significance for managing accidents: the lack of an atomic energy law, the inadequacy of the technical standards, the lack of a verified program package for nuclear-station designs in order to calculate the beyond-design accidents and analyze risks, and a lack of approach by designers to such accidents on the basis of international recommendations. This paper gives a brief description of three-forming points in the scientific activity: the general concept of nuclear-station safety, methods of analyzing and providing accident management, and the sequence of actions developed by specialists at this institute in recent years

  3. Development of accident event trees and evaluation of safety system failure modes for the nuclear ultra large crude carrier

    International Nuclear Information System (INIS)

    Lewe, C.K.; Coffey, R.S.; Goodwin, E.F.; Maltese, J.G.; Pyatt, D.W.

    1978-01-01

    A method of applying the probabilistic accident event tree methodology to safety assessments of a nuclear powered Ultra Large Crude Carrier is presented. Also presented are the procedures by which an external accident initiating event, such as a ship collision, may be correlated with the probabilities of damage to the ship's safety systems and to their ultimate availabilities to perform required safety functions

  4. Licensee Event Report sequence coding and search procedure workshop

    International Nuclear Information System (INIS)

    Cottrell, W.B.; Gallaher, R.B.

    1981-01-01

    Since mid-1980, the Office for Analysis and Evaluation of Operational Data (AEOD) of the Nuclear Regulatory Commission (NRC) has been developing procedures for the systematic review and analysis of Licensee Event Reports (LERs). These procedures generally address several areas of concern, including identification of significant trends and patterns, event sequence of occurrences, component failures, and system and plant effects. The AEOD and NSIC conducted a workshop on the new coding procedure at the American Museum of Science and Energy in Oak Ridge, TN, on November 24, 1980

  5. Results of the event sequence reliability benchmark exercise

    International Nuclear Information System (INIS)

    Silvestri, E.

    1990-01-01

    The Event Sequence Reliability Benchmark Exercise is the fourth of a series of benchmark exercises on reliability and risk assessment, with specific reference to nuclear power plant applications, and is the logical continuation of the previous benchmark exercises on System Analysis Common Cause Failure and Human Factors. The reference plant is the Nuclear Power Plant at Grohnde Federal Republic of Germany a 1300 MW PWR plant of KWU design. The specific objective of the Exercise is to model, to quantify and to analyze such event sequences initiated by the occurrence of a loss of offsite power that involve the steam generator feed. The general aim is to develop a segment of a risk assessment, which ought to include all the specific aspects and models of quantification, such as common canal failure, Human Factors and System Analysis, developed in the previous reliability benchmark exercises, with the addition of the specific topics of dependences between homologous components belonging to different systems featuring in a given event sequence and of uncertainty quantification, to end up with an overall assessment of: - the state of the art in risk assessment and the relative influences of quantification problems in a general risk assessment framework. The Exercise has been carried out in two phases, both requiring modelling and quantification, with the second phase adopting more restrictive rules and fixing certain common data, as emerged necessary from the first phase. Fourteen teams have participated in the Exercise mostly from EEC countries, with one from Sweden and one from the USA. (author)

  6. Application of accident progression event tree technology to the Savannah River Site Defense Waste Processing Facility SAR analysis

    International Nuclear Information System (INIS)

    Brandyberry, M.D.; Baker, W.H.; Wittman, R.S.; Amos, C.N.

    1993-01-01

    The Accident Analysis in the Safety Analysis Report (SAR) for the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) has recently undergone an upgrade. Non-reactor SARs at SRS (and other Department of Energy (DOE) sites) use probabilistic techniques to assess the frequency of accidents at their facilities. This paper describes the application of an extension of the Accident Progression Event Tree (APET) approach to accidents at the SRS DWPF. The APET technique allows an integrated model of the facility risk to be developed, where previous probabilistic accident analyses have been limited to the quantification of the frequency and consequences of individual accident scenarios treated independently. Use of an APET allows a more structured approach, incorporating both the treatment of initiators that are common to more than one accident, and of accident progression at the facility

  7. International exchange of radiological information in the event of a nuclear accident - future perspectives

    International Nuclear Information System (INIS)

    De-Cort, M.; De-Vries, G.; Breitenbach, L.; Leeb, H.; Weiss, W.

    1996-01-01

    Immediately after the Chernobyl accident most European countries established or enhanced their national radioactivity monitoring and information systems. The large transboundary effect of the radioactive release also triggered the need for bilateral and international agreements on the exchange of radiological information in case of a nuclear accident. Based on the experiences gained from existing bi- and multilateral data exchange the Commission of the European Communities has made provision for and is developing technical systems to exchange information of common interest. Firstly the existing national systems and systems based on bilateral agreements are summarized. The objectives and technical realizations of the EC international information exchange systems ECURIE and EURDEP, are described. The experiences gained over the past few years and the concepts for the future, in which central and eastern European countries will be included, are discussed. The benefits that would result from improving the international exchange of radiological information in the event of a future nuclear accident are further being described

  8. A study on hydrogen deflagration for selected severe accident sequences in Ringhals 3

    Energy Technology Data Exchange (ETDEWEB)

    Gustavsson, V.; Moeller, E. [SwedPower AB (Sweden)

    2002-01-01

    In this report, we have investigated the most important severe accident sequences in Ringhals 3, a Westinghouse 3-loop PWR, concerning hydrogen generation and containment pressure at hydrogen deflagration. In order to analyze the accident sequences and to calculate the hydrogen production, the computer code MAAP (Modular Accident Analysis Program) was used. Six accident sequences were studied, where four were LOCA cases and two transients. MAAP gives the evolution of the accident and particularly the pressure in the containment and the production of hydrogen as a function of time. The pressure peaks at deflagration were calculated by the method AICC-Adiabatic Isochoric Complete Combustion. The results from these calculations are conservative for two reasons. Adiabatic combustion means that the heat losses to structures in the containment are neglected. The combustion is also assumed to occur once and all available hydrogen is burned. The maximum pressure in five analysed cases was compared with the failure pressure of the containment. In the LOCA case, 373 kg hydrogen was burned and the resulting peak pressure in the containment was 0,53 MPa. In the transient, where 720 kg hydrogen was burned, the peak pressure was 0,69 MPa. This is the same as the failure pressure of the containment. Finally, in the conservative case, 980 kg hydrogen was burned and the resulting peak pressure 0,96 MPa. However, it should be noted that these conclusions are conservative from two points of view. Firstly a more realistic (than AICC) calculation of the peak pressure would give a lower value than 0,69 MPa. Secondly, there is conservatism in the evaluation of the failure pressure. (au)

  9. 30 years of the Goiania Accident: a comparative study with other radioactivity dispersion events

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Ricardo Bastos; Vicente, Roberto, E-mail: rbsmith@ipen.br, E-mail: rvicente@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The year 2017 marks 30 years since the radioactive accident that occurred in the city of Goiania, capital of the state of Goias. It was the largest radiological accident in Brazil, and one of the largest in the world occurring outside nuclear facilities. Regarding the accidents at nuclear power plants, two of the biggest were Chernobyl in Ukraine, a year and a half before Goiania, and the Fukushima accident in Japan, in 2011. Different amounts of radioactive material were dispersed in the environment in each of these events. However, each one’s main pathway of dispersion was different: the accident of Goiania was terrestrial, Chernobyl was at the atmosphere, and Fukushima was mainly in the ocean. This work aims to study these different amounts, comparing such activities. In addition, it proposes to compare the sea dispersion of Fukushima with the amount of radioactive waste dumped in the oceans, when the release of radioactive waste at sea was permitted. It also proposes to compare the Chernobyl aerial dispersion with the radioactive material dissipated in the atmosphere, resulting from the more than 500 atmospheric nuclear tests conducted between 1945 and 1962 by the United States, the former Soviet Union, England, France and China. (author)

  10. 30 years of the Goiania Accident: a comparative study with other radioactivity dispersion events

    International Nuclear Information System (INIS)

    Smith, Ricardo Bastos; Vicente, Roberto

    2017-01-01

    The year 2017 marks 30 years since the radioactive accident that occurred in the city of Goiania, capital of the state of Goias. It was the largest radiological accident in Brazil, and one of the largest in the world occurring outside nuclear facilities. Regarding the accidents at nuclear power plants, two of the biggest were Chernobyl in Ukraine, a year and a half before Goiania, and the Fukushima accident in Japan, in 2011. Different amounts of radioactive material were dispersed in the environment in each of these events. However, each one’s main pathway of dispersion was different: the accident of Goiania was terrestrial, Chernobyl was at the atmosphere, and Fukushima was mainly in the ocean. This work aims to study these different amounts, comparing such activities. In addition, it proposes to compare the sea dispersion of Fukushima with the amount of radioactive waste dumped in the oceans, when the release of radioactive waste at sea was permitted. It also proposes to compare the Chernobyl aerial dispersion with the radioactive material dissipated in the atmosphere, resulting from the more than 500 atmospheric nuclear tests conducted between 1945 and 1962 by the United States, the former Soviet Union, England, France and China. (author)

  11. Radiodosimetry and preventive measures in the event of a nuclear accident. Proceedings of an international symposium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    An international symposium on Radiodosimetry and Preventive Measures in the Event of a Nuclear Accident was held in Cracow, Poland, from 26 to 28 May 1994. The symposium was organized by the Polish Society for Nuclear Medicine, and co-sponsored by the IAEA. Over 40 experts from Belarus, Latvia, Lithuania, Germany, Poland, the Russian Federation, Sweden and Switzerland participated. The aim of the Symposium was to review models of iodine kinetics used in the calculation of internal radiation doses to the thyroid after the Chernobyl accident, to discuss internal and external radiation dose to the thyroid in terms or risk of thyroid cancer, and to present data on the incidence rate of thyroid cancer in the selected iodine deficient area in Poland. A part of the symposium was dedicated to the physiological basis of iodine prophylaxis and emergency planning for a nuclear accident. Recommendations of the IAEA on preventive measures in the event of a nuclear accident were also addressed. These proceedings contain the full text of the eight invited papers presented at the symposium. Refs, figs, tabs.

  12. Radiodosimetry and preventive measures in the event of a nuclear accident. Proceedings of an international symposium

    International Nuclear Information System (INIS)

    1996-08-01

    An international symposium on Radiodosimetry and Preventive Measures in the Event of a Nuclear Accident was held in Cracow, Poland, from 26 to 28 May 1994. The symposium was organized by the Polish Society for Nuclear Medicine, and co-sponsored by the IAEA. Over 40 experts from Belarus, Latvia, Lithuania, Germany, Poland, the Russian Federation, Sweden and Switzerland participated. The aim of the Symposium was to review models of iodine kinetics used in the calculation of internal radiation doses to the thyroid after the Chernobyl accident, to discuss internal and external radiation dose to the thyroid in terms or risk of thyroid cancer, and to present data on the incidence rate of thyroid cancer in the selected iodine deficient area in Poland. A part of the symposium was dedicated to the physiological basis of iodine prophylaxis and emergency planning for a nuclear accident. Recommendations of the IAEA on preventive measures in the event of a nuclear accident were also addressed. These proceedings contain the full text of the eight invited papers presented at the symposium. Refs, figs, tabs

  13. Development of technique for estimating primary cooling system break diameter in predicting nuclear emergency event sequence

    International Nuclear Information System (INIS)

    Tatebe, Yasumasa; Yoshida, Yoshitaka

    2012-01-01

    If an emergency event occurs in a nuclear power plant, appropriate action is selected and taken in accordance with the plant status, which changes from time to time, in order to prevent escalation and mitigate the event consequences. It is thus important to predict the event sequence and identify the plant behavior resulting from the action taken. In predicting the event sequence during a loss-of-coolant accident (LOCA), it is necessary to estimate break diameter. The conventional method for this estimation is time-consuming, since it involves multiple sensitivity analyses to determine the break diameter that is consistent with the plant behavior. To speed up the process of predicting the nuclear emergency event sequence, a new break diameter estimation technique that is applicable to pressurized water reactors was developed in this study. This technique enables the estimation of break diameter using the plant data sent from the safety parameter display system (SPDS), with focus on the depressurization rate in the reactor cooling system (RCS) during LOCA. The results of LOCA analysis, performed by varying the break diameter using the MAAP4 and RELAP5/MOD3.2 codes, confirmed that the RCS depressurization rate could be expressed by the log linear function of break diameter, except in the case of a small leak, in which RCS depressurization is affected by the coolant charging system and the high-pressure injection system. A correlation equation for break diameter estimation was developed from this function and tested for accuracy. Testing verified that the correlation equation could estimate break diameter accurately within an error of approximately 16%, even if the leak increases gradually, changing the plant status. (author)

  14. Accident sequences evaluation using SFATs for low power and shutdown operation of pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Kim, Chansoo; Chung, Chang-Hyun; Yang, Huichang

    2004-01-01

    To maintain the level of defense-in-depth safety of Pressurized Heavy Water Reactor (PHWR) during LP/SD operation, the qualitative risk evaluation methods such as Safety Function Assessment Trees (SFATs) are required. Therefore SFATs are suggested to assess and manage the PHWR safety in LP/SD. Before this study, safety functions of PHWR were classified into 7 groups; Reactivity Control, Core Cooling, Secondary Heat Removal, Primary Heat Transport Inventory, Essential Electrical Power, Cooling Water, and Containment Integrity. The SFATs for PHWR LP/SD operations were developed along with the Plant Outage Status (POS) variation, and totally 38 SFATs were developed for Wolsung Unit 2. For the verification of SFATs logics developed, top 5 accident sequences those contribute the CDF of PHWR were selected, and plant safety status were evaluated for those accident sequences. Accident sequences such as DCC-4 (Dual Control Computer Failure), CL4-16 (Total Loss of Class IV Power), and FWPV-11 (Loss of Feedwater Supply to SG due to Failure of Pumps/Values) were included. In this research the evaluation of plant safety status by accident sequences using SFATs and the verification of the SFATs were performed. Through the verification of SFAT logics, the enhancements to the internal logics of the SFATs were made, and the dependencies between safety systems and support systems were considered. It is expected the defense-in-depth evaluation model of PHW just as SFATs can be utilized in the configuration risk management program (CRMP) development and improve technical specifications development for Korean PHWRs. (author)

  15. Containment response to a severe accident (TMLB sequence) with and without mitigation strategies

    International Nuclear Information System (INIS)

    Passalacqua, R.

    2004-01-01

    A loss of SG feed-water (TMLB sequence) for a prototypic PWR 900 MWe with a multi-compartment configuration (with 11 and 16 cells nodalization) has been calculated by the author using the ASTEC code in the frame of the EVITA project (5th Framework Programme, FWP). A variety of hypothesis (e.g. activation of sprays and hydrogen recombiners) and possible consequences of these assumptions (cavity flooding, hydrogen combustion, etc.) have been made in order to evaluate the global reactor containment building response (pressure, aerosol/FP concentration, etc.). The need to dispose of severe accident management guidelines (SAMGs) is increasing. These guidelines are meant for nuclear plants' operators in order to allow them to apply mitigation strategies all along a severe accident, which, only in its initial phase, may last several days. The purpose of this paper is to outline the influence on the containment load of most common accident occurrences and operators actions, which is essential in establishing SAMGs. ASTEC (Accident Source Term Evaluation Code) is a computer code for the evaluation of the consequences of a postulated nuclear plant severe accident sequence. ASTEC is a computer tool currently under joint development by the Institut de Radioprotection et de Surete Nucleaire (IRSN), France, and Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS), Germany. The aim of the development is to create a fast running integral code package, reliable in all simulations of a severe accident, to be used for level-2 PSA analysis. It must be said that several recent developments have significantly improved the best-estimate models of ASTEC and a new version (ASTEC V1.0) has been released mid-2002. Nevertheless, the somehow obsolete ASTECv0.3 version here used, has given results very useful for the estimation of the global risk of a nuclear plant. Moreover, under the current 6th FWP (Sustainable Integration of EU Research on Severe Accident Phenomenology and Management), the

  16. On-site habitability in the event of an accident at a nuclear facility

    International Nuclear Information System (INIS)

    1989-01-01

    This publication is intended to provide technical guidance and a methodology for regulatory bodies, designers, constructors and operators of nuclear facilities to assist them in assessing the current situation as regards on-site habitability for their specific nuclear facilities. Initially, the aim will be to ensure that the ''vital areas'' of the facility which are necessary for the safe operation and shutdown of the facility will remain habitable, in some cases continuously and in others transiently, in the event of an accident inside or outside the installation. The assessment procedure can be used not only for potential radiation accidents but also to consider the effects on habitability of those probable non-radiological events which, if not correctly and effectively countered, could lead to the development of potentially unsafe conditions in the facility itself. 30 refs, 4 figs, 8 tabs

  17. Three Mile Island accident: a case study of life event appraisal

    International Nuclear Information System (INIS)

    Goldsteen, R.L.

    1983-01-01

    This research investigates community reactions to the accident at the Three Mile Island (TMI) nuclear powered electric generating plant in March, 1979. The investigation is placed in the context of life event research and chooses an appraisal orientation. Three innovations are argued: 1) perceived consequences of the event best predict reactions to it, 2) the attitudes of significant others toward the event influence reactions to the accident under certain circumstances, and 3) sense of well-being is a good outcome measure for a general population. The hypotheses posit that the attitudes of others will affect sense of well-being only when individual attitudes concerning the consequences of the accident are moderate; when individual attitudes are extreme, the attitudes of others will have no demonstrable effect on outcomes. The findings did not support all the prediction of the hypotheses. However, they indicate that perceived consequences are the best predictors of sense of well-being and that an individual's attitudinal position, his strength of attitude, and the nature of the stimulus are highly related to whether or not an individual will be influenced by the views of others

  18. Molecular Characterization of Transgenic Events Using Next Generation Sequencing Approach.

    Science.gov (United States)

    Guttikonda, Satish K; Marri, Pradeep; Mammadov, Jafar; Ye, Liang; Soe, Khaing; Richey, Kimberly; Cruse, James; Zhuang, Meibao; Gao, Zhifang; Evans, Clive; Rounsley, Steve; Kumpatla, Siva P

    2016-01-01

    Demand for the commercial use of genetically modified (GM) crops has been increasing in light of the projected growth of world population to nine billion by 2050. A prerequisite of paramount importance for regulatory submissions is the rigorous safety assessment of GM crops. One of the components of safety assessment is molecular characterization at DNA level which helps to determine the copy number, integrity and stability of a transgene; characterize the integration site within a host genome; and confirm the absence of vector DNA. Historically, molecular characterization has been carried out using Southern blot analysis coupled with Sanger sequencing. While this is a robust approach to characterize the transgenic crops, it is both time- and resource-consuming. The emergence of next-generation sequencing (NGS) technologies has provided highly sensitive and cost- and labor-effective alternative for molecular characterization compared to traditional Southern blot analysis. Herein, we have demonstrated the successful application of both whole genome sequencing and target capture sequencing approaches for the characterization of single and stacked transgenic events and compared the results and inferences with traditional method with respect to key criteria required for regulatory submissions.

  19. Molecular Characterization of Transgenic Events Using Next Generation Sequencing Approach.

    Directory of Open Access Journals (Sweden)

    Satish K Guttikonda

    Full Text Available Demand for the commercial use of genetically modified (GM crops has been increasing in light of the projected growth of world population to nine billion by 2050. A prerequisite of paramount importance for regulatory submissions is the rigorous safety assessment of GM crops. One of the components of safety assessment is molecular characterization at DNA level which helps to determine the copy number, integrity and stability of a transgene; characterize the integration site within a host genome; and confirm the absence of vector DNA. Historically, molecular characterization has been carried out using Southern blot analysis coupled with Sanger sequencing. While this is a robust approach to characterize the transgenic crops, it is both time- and resource-consuming. The emergence of next-generation sequencing (NGS technologies has provided highly sensitive and cost- and labor-effective alternative for molecular characterization compared to traditional Southern blot analysis. Herein, we have demonstrated the successful application of both whole genome sequencing and target capture sequencing approaches for the characterization of single and stacked transgenic events and compared the results and inferences with traditional method with respect to key criteria required for regulatory submissions.

  20. Preliminary Analysis of Aircraft Loss of Control Accidents: Worst Case Precursor Combinations and Temporal Sequencing

    Science.gov (United States)

    Belcastro, Christine M.; Groff, Loren; Newman, Richard L.; Foster, John V.; Crider, Dennis H.; Klyde, David H.; Huston, A. McCall

    2014-01-01

    Aircraft loss of control (LOC) is a leading cause of fatal accidents across all transport airplane and operational classes, and can result from a wide spectrum of hazards, often occurring in combination. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of conditions and uncertainties, including multiple hazards, and their validation must provide a means of assessing system effectiveness and coverage of these hazards. This requires the definition of a comprehensive set of LOC test scenarios based on accident and incident data as well as future risks. This paper defines a comprehensive set of accidents and incidents over a recent 15 year period, and presents preliminary analysis results to identify worst-case combinations of causal and contributing factors (i.e., accident precursors) and how they sequence in time. Such analyses can provide insight in developing effective solutions for LOC, and form the basis for developing test scenarios that can be used in evaluating them. Preliminary findings based on the results of this paper indicate that system failures or malfunctions, crew actions or inactions, vehicle impairment conditions, and vehicle upsets contributed the most to accidents and fatalities, followed by inclement weather or atmospheric disturbances and poor visibility. Follow-on research will include finalizing the analysis through a team consensus process, defining future risks, and developing a comprehensive set of test scenarios with correlation to the accidents, incidents, and future risks. Since enhanced engineering simulations are required for batch and piloted evaluations under realistic LOC precursor conditions, these test scenarios can also serve as a high-level requirement for defining the engineering simulation enhancements needed for generating them.

  1. Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

    Energy Technology Data Exchange (ETDEWEB)

    Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

  2. Policy elements for post-accident management in the event of nuclear accident. Document drawn up by the Steering Committee for the Management of the Post-Accident Phase of a Nuclear Accident (CODIRPA). Final version - 5 October 2012

    International Nuclear Information System (INIS)

    2012-01-01

    Pursuant to the Inter-ministerial Directive on the Action of the Public Authorities, dated 7 April 2005, in the face of an event triggering a radiological emergency, the National directorate on nuclear safety and radiation protection (DGSNR), which became the Nuclear safety authority (ASN) in 2006, was tasked with working the relevant Ministerial offices in order to set out the framework and outline, prepare and implement the provisions needed to address post-accident situations arising from a nuclear accident. In June 2005, the ASN set up a Steering committee for the management of the post-accident phase in the event of nuclear accident or a radiological emergency situation (CODIRPA), put in charge of drafting the related policy elements. To carry out its work, CODIRPA set up a number of thematic working groups from 2005 on, involving in total several hundred experts from different backgrounds (local information commissions, associations, elected officials, health agencies, expertise agencies, authorities, etc.). The working groups reports have been published by the ASN. Experiments on the policy elements under construction were carried out at the local level in 2010 across three nuclear sites and several of the neighbouring municipalities, as well as during national crisis drills conducted since 2008. These works gave rise to two international conferences organised by ASN in 2007 and 2011. The policy elements prepared by CODIRPA were drafted in regard to nuclear accidents of medium scale causing short-term radioactive release (less than 24 hours) that might occur at French nuclear facilities equipped with a special intervention plan (PPI). They also apply to actions to be carried out in the event of accidents during the transport of radioactive materials. Following definitions of each stage of a nuclear accident, this document lists the principles selected by CODIRPA to support management efforts subsequent to a nuclear accident. Then, it presents the main

  3. Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Valle Cepero, Reinaldo

    2004-01-01

    The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

  4. Nuclear accidents at the Fukushima Dai-ichi power plant. History, events and consequences

    International Nuclear Information System (INIS)

    Berniolles, Jean Marc

    2011-01-01

    Written few weeks after the accident, this article first recalls the circumstances (earthquake and tsunami), and then describes the accidental process within the primary vessels of the Fukushima Dai-ichi number 1, 2 and 3 reactors. The author then describes the interventions which aimed at cooling these three reactors, the problem faced for the storage of used fuels, and then the sequence of accidents: loss of cooling means leading to an explosion, problems faced in the different storage pools. He describes the various steps of recovery (primary cooling, electricity supply), discusses the consequences in terms of radioactivity releases in the plant environment with a comparison with Chernobyl, and also in terms of nature and quantity of radioactive elements. He comments radioactivity controls and measurements, evacuation measures, measurements performed by the IAEA, measurements of sea radioactivity, and the establishment of maps of ground radioactivity around the plant. He discusses the perspectives associated with these measurements for the surroundings of the Fukushima site

  5. Radioactive Reversal? The Fukushima Accident as a Focusing Event for Comparative Policy Change on Nuclear Energy

    Science.gov (United States)

    Sanchez, Victoria Justine

    This dissertation project examines the 2011 Fukushima nuclear accident as a focusing event for policy change on nuclear energy. For example, following the accident, Germany (and much of Europe) experienced a reversal of policy on nuclear energy. Conversely, many others such as China, Russia, and France, did not exhibit such a retraction against nuclear power, albeit with public debate about the risks and consequences of accidents. Why has there been dramatic policy change in some cases but not others? The political and literal fallout of Fukushima has provoked a wave of policy change towards nuclear energy at the national level. Through qualitative and quantitative measures, we can view Fukushima as an impetus for comparing the dynamics of nuclear policy change. Quantitatively, this project employs logistic regression to explore variables such as regime type, energy security, trade supply and demand, climate change concerns, and public acceptance are related to policy outcomes and change on nuclear energy in the post-Fukushima context of 49 different countries. Qualitatively, country cases (Russia, Germany, and Canada) are assessed into three categories based on the outcome of policy decisions on nuclear energy following Fukushima for a richer analysis. Beyond the Fukushima example, we can hope to better understand how political focusing events can gain influence in an international context.

  6. Method for improving accident sequence recognition in nuclear power plant control rooms

    International Nuclear Information System (INIS)

    Heising, C.D.; Dinsmore, S.C.

    1983-01-01

    This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to construct and quantitatively evaluate an alarm analysis system for the engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide system failure mode information which can be used to construct alarm trees. These alarm trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel. In the alarm tree, possible states of each instrumented alarem are identified as true or false. In addition, a warning status is defined and integrated into an alarm analysis routine. The impact of this additional status conditioned on the Boolean laws used to evaluate the alarm trees is examined. An application is described for BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents

  7. International policy on intervention in the event of a nuclear accident

    International Nuclear Information System (INIS)

    Jensen, P.H.; Crick, M.J.; Gonzalez, A.J.

    1996-01-01

    Criteria for taking particular protective actions with the aim of preventing or reducing radiation exposures to the population or to workers in the event of a nuclear accident or radiological emergency can be established on the basis of radiological protection principles for intervention situations. It is of utmost importance that pre-established intervention levels for different protective measures form an integral part of an emergency response plan. Generic optimized intervention levels and their derived operational quantities based on the principles given in this paper are judged to provide protection that would be justified and reasonable optimized for a wide range of accident situations although they can only be used as guidelines. Any specific optimization would lead to intervention levels that might be either higher or lower than those emerging from a generic optimization. (author). 9 refs

  8. Initiating events of accidents in the practice of oil well logging in Cuba

    International Nuclear Information System (INIS)

    Alles Leal, A.; Perez Reyes, Y.; Dumenigo Gonzalez, C.

    2013-01-01

    The oil well logging is an extremely important activity within the oil industry, but in turn, brings risks that occasionally result in damage to health, the environment and economic losses. In this context, risk analysis has become an important tool to control them through their prediction and the study of the factors that determine them, enabling substantiated decisions to, first, foresee accidents and, secondly, to minimize their consequences. This paper proposes the elaboration of a list of initiating events of accidents in the practice of oil well logging which is one of the most important aspects for further evaluation of radiation safety of this practice. For its determination the technique employed to identify risks was 'Failure Modes and Effects Analysis (FMEA)' by applying it to the different stages and processes of practice. (Author)

  9. International policy on intervention in the event of a nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, P H [Risoe National Lab., Roskilde (Denmark); Crick, M J; Gonzalez, A J [International Atomic Energy Agency, Vienna (Austria)

    1996-08-01

    Criteria for taking particular protective actions with the aim of preventing or reducing radiation exposures to the population or to workers in the event of a nuclear accident or radiological emergency can be established on the basis of radiological protection principles for intervention situations. It is of utmost importance that pre-established intervention levels for different protective measures form an integral part of an emergency response plan. Generic optimized intervention levels and their derived operational quantities based on the principles given in this paper are judged to provide protection that would be justified and reasonable optimized for a wide range of accident situations although they can only be used as guidelines. Any specific optimization would lead to intervention levels that might be either higher or lower than those emerging from a generic optimization. (author). 9 refs.

  10. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

    This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

  11. A Preliminary Neutral Framework for the Accident Sequence Evaluation for a Hydrogen Conversion Reactor

    International Nuclear Information System (INIS)

    Han, Seok Jung; Yang, Joon Eon

    2005-01-01

    A framework for an early stage PSA for a hydrogen conversion reactor has been proposed in this paper. The approach is based on a functional and top-down approach. A main concerning point of this approach is to use a design neutral framework. A design neutral framework of PSA can provide a flexibility to apply to several candidate design concepts or options. This neutral-framework idea was borrowed from a proposed regulatory framework in US NRC. The feasibility of our proposed approach has been assessed to be applied in an accident sequence analysis for a hydrogen conversion reactor

  12. Current status and issues of external event PSA for extreme natural hazards after Fukushima accident

    International Nuclear Information System (INIS)

    Choi, In-Kil; Hahm, Daegi; Kim, Min Kyu

    2014-01-01

    Extreme external events is emerged as significant risk contributor to the nuclear power plants after Fukushima Daiichi accident due to the catastrophic earthquake followed by great tsunami greater than a design basis. This accident shows that the extreme external events have the potential to simultaneously affect redundant and diverse safety systems and thereby induce common cause failure or common cause initiators. The probabilistic risk assessment methodology has been used for the risk assessment and safety improvement against the extreme natural hazards. The earthquake and tsunami hazard is an important issue for the nuclear industry in Korea. In this paper, the role and application of probabilistic safety assessment for the post Fukushima action will be introduced. For the evaluation of the extreme natural hazard, probabilistic seismic and tsunami hazard analysis is being performed for the safety enhancement. The research activity on the external event PSA and its interim results will be introduced with the issues to be solved in the future for the reliability enhancement of the risk analysis results. (authors)

  13. Technical organization of safety authorities for the event of an accident at a nuclear installation

    International Nuclear Information System (INIS)

    Scherrer, J.; Evrard, J.M.; Ney, J.

    1986-01-01

    Within the general context of nuclear safety, the Central Nuclear Installation Safety Service of the French Ministry for Industry and its technical backup, the Institute for Radiation Protection and Nuclear Safety of the CEA (Atomic Energy Commission), have established a special organization designed to provide real-time forecasts of the evolution of a nuclear accident situation with sufficient forewarning for the local representative of the Government (the Commissaire de la Republique in the Departement affected) to implement, as required, effective countermeasures to protect the population - for example, confinement indoors or evacuation. Descriptions are given of the principles of this organization and the particular precautions taken to confront the problems of mobilizing experts and of dealing with the saturation of normal telecommunications channels to be expected in the event of a nuclear accident. The organization set up for the installations belonging to Electricite de France is given as a detailed example. Particular stress is placed on the organizational arrangements of the Institute for Radiation Protection and Nuclear Safety designed to provide the emergency teams with the evaluation and forecasting tools they require to carry out their tasks. The procedures are on the whole well developed for atmospheric radioactivity transport, for which operational models already exist. Computer-backed methods with improved performance are at present being developed. A method of forecasting the behaviour of the releases resulting from nuclear accidents is set out for pressurized water reactors, based on evaluating the physical state of the installation, confinement integrity, availability of safety and backup systems, support systems and feed sources and on forecasting how this state will develop on the basis of measured and inferred physical values transmitted from the affected power station through a national network. The experience acquired during accident

  14. Organizational forms of medical care in the event of radiation accidents in the German Democratic Republic

    International Nuclear Information System (INIS)

    Nack, P.; Arndt, D.; Schuettmann, W.

    1977-01-01

    Medical care of radiation casualties in the German Democratic Republic (GDR) is organized on two levels. On the level of users the responsible Medical Officers guarantee both the routine control of persons occupationally exposed to radiation and first aid in the event of accidents. On the second level medical treatment is given either in the Clinical Department of the National Board of Nuclear Safety and Radiation Protection or in specialized national health system clinics having facilities for intensive medical care. A decision on hospitalization is made according to the conditions of the accident and the necessary diagnostic and therapeutic measures as a rule are based on consultations between the responsible Medical Officer and the departments of the Board (Emergency Assistance Service, Clinical Department, Consultative Committee). For serious cases where haematological complications can be expected, a central medical clinic with facilities for bone-marrow transplants is available. The casualties are treated in local clinics which are provided with continuous support and advice by the Board. This support consists in: (i) immediate activity by a consultative committee of the Board's physicians and scientists experienced and trained in radiation protection and the treatment of radiation accidents; (ii) the requirement of compulsory examination methods and take-over of specialized laboratory investigations; and (iii) the use of a mobile emergency measuring system in cases of additional incorporation. It is the main principle of medical care in case of radiation accidents to consult, as early as possible, a medical consultative committee of the Board in the field of radiation protection at each step of medical care. (author)

  15. Development of accident sequence precursors methodologies for core damage Probabilities in NPPs

    International Nuclear Information System (INIS)

    Munoz, R.; Minguez, E.; Melendez, E.; Sanchez-Perea, M.; Izquierdo, J.M.

    1998-01-01

    Several licensing programs have focused on the evaluation of the importance of operating events occurred in NPPs. Some have worked the dynamic aspects of the sequence of events involved, reproducing the incidents, while others are based on PSA applications to incident analysis. A method that controls the two above approaches to determine risk analysis derives from the Integrated Safety Assessment methodology (ISA). The dynamics of the event is followed by transient simulation in tree form, building a Setpoint or Deterministic Dynamic Event Tree (DDET). When a setpoint is reached, the actuation of a protection is triggered, then the tree is opened in branches corresponding to different functioning states. The engineering simulator with the new states followers each branch. One of these states is the nominal one, which is the PSA is associated to the success criterion of the system. The probability of the sequence is calculated in parallel to the dynamics. The following tools should perform the couple simulation: 1. A Scheduler that drives the simulation of the different sequences, and open branches upon demand. It will be the unique generator of processes while constructing the tree calculation, and will develop the computation in a distributed computational environment. 2. The Plant Simulator, which models the plant systems and the operator actions throughout a sequence. It receives the state of the equipment in each sequence and must provide information about setpoint crossing to the Scheduler. It will receive decision flags to continue or to stop each sequence, and to send new conditions to other plant simulators. 3. The Probability Calculator, linked only to the Scheduler, is the fault trees associated with each event tree header and performing their Boolean product. (Author)

  16. Radiological Emergency Preparedness after the Early Phase of an Accident : Focusing on an Air Contamination Event

    International Nuclear Information System (INIS)

    Jeong, Hyo Joon; Hwang, Won Tae; Kim, Eun Han; Han, Moon Hee

    2010-01-01

    Toxic materials in an urban area can be caused by a variety of events, such as accidental releases on industrial complexes, accidents during the transportation of hazardous materials and intentional explosions. Most governments around the world and their citizens have become increasingly worried about intentional accidents in urban area after the 911 terrorist attack in the United States of America. Even though there have been only a few attempted uses of Radiological Dispersal Devices (RDDs), accidental releases have occurred many times at commercial nuclear power plants and nuclear waste disposal sites. When an intentional release of radioactive materials occurs in an urban area, air quality for radioactive materials in the environment is of great importance to take action for countermeasures and environmental risk assessments. Atmospheric modeling is part of the decision making tasks and that it is particularly important for emergency managers as they often need to take actions quickly on very inadequate information(1). A simple model such as HOTSPOT required wind direction and source term would be enough to support the decision making in the early phase of an accident, but more sophisticated atmospheric modeling is required to adjust decontamination area and relocation etc after the early phase of an accidental event. In this study, we assume an explosion of 137 Cs using RDDs in the metropolitan area of Soul, South Korea. California Puff Model (CALPUFF) is used to calculate an atmospheric dispersion and transport for 137 Cs. Atmospheric dispersion and quantitative radiological risk analysis for 137 Cs were performed assuming an intentional explosion in the metropolitan area of Soul, South Korea after the early phase of emergency. These kinds of atmospheric modeling and risk analysis could provide a means for decision makers to take action on important issues such as the cleanup of the contaminated area and countermeasures to protect the public caused by

  17. A PC Mathcad-based computational aid for severe accident analysis and its application to a BWR small LOCA sequence

    International Nuclear Information System (INIS)

    Wu, Laung-Kuang T.; Lee, S.J.

    2004-01-01

    A PC-based Mathcad program is used to develop a computational aid for analyzing severe accident phenomena. This computational aid uses simple engineering expressions and empirical correlations to estimate key quantities and timings at various stages of accident progressions. In this paper, the computational aid is applied to analyze an early phase of a BWR small LOCA sequence. The accident phenomena analyzed include: break flow rates, boiled-up water level in the core, core uncovery time, depressurization of the reactor pressure vessel, core heat-up, onset of clad oxidation, hydrogen generation, and onset of fuel relocation. The results are compared with those obtained running the MAAP 3.0B code. This PC-based computational aid can be used to train plant personnel in understanding severe accident phenomena and to assist them in managing severe accidents. (author)

  18. 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)

  19. Station blackout transient at the Browns Ferry Unit 1 Plant: a severe accident sequence analysis (SASA) program study

    International Nuclear Information System (INIS)

    Schultz, R.R.

    1982-01-01

    Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators

  20. CNE (Embalse nuclear power plant): probabilistic safety study. Loss of service water. Probabilistic evaluation and analysis through events sequence

    International Nuclear Information System (INIS)

    Couto, A.J.; Perez, S.S.

    1987-01-01

    This work is part of a study on the service water systems of the Embalse nuclear power plant from a safety point of view. The faults of service water systems of high and low pressure that can lead to situations threatening the plant safety were analyzed in a previous report. The event 'total loss of low pressure service water' causes the largest number of such conditions. Such event is an operational incident that can lead to an accident situation due to faults in the required process systems or by omission of a procedure. The annual frequency of the event 'total loss of low pressure service water' is calculated. The main contribution comes from pump failure. The evaluation of the accident sequences shows that the most direct way to the liberation of fission products is the loss of steam generators as heat sink. The contributions to small and large LOCA and electric supply loss are analyzed. The sequence that leads to tritium release through boiling of moderator is also evaluated. (Author)

  1. Versatile single-chip event sequencer for atomic physics experiments

    Science.gov (United States)

    Eyler, Edward

    2010-03-01

    A very inexpensive dsPIC microcontroller with internal 32-bit counters is used to produce a flexible timing signal generator with up to 16 TTL-compatible digital outputs, with a time resolution and accuracy of 50 ns. This time resolution is easily sufficient for event sequencing in typical experiments involving cold atoms or laser spectroscopy. This single-chip device is capable of triggered operation and can also function as a sweeping delay generator. With one additional chip it can also concurrently produce accurately timed analog ramps, and another one-chip addition allows real-time control from an external computer. Compared to an FPGA-based digital pattern generator, this design is slower but simpler and more flexible, and it can be reprogrammed using ordinary `C' code without special knowledge. I will also describe the use of the same microcontroller with additional hardware to implement a digital lock-in amplifier and PID controller for laser locking, including a simple graphics-based control unit. This work is supported in part by the NSF.

  2. The fuzzy set theory application to the analysis of accident progression event trees with phenomenological uncertainty issues

    International Nuclear Information System (INIS)

    Chun, Moon-Hyun; Ahn, Kwang-Il

    1991-01-01

    Fuzzy set theory provides a formal framework for dealing with the imprecision and vagueness inherent in the expert judgement, and therefore it can be used for more effective analysis of accident progression of PRA where experts opinion is a major means for quantifying some event probabilities and uncertainties. In this paper, an example application of the fuzzy set theory is first made to a simple portion of a given accident progression event tree with typical qualitative fuzzy input data, and thereby computational algorithms suitable for application of the fuzzy set theory to the accident progression event tree analysis are identified and illustrated with example applications. Then the procedure used in the simple example is extended to extremely complex accident progression event trees with a number of phenomenological uncertainty issues, i.e., a typical plant damage state 'SEC' of the Zion Nuclear Power Plant risk assessment. The results show that the fuzzy averages of the fuzzy outcomes are very close to the mean values obtained by current methods. The main purpose of this paper is to provide a formal procedure for application of the fuzzy set theory to accident progression event trees with imprecise and qualitative branch probabilities and/or with a number of phenomenological uncertainty issues. (author)

  3. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  4. Ultimate Electrical Means for Severe Accident and Multi Unit Event Management

    International Nuclear Information System (INIS)

    Guisez, Xavier

    2015-01-01

    Following the Multi Unit Severe Accident that occurred at Fukushima as a result of the tsunami on 11 March 2011, the European Council decided to submit its Nuclear Power Plants to a Stress Test. In Belgium, this Stress Test, named BEST (Belgian Stress Test), was successfully concluded at the end of 2011. Nevertheless, Electrabel decided, in agreement with the Authorities, to start a beyond design basis action plan, with the goal to mitigate the consequences of a Beyond Design Basis Accident and a Multi Unit Event. Consequently, this has led to an improvement of the robustness of its Nuclear Power Plants. Considering the importance of electrical power supply to a nuclear power plant, a significant part of this action plan consisted of setting up a mobile, 'plug and play' method for the electrical power supply to some major safety systems. In order to install this ultimate power supply, three factors were retained as essential. First, important reactor monitoring instrumentation is preserved. Second, core cooling is provided at all times. Finally, systems are easily made operational within a very short delay of time. During normal operation and Design Basis Events, core cooling is provided by High Voltage equipment. However, during high stress circumstances, it is too complex to realize connections on this equipment. Therefore, analysis was performed to realize core cooling with, easier to handle, Low Voltage equipment. These systems are powered by several GenSets, especially designed and manufactured for this application. The outcome of this project are easy to use, ultimate means, that supply electric power to important safety systems in order to drastically reduce the risk of core damage, during a beyond design basis event. Additionally, for all ultimate means, procedures and training modules were developed for the operators. (authors)

  5. The significance of water hammer events to public dose from reactor accidents: A probabilistic assessment

    International Nuclear Information System (INIS)

    Amico, P.J.; Ferrell, W.L.; Rubin, M.P.

    1984-01-01

    A probabilistic assessment was made of the effects on public dose of water hammer events in LWRs. The analysis utilized actual historical water hammer data to determine if the water hammer events contributed either to system failure rates or initiating event frequencies. Representative PRAs were used to see if changes in initiating events and/or system failures caused by water hammer resulted in new values for the dominant sequences in the PRAs. New core melt frequencies were determined and carried out to the subsequent increase in public dose. It is concluded that water hammer is not a significant problem with respect to risk to the public for either BWRs or PWRs. (orig./HP)

  6. 'It was a freak accident': an analysis of the labelling of injury events in the US press.

    Science.gov (United States)

    Smith, Katherine C; Girasek, Deborah C; Baker, Susan P; Manganello, Jennifer A; Bowman, Stephen M; Samuels, Alicia; Gielen, Andrea C

    2012-02-01

    Given that the news media shape our understanding of health issues, a study was undertaken to examine the use by the US media of the expression 'freak accident' in relation to injury events. This analysis is intended to contribute to the ongoing consideration of lay conceptualisation of injuries as 'accidents'. LexisNexis Academic was used to search three purposively selected US news sources (Associated Press, New York Times and Philadelphia Inquirer) for the expression 'freak accident' over 5 years (2005-9). Textual analysis included both structured and open coding. Coding included measures for who used the expression within the story, the nature of the injury event and the injured person(s) being reported upon, incorporation of prevention information within the story and finally a phenomenological consideration of the uses and meanings of the expression within the story context. Results The search yielded a dataset of 250 human injury stories incorporating the term 'freak accident'. Injuries sustained by professional athletes dominated coverage (61%). Fewer than 10% of stories provided a clear and explicit injury prevention message. Stories in which journalists employed the expression 'freak accident' were less likely to include prevention information than stories in which the expression was used by people quoted in the story. Journalists who frame injury events as freak accidents may be an appropriate focus for advocacy efforts. Effective prevention messages should be developed and disseminated to accompany injury reporting in order to educate and protect the public.

  7. Action to be taken in the event of a radiological accident

    International Nuclear Information System (INIS)

    Bresson, G.; Nenot, J.C.

    1977-01-01

    In the event of a radiological accident affecting people, the measures that have to be taken are the responsibility of a large number of persons whose original disciplines differ widely. In the interest of efficiency, it is obviously essential that these measures should be co-ordinated; this implies smoothly functioning liaison between the persons responsible for action at different levels. These levels of action are numerous and differ very considerably; they include, in the first place, the links between the nuclear facility and the medical authorities, either directly within the hospital framework or through the intermediary of an industrial medicine service; then the links between the hospital sector and the large number of experts concerned with the highly specialized aspects of the diagnostic, therapeutic and prognostic problems of irradiation or contamination by radioisotopes; lastly, the links between these various specialists. In view of the wide variety of the parameters involved in accidents, the organization of the action to be taken cannot be encompassed within a rigid framework, especially as it should be possible to apply this organization at both the national and the international level, taking into account the diversity of the medico-legal aspects. The efficiency of the means applied is therefore governed by the flexibility of the procedure; however, the relative scarcity of accidents, i.e. the absence of any involvement of persons and equipment on a true scale, makes it imperative that a high degree of precision be applied in preparing emergency plans, since the omission of one step or one link may have serious or irreparable consequences which cannot always be offset by improvization. The outline of such an operational organization is presented and discussed in the light of past experience. (author)

  8. Application of a Software tool for Evaluating Human Factors in Accident Sequences

    International Nuclear Information System (INIS)

    Queral, Cesar; Exposito, Antonio; Gonzalez, Isaac; Quiroga, Juan Antonio; Ibarra, Aitor; Hortal, Javier; Hulsund, John-Einar; Nilsen, Svein

    2006-01-01

    The Probabilistic Safety Assessment (PSA) includes the actions of the operator like elements in the set of the considered protection performances during accident sequences. Nevertheless, its impact throughout a sequence is not analyzed in a dynamic way. In this sense, it is convenient to make more detailed studies about its importance in the dynamics of the sequences, letting make studies of sensitivity respect to the human reliability and the response times. For this reason, the CSN is involved in several activities oriented to develop a new safety analysis methodology, the Integrated Safety Assessment (ISA), which must be able to incorporate operator actions in conventional thermo-hydraulic (TH) simulations. One of them is the collaboration project between CSN, HRP and the DSE-UPM that started in 2003. In the framework of this project, a software tool has been developed to incorporate operator actions in TH simulations. As a part of the ISA, this tool permits to quantify human error probabilities (HEP) and to evaluate its impact in the final state of the plant. Independently, it can be used for evaluating the impact of the execution by operators of procedures and guidelines in the final state of the plant and the evaluation of the allowable response times for the manual actions of the operator. The results obtained in the first pilot case are included in this paper. (authors)

  9. Principles for establishing intervention levels for the protection of the public in the event of a nuclear accident or radiological emergency

    International Nuclear Information System (INIS)

    1985-01-01

    This Safety Guide is based on the report of an Advisory Group which met in Vienna in October 1984 in order to develop guidance on the radiation protection principles concerning emergency response planning and the establishment of intervention levels to be applied for the protection of the public in the event of a nuclear accident or radiological emergency. It considers the relationship between emergency response planning and various accident sequences, examines the pathways for radiation exposure and the sources of advice to decision makers during each of the three main accident phases, and specifies the dosimetric quantities that apply. The relevant pathological effects that must be protected against are summarized and the measures that may need to be implemented to provide protection with respect to each of the exposure pathways are discussed. It sets out the principles which underline decisions on intervention planning for each of the accident phases, gives guidance on dose values for the introduction of relevant protective measures and considers the application of cost-benefit analysis and the determination of the optimum dose level at which to withdraw protective measures

  10. Organization of the French emergency teams in the event of a radiological accident

    Energy Technology Data Exchange (ETDEWEB)

    Dumon, F. [Faculte de Pharmacie, 13 - Marseille (France); Pizzocaro, Y. [CSP, Risques Technologiques, 83 - Toulon (France)

    2001-07-01

    Nowadays, the intervention in ionizing environment is increasingly probable. It is still rare, but with the development of the nuclear programme of electricity production which was held in the french past and the significant rise in the use of the radioelements, as well in the medical field as industrial, the radioactive risk cannot be neglected. Technical and human resources, brought by mobile emergency teams called CMIR, were thus implemented to ensure either the safety of only hard-working exposed to the ionizing radiations, but also that of the population. In France, the organization of the public authorities in the event of nuclear accident, fixed by Directives of the Prime Minister which relate to nuclear safety, protection against radiation, the law and order and the civil safety, is described in Particular Plan for Intervention (PPI). (author)

  11. The protection of on-site personnel in the event of a radiological accident

    International Nuclear Information System (INIS)

    Morrey, M.; Simister, D.N.

    2003-01-01

    The National Radiological Protection Board (NPRB) is responsible in the UK for advising Government and other responsible bodies on the principles for responding to radiological emergencies. NRPB has published appropriate advice on the off-site protection of the public and on the protection of workers involved in taking mitigating actions to reduce the exposure of others. This paper puts forward a suggested framework for the protection of on-site personnel in the event of a radiological emergency which might include a criticality accident. This framework both dovetails with existing planning for the protection of members of the public off-site, and also takes account of specific differences between the situations on and off-site. (author)

  12. Organization of the French emergency teams in the event of a radiological accident

    International Nuclear Information System (INIS)

    Dumon, F.; Pizzocaro, Y.

    2001-01-01

    Nowadays, the intervention in ionizing environment is increasingly probable. It is still rare, but with the development of the nuclear programme of electricity production which was held in the french past and the significant rise in the use of the radioelements, as well in the medical field as industrial, the radioactive risk cannot be neglected. Technical and human resources, brought by mobile emergency teams called CMIR, were thus implemented to ensure either the safety of only hard-working exposed to the ionizing radiations, but also that of the population. In France, the organization of the public authorities in the event of nuclear accident, fixed by Directives of the Prime Minister which relate to nuclear safety, protection against radiation, the law and order and the civil safety, is described in Particular Plan for Intervention (PPI). (author)

  13. Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code

    International Nuclear Information System (INIS)

    Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges

    2003-01-01

    Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described

  14. Procedural and submittal guidance for the individual plant examination of external events (IPEEE) for severe accident vulnerabilities

    International Nuclear Information System (INIS)

    Chen, J.T.; Chokshi, N.C.; Kenneally, R.M.; Kelly, G.B.; Beckner, W.D.; McCracken, C.; Murphy, A.J.; Reiter, L.; Jeng, D.

    1991-06-01

    Based on a Policy statement on Severe Accidents, the licensee of each nuclear power plant is requested to perform an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents and cost-effective safety improvements that reduce or eliminate the important vulnerabilities. This document presents guidance for performing and reporting the results of the individual plant examination of external events (IPEEE). The guidance for reporting the results of the individual plant examination of internal events (IPE) is presented in NUREG-1335. 53 refs., 1 figs., 2 tabs

  15. Reactor accidents of four decades

    International Nuclear Information System (INIS)

    Szabo, Z.

    1982-11-01

    The report covers the period between 1942 and June 30, 1982. A detailed description and a comparative analysis of reactor accidents and chemical-processing-plant excursions are presented. The analysis takes into account the following points: causes (design, maintenance, operation); events (initiating event and sequence of events); consequences (environmental impacts, personnel effects and equipment damages). (author)

  16. Event Sequence Analysis of the Air Intelligence Agency Information Operations Center Flight Operations

    National Research Council Canada - National Science Library

    Larsen, Glen

    1998-01-01

    This report applies Event Sequence Analysis, methodology adapted from aircraft mishap investigation, to an investigation of the performance of the Air Intelligence Agency's Information Operations Center (IOC...

  17. Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP

    International Nuclear Information System (INIS)

    Kaliatka, A.; Uspuras, E.

    2001-01-01

    The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

  18. Thirteen- and Sixteen-Month-Olds' Long-Term Recall of Event Sequences.

    Science.gov (United States)

    Hertsgaard, L.; Bauer, P. J.

    In two experiments, the ability of children younger than 20 months to engage in delayed ordered recall was investigated. In the first experiment, 13- and 16-month-old children were presented with 2-step event sequences and tested for recall, first, immediately following the event and second, after a one-week delay. Sequences were novel-causal,…

  19. Constraining the magnitude of the largest event in a foreshock-main shock-aftershock sequence

    Science.gov (United States)

    Shcherbakov, Robert; Zhuang, Jiancang; Ogata, Yosihiko

    2018-01-01

    Extreme value statistics and Bayesian methods are used to constrain the magnitudes of the largest expected earthquakes in a sequence governed by the parametric time-dependent occurrence rate and frequency-magnitude statistics. The Bayesian predictive distribution for the magnitude of the largest event in a sequence is derived. Two types of sequences are considered, that is, the classical aftershock sequences generated by large main shocks and the aftershocks generated by large foreshocks preceding a main shock. For the former sequences, the early aftershocks during a training time interval are used to constrain the magnitude of the future extreme event during the forecasting time interval. For the latter sequences, the earthquakes preceding the main shock are used to constrain the magnitudes of the subsequent extreme events including the main shock. The analysis is applied retrospectively to past prominent earthquake sequences.

  20. Characterization of GM events by insert knowledge adapted re-sequencing approaches

    OpenAIRE

    Yang, Litao; Wang, Congmao; Holst-Jensen, Arne; Morisset, Dany; Lin, Yongjun; Zhang, Dabing

    2013-01-01

    Detection methods and data from molecular characterization of genetically modified (GM) events are needed by stakeholders of public risk assessors and regulators. Generally, the molecular characteristics of GM events are incomprehensively revealed by current approaches and biased towards detecting transformation vector derived sequences. GM events are classified based on available knowledge of the sequences of vectors and inserts (insert knowledge). Herein we present three insert knowledge-ad...

  1. Tacit Driving Knowledge, Emotional Intelligence, Stressful Events and Accident Risk: Traffic Safety Implications

    National Research Council Canada - National Science Library

    Legree, Peter

    1999-01-01

    ... and the driver's internal or emotional state. The tests were administered with a battery of conventional cognitive tests, personality instruments and situational variables chosen to predict accident involvement...

  2. Predicting Consequences of Technological Disasters from Natural Hazard Events: Challenges and Opportunities Associated with Industrial Accident Data Sources

    Science.gov (United States)

    Wood, M.

    2009-04-01

    The increased focus on the possibility of technological accidents caused by natural events (Natech) is foreseen to continue for years to come. In this case, experts in prevention, mitigation and preparation activities associated with natural events will increasingly need to borrow data and expertise traditionally associated with the technological fields to carry out the work. An important question is how useful is the data for understanding consequences from such natech events. Data and case studies provided on major industrial accidents tend to focus on lessons learned for re-engineering the process. While consequence data are reported at least nominally in most reports, their precision, quality and completeness is often lacking. Consequences that are often or sometimes available but not provided can include severity and type of injuries, distance of victims from the source, exposure measurements, volume of the release, population in potentially affected zones, and weather conditions. Yet these are precisely the type of data that will aid natural hazard experts in land-use planning and emergency response activities when a Natech event may be foreseen. This work discusses the results of a study of consequence data from accidents involving toxic releases reported in the EU's MARS accident database. The study analysed the precision, quality and completeness of three categories of consequence data reported: the description of health effects, consequence assessment and chemical risk assessment factors, and emergency response information. This work reports on the findings from this study and discusses how natural hazards experts might interact with industrial accident experts to promote more consistent and accurate reporting of the data that will be useful in consequence-based activities.

  3. Proposal of methodology of tsunami accident sequence analysis induced by earthquake using DQFM methodology

    International Nuclear Information System (INIS)

    Muta, Hitoshi; Muramatsu, Ken

    2017-01-01

    Since the Fukushima-Daiichi nuclear power station accident, the Japanese regulatory body has improved and upgraded the regulation of nuclear power plants, and continuous effort is required to enhance risk management in the mid- to long term. Earthquakes and tsunamis are considered as the most important risks, and the establishment of probabilistic risk assessment (PRA) methodologies for these events is a major issue of current PRA. The Nuclear Regulation Authority (NRA) addressed the PRA methodology for tsunamis induced by earthquakes, which is one of the methodologies that should be enhanced step by step for the improvement and maturity of PRA techniques. The AESJ standard for the procedure of seismic PRA for nuclear power plants in 2015 provides the basic concept of the methodology; however, details of the application to the actual plant PRA model have not been sufficiently provided. This study proposes a detailed PRA methodology for tsunamis induced by earthquakes using the DQFM methodology, which contributes to improving the safety of nuclear power plants. Furthermore, this study also states the issues which need more research. (author)

  4. Iodine chemistry effect on source term assessments. A MELCOR 186 YT study of a PWR severe accident sequence

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Otero, Bernadette

    2009-01-01

    Level-2 Probabilistic Safety Analysis has demonstrated to be a powerful tool to give insights into multiple aspects concerning severe accidents: phenomena with the greatest potential to lead to containment failure, safety systems performance and, even, to identify any additional accident management that could mitigate the consequences of such an even, etc. A major result of level-2 PSA is iodine content in Source Term since it is the main responsible for the radiological impact during the first few days after a hypothetical severe accident. Iodine chemistry is known to considerably affect iodine behavior and although understanding has improved substantially since the early 90's, a thorough understanding is still missing and most PSA studies do not address it when assessing severe accident scenarios. This paper emphasizes the quantitative and qualitative significance of considering iodine chemistry in level-2 PSA estimates. To do so a cold leg break, low pressure severe accident sequence of an actual pressurized water reactor has been analyzed with the MELCOR 1.8.6 YT code. Two sets of calculations, with and without chemistry, have been carried out and compared. The study shows that iodine chemistry could result in an iodine release to environment about twice higher, most of which would consist of around 60% of iodine in gaseous form. From these results it is concluded that exploratory studies on the potential effect of iodine chemistry on source term estimates should be carried out. (author)

  5. Identification of important ''PIUS'' design considerations and accident sequences using qualitative plant assessment techniques

    International Nuclear Information System (INIS)

    Higgins, J.; Fullwood, R.; Kroeger, P.; Youngblood, R.

    1992-01-01

    The PIUS (Process Inherent Ultimate Safety) reactor is an advanced design nuclear power plant that uses passive safety features and basic physical processes to address safety concerns. Brookhaven National Laboratory (BNL) performed a detailed study of the PIUS design for the NRC using primarily qualitative engineering analysis techniques. Some quantitative methods were also employed. There are three key initial areas of analysis: FMECA, HAZOP, and deterministic analyses, which are described herein. Once these three analysis methods were completed, the important findings from each of the methods were assembled into thePIUS Interim Table (PIT). This table thus contains a first cut sort of the important design considerations and features of the PIUS reactor. The table also identifies some potential initiating events and systems used for mitigating these initiators. The next stage of the analysis was the construction of event trees for each of the identified initiators. The most significant sequences were then determined qualitatively, using, some quantitative input. Finally, overall insights on the PIUS design developed from the PIT and from the event tree analysis were developed and presented

  6. Phenomenological analyses and their application to the Defense Waste Processing Facility probabilistic safety analysis accident progression event tree. Revision 1

    International Nuclear Information System (INIS)

    Kalinich, D.A.; Thomas, J.K.; Gough, S.T.; Bailey, R.T.; Kearnaghan, D.P.

    1995-01-01

    In the Defense Waste Processing Facility (DWPF) Safety Analysis Reports (SARs) for the Savannah River Site (SRS), risk-based perspectives have been included per US Department of Energy (DOE) Order 5480.23. The NUREG-1150 Level 2/3 Probabilistic Risk Assessment (PRA) methodology was selected as the basis for calculating facility risk. The backbone of this methodology is the generation of an Accident Progression Event Tree (APET), which is solved using the EVNTRE computer code. To support the development of the DWPF APET, deterministic modeling of accident phenomena was necessary. From these analyses, (1) accident progressions were identified for inclusion into the APET; (2) branch point probabilities and any attendant parameters were quantified; and (3) the radionuclide releases to the environment from accidents were determined. The phenomena of interest for accident progressions included explosions, fires, a molten glass spill, and the response of the facility confinement system during such challenges. A variety of methodologies, from hand calculations to large system-model codes, were used in the evaluation of these phenomena

  7. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  8. Current status of low power/shutdown PSA and accident sequence analysis for loss of RHR during mid-loop operation

    International Nuclear Information System (INIS)

    Park, Chang Kyu; Choi, Young; Kim, Tae Woon; Jin, Young Ho

    1994-07-01

    Probabilistic safety assessment (PSA) has been applied to only full-power operation of nuclear power plant (NPP), but some events which were recently occurred could reach severe plant damage state. Thus, various countries around the world have focused their interests on the evaluation for low power/shutdown (LP/S) operation. This report covers the main stream of LP/S PSA methodology, current status of LP/S PSA practices and results, and accident sequence analysis for loss of RHR during mid-loop operation. Therefore this report would be helpful for us to practice LP/S PSA for YGN 5,6 NPP which will be built in the near future. Also the results of accident sequence analysis show that operator's mis-diagnosis and failure of recovery action would initiate core damage during LP/S operation. In summary, overall environmental improvements (equipments, procedures, Tech Spec, etc, ...) and operating support system will be very useful to reduce risk during LP/S operation. (Author) 5 figs., 9 tabs

  9. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

  10. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis

  11. Device for bonding iodine in the event of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Hladik, O.

    1988-01-01

    A device for bonding iodine, in particular radioiodine released during nuclear reactor accidents, is presented. Radioiodine is bonded, even at high temperatures, so that it is neither volatile nor soluble

  12. Rules concerning radiation protection measures to be taken in the event of accidents with gamma radiography

    International Nuclear Information System (INIS)

    1981-01-01

    These rules issued by the State Institute of Radiation Hygiene (SIS) set out the obligations of licensees and radiographers regarding radiation protection and accident prevention in relation to gamma radiography. (NEA)

  13. The sequence coding and search system: An approach for constructing and analyzing event sequences at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Mays, G.T.

    1989-04-01

    The US Nuclear Regulatory Commission (NRC) has recognized the importance of the collection, assessment, and feedstock of operating experience data from commercial nuclear power plants and has centralized these activities in the Office for Analysis and Evaluation of Operational Data (AEOD). Such data is essential for performing safety and reliability analyses, especially analyses of trends and patterns to identify undesirable changes in plant performance at the earliest opportunity to implement corrective measures to preclude the occurrences of a more serious event. One of NRC's principal tools for collecting and evaluating operating experience data is the Sequence Coding and Search System (SCSS). The SCSS consists of a methodology for structuring event sequences and the requisite computer system to store and search the data. The source information for SCSS is the Licensee Event Report (LER), which is a legally required document. This paper describes the objective SCSS, the information it contains, and the format and approach for constructuring SCSS event sequences. Examples are presented demonstrating the use SCSS to support the analysis of LER data. The SCSS contains over 30,000 LERs describing events from 1980 through the present. Insights gained from working with a complex data system from the initial developmental stage to the point of a mature operating system are highlighted

  14. Radiation protection service for a nucleonic control system of continuous casting plant after events of accident

    International Nuclear Information System (INIS)

    Chakrabarti, Santanu; Massand, O.P.

    1998-01-01

    Extensive use of nucleonic control systems like level controllers was observed during radiation protection surveys in industries such as refineries, steel plants etc., located in the eastern region of India. There were two accidents at continuous casting plant in 1995 which affected the nucleonic control system installed in 1992. The authorities contacted Bhabha Atomic Research Centre (BARC) for radiation protection surveys for the involved nucleonic gauges. The present paper describes the radiation protection services rendered by BARC during such accidents. (author)

  15. Causal Factors and Adverse Events of Aviation Accidents and Incidents Related to Integrated Vehicle Health Management

    Science.gov (United States)

    Reveley, Mary S.; Briggs, Jeffrey L.; Evans, Joni K.; Jones, Sharon M.; Kurtoglu, Tolga; Leone, Karen M.; Sandifer, Carl E.

    2011-01-01

    Causal factors in aviation accidents and incidents related to system/component failure/malfunction (SCFM) were examined for Federal Aviation Regulation Parts 121 and 135 operations to establish future requirements for the NASA Aviation Safety Program s Integrated Vehicle Health Management (IVHM) Project. Data analyzed includes National Transportation Safety Board (NSTB) accident data (1988 to 2003), Federal Aviation Administration (FAA) incident data (1988 to 2003), and Aviation Safety Reporting System (ASRS) incident data (1993 to 2008). Failure modes and effects analyses were examined to identify possible modes of SCFM. A table of potential adverse conditions was developed to help evaluate IVHM research technologies. Tables present details of specific SCFM for the incidents and accidents. Of the 370 NTSB accidents affected by SCFM, 48 percent involved the engine or fuel system, and 31 percent involved landing gear or hydraulic failure and malfunctions. A total of 35 percent of all SCFM accidents were caused by improper maintenance. Of the 7732 FAA database incidents affected by SCFM, 33 percent involved landing gear or hydraulics, and 33 percent involved the engine and fuel system. The most frequent SCFM found in ASRS were turbine engine, pressurization system, hydraulic main system, flight management system/flight management computer, and engine. Because the IVHM Project does not address maintenance issues, and landing gear and hydraulic systems accidents are usually not fatal, the focus of research should be those SCFMs that occur in the engine/fuel and flight control/structures systems as well as power systems.

  16. Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

    1999-01-01

    In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10 -4 and 10 -6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10 -9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

  17. A trend analysis of human error events for proactive prevention of accidents. Methodology development and effective utilization

    International Nuclear Information System (INIS)

    Hirotsu, Yuko; Ebisu, Mitsuhiro; Aikawa, Takeshi; Matsubara, Katsuyuki

    2006-01-01

    This paper described methods for analyzing human error events that has been accumulated in the individual plant and for utilizing the result to prevent accidents proactively. Firstly, a categorization framework of trigger action and causal factors of human error events were reexamined, and the procedure to analyze human error events was reviewed based on the framework. Secondly, a method for identifying the common characteristics of trigger action data and of causal factor data accumulated by analyzing human error events was clarified. In addition, to utilize the results of trend analysis effectively, methods to develop teaching material for safety education, to develop the checkpoints for the error prevention and to introduce an error management process for strategic error prevention were proposed. (author)

  18. Object-Oriented Query Language For Events Detection From Images Sequences

    Science.gov (United States)

    Ganea, Ion Eugen

    2015-09-01

    In this paper is presented a method to represent the events extracted from images sequences and the query language used for events detection. Using an object oriented model the spatial and temporal relationships between salient objects and also between events are stored and queried. This works aims to unify the storing and querying phases for video events processing. The object oriented language syntax used for events processing allow the instantiation of the indexes classes in order to improve the accuracy of the query results. The experiments were performed on images sequences provided from sport domain and it shows the reliability and the robustness of the proposed language. To extend the language will be added a specific syntax for constructing the templates for abnormal events and for detection of the incidents as the final goal of the research.

  19. Initiating events and accidental sequences taken into account in the CAREM reactor design

    International Nuclear Information System (INIS)

    Kay, J.M.; Felizia, E.R.; Navarro, N.R.; Caruso, G.J.

    1990-01-01

    The advance made in the nuclear security evaluation of the CAREM reactor is presented. It was carried out using the Security Probabilistic Analysis (SPA). The latter takes into account the different phases of identification and solution of initiating events and the qualitative development of event trees. The method of identification of initiating events is the Master Logical Diagram (MLD), whose deductive basis makes it appropriate for a new design like the one described. The qualitative development of the event trees associated to the identified initiating events, allows identification of those accidental sequences which are to have the security systems in the reactor. (Author) [es

  20. Iodine prophylaxis following nuclear accidents - a concept how to distribute potassium-iodide tablets out of the central stocks in the event of an accident

    International Nuclear Information System (INIS)

    Portius, U.

    2007-01-01

    With its recommendation ''Iodine prophylaxis following nuclear accidents'' (1996) and its reports of 1997 and 2001 the German Commission on Radiological Protection (SSK) followed the recommendations of the WHO ''Guidelines for iodine prophylaxis following nuclear accidents'' of 1989. The intervention levels were lowered (50 mSv for children/adolescents (up to the age of 18 years) and pregnant women, 250 mSv for adults), the iodine prophylaxis was restricted to persons up to the age of 45 years and the recommended dosage of stable iodine was changed. Due to the lowered reference levels the radius of 25 km around a nuclear power plant that had been the planning radius for the distribution of iodine tablets so far was extended to 100 km. Based on these recommendations the German authorities began to set up new strategies for the provision and distribution of potassium-iodide tablets (iodine tablets). Since 2004, within the radius of 25 km the iodine tablets are pre-distributed to households and/or stored at several points in the municipality for persons up to the age of 45 years. For the new planning radius of 25-100 km iodine tablets are stored in 8 central stocks in Germany for children/adolescents (up to the age of 18 years) and pregnant women. A working group with representatives from federal and Laender authorities has developed a distribution strategy for the distribution out of these central stocks in the event of an accident. It describes a possibility of organising and implementing the distribution of the iodine tablets within the radius of 25-100 km in a nationwide standardised way. (orig.)

  1. A new approach to incorporate operator actions in the simulation of accident sequences

    International Nuclear Information System (INIS)

    Antonio Exposito; Juan Antonio Quiroga; Javier Hortal; John-Einar Hulsund

    2006-01-01

    Full text of publication follows: Nowadays, simulation-based human reliability analysis (HRA) methods seem to provide a new direction for the development of advanced methodologies to study operator actions effect during accident sequences. Due to this, the Spanish Nuclear Safety Council (CSN) started a working group which has, among other objectives, to develop such simulation-based HRA methodology. As a result of its activities, a new methodology, named Integrated Safety Assessment (ISA), has been developed and is currently being incorporated into licensing activities at CSN. One of the key aspects of this approach is the incorporation of the capability to simulate operator actions, expanding the ISA methodology scopes to make HRA studies. For this reason, CSN is involved in several activities oriented to develop a new tool, which must be able to incorporate operator actions in conventional thermohydraulic (TH) simulations. One of them is the collaboration project between CSN, Halden Reactor Project (HRP) and the Department of Energy Systems (DSE) of the Polytechnic University of Madrid that started in 2003. The basic aim of the project is to develop a software tool that consists of a closed-loop plant/operator simulator, a thermal hydraulic (TH) code for simulating the plant transient and the procedures processor to give the information related with operator actions to the TH code, both coupled by a data communication system which allows the information exchange. For the plant simulation we have a plant transient simulator code (TRETA/TIZONA for PWR/BWR NPPs respectively), developed by the CSN, with PWR/BWR full scope models. The functionality of these thermalhydraulic codes has been expanded, allowing control the overall information flow between coupled codes, simulating the TH transient and determining when the operator actions must be considered. In the other hand, we have the COPMA-III code, a computerized procedure system able to manage XML operational

  2. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States)] [and others

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

  3. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendices F-H, Volume 2, Part 4

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis

  4. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M.; Bley, D.; Johnson, D.

    1994-06-01

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis

  5. Assessment and limitation of radioactivity transfers in the event of a postulated severe PWR accident

    International Nuclear Information System (INIS)

    Gauvain, J.

    1992-01-01

    This report constitutes the supporting material for a lecture on severe accidents which could occur on PWR type nuclear reactors. It is assumed for present purposes that the reader has at least a rudimentary acquaintance with the basics of general physics if not with the operating processes of these reactors. After defining what is meant by a ''severe accident'' on a reactor, the possible phenomenology of such an accident is qualitatively described: loss of coolant and loss of containment integrity. A certain number of elements are then given for the quantitative assessment of these phenomena involving possible radioactivity transfers within and outside the plant. In conclusion, available means are indicated for the limitation and control of these environmental transfers. (author). 5 refs, figs

  6. Fukushima. The accident sequence and important causes. Pt. 2/3; Fukushima. Unfallablauf und wesentliche Ursachen. T. 2/3

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany). Bereich Nukleartechnik und Anlagensicherheit

    2013-07-01

    In this part on the accident sequence in the NPP Fukushima Daiichi on March 11, 2011 the important safety systems of a nuclear power plant are described, including the design of a nuclear boiling water reactor with Mark-II type containment, the high-pressure injection system and the systems for afterheat removal. The chronology of the accident progress in the NPP units 1-3 is described. The units 4-6 were shutdown due to revision work. Due to the earthquake an electric power transformation station close to the NPP site and the power poles were destroyed, the redundant power supply of the neighboring electricity supplier Tohoku did not work. All emergency diesel generators were flooded and destroyed resulting in the so-called station blackout. Firefighting trucks and materials for radiation protection and the infrastructure at the NPP site were destroyed. The release of radioactivity induced a severe contamination of the reactor site.

  7. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Analysis of core damage frequency from internal events for Plant Operational State 5 during a refueling outage. Volume 2, Part 2: Internal Events Appendices A to H

    International Nuclear Information System (INIS)

    Darby, J.; Whitehead, D.; Staple, B.; Dandini, V.

    1994-06-01

    This document contains the accident sequence analysis of internally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 during a refueling outage. The report documents the methodology used during the analysis, describes the results from the application of the methodology, and compares the results with the results from two full power analyses performed on Grand Gulf

  8. ISVASE: identification of sequence variant associated with splicing event using RNA-seq data.

    Science.gov (United States)

    Aljohi, Hasan Awad; Liu, Wanfei; Lin, Qiang; Yu, Jun; Hu, Songnian

    2017-06-28

    Exon recognition and splicing precisely and efficiently by spliceosome is the key to generate mature mRNAs. About one third or a half of disease-related mutations affect RNA splicing. Software PVAAS has been developed to identify variants associated with aberrant splicing by directly using RNA-seq data. However, it bases on the assumption that annotated splicing site is normal splicing, which is not true in fact. We develop the ISVASE, a tool for specifically identifying sequence variants associated with splicing events (SVASE) by using RNA-seq data. Comparing with PVAAS, our tool has several advantages, such as multi-pass stringent rule-dependent filters and statistical filters, only using split-reads, independent sequence variant identification in each part of splicing (junction), sequence variant detection for both of known and novel splicing event, additional exon-exon junction shift event detection if known splicing events provided, splicing signal evaluation, known DNA mutation and/or RNA editing data supported, higher precision and consistency, and short running time. Using a realistic RNA-seq dataset, we performed a case study to illustrate the functionality and effectiveness of our method. Moreover, the output of SVASEs can be used for downstream analysis such as splicing regulatory element study and sequence variant functional analysis. ISVASE is useful for researchers interested in sequence variants (DNA mutation and/or RNA editing) associated with splicing events. The package is freely available at https://sourceforge.net/projects/isvase/ .

  9. Characterization of GM events by insert knowledge adapted re-sequencing approaches.

    Science.gov (United States)

    Yang, Litao; Wang, Congmao; Holst-Jensen, Arne; Morisset, Dany; Lin, Yongjun; Zhang, Dabing

    2013-10-03

    Detection methods and data from molecular characterization of genetically modified (GM) events are needed by stakeholders of public risk assessors and regulators. Generally, the molecular characteristics of GM events are incomprehensively revealed by current approaches and biased towards detecting transformation vector derived sequences. GM events are classified based on available knowledge of the sequences of vectors and inserts (insert knowledge). Herein we present three insert knowledge-adapted approaches for characterization GM events (TT51-1 and T1c-19 rice as examples) based on paired-end re-sequencing with the advantages of comprehensiveness, accuracy, and automation. The comprehensive molecular characteristics of two rice events were revealed with additional unintended insertions comparing with the results from PCR and Southern blotting. Comprehensive transgene characterization of TT51-1 and T1c-19 is shown to be independent of a priori knowledge of the insert and vector sequences employing the developed approaches. This provides an opportunity to identify and characterize also unknown GM events.

  10. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    Dallman, R.J.; Gottula, R.C.; Holcomb, E.E.; Jouse, W.C.; Wagoner, S.R.; Wheatley, P.D.

    1987-05-01

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

  11. Risk evaluation method for faults by engineering approach. (2) Application concept of margin analysis utilizing accident sequences

    International Nuclear Information System (INIS)

    Kamiya, Masanobu; Kanaida, Syuuji; Kamiya, Kouichi; Sato, Kunihiko; Kuroiwa, Katsuya

    2016-01-01

    The influence of the fault displacement on the facility should to be evaluated not only by the activity of the fault but also by obtaining risk information by considering scenarios including such as the frequency and the degree of the hazard, which should be an appropriate approach for nuclear safety. An applicable concept of margin analysis utilizing accident sequences for evaluating the influence of the fault displacement is proposed. By use of this analysis, we can evaluate of the safety functions and margin for core damage, verify the efficiency of equipment of portable type and make a decision to take additional measures to reduce the risk by using obtained risk information. (author)

  12. Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

    2015-05-15

    The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

  13. Derivation of working levels for animal feedstuffs for use in the event of a future nuclear accident

    International Nuclear Information System (INIS)

    Nisbet, A.; Woodman, R.; Brown, J.

    1998-04-01

    In the event of a future nuclear accident, European Council Food Intervention Levels (CFILs) would be legally binding for foodstuffs marketed in the UK. Practical guidance has been developed on the activity concentrations of radiocaesium and radiostrontium in animal feedstuffs that would give rise to concentrations equivalent to the relevant CFIL in the final animal product. The animals considered were dairy and beef cattle, lambs, pigs, broiler chickens and laying hens. Typical diets have been derived for each animal. The NRPB foodchain model FARMLAND has been used to predict activity concentrations in different feedstuffs for accidents occurring at different times of the year. The predicted concentrations were combined with the data on dietary composition, information on feed-to-product transfer and the relevant CFIL to estimate the corresponding Working levels in Animal Feedstuffs (WAFs). The calculations were carried out using a dedicated software system called SILAFOD. This flexible system can be used to carry out more specific assessments. A handbook that accompanies this report contains detailed information on animal diets, contributions from various feedstuffs to intakes of activity and the corresponding WAFs. The early phase after an accident and the longer-term phase are both considered. The work received partial financial support from the Ministry of Agriculture, Fisheries and Food, Radiological Safety and Nutrition Division. (author)

  14. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    Science.gov (United States)

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  15. The accident of the Fukushima-Daiichi nuclear plant. Status two years after the event

    International Nuclear Information System (INIS)

    2013-03-01

    In a first part, this report briefly recalls the circumstances and occurrence of the accident, gives an overview of actions undertaken by the IRSN (calculations of installation damages, modelling of contaminated air movements, simulations of radionuclide dispersion in the sea environment, information of French nationals in Japan, press and public information), and an overview of strength tests of nuclear installations (additional safety assessments and European stress tests). The second part gives an overview of the situation in Japan two years after the accident: evolution of governance in terms of nuclear risk management, condition of the Fukushima plant in January 2013, health and environmental impact and post-accidental management, actions undertaken by the IRSN (assessment of doses potentially received by populations, strengthening of cooperation between Japan and France in the field of severe accidents, participation to the Fukushima Dialogue). The third part presents the contribution of the IRSN to the strengthening of nuclear safety and radiation protection at the international level, at the European level, and in France

  16. Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

    International Nuclear Information System (INIS)

    Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

    1998-01-01

    Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

  17. Investigations of postulated accident sequences for the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Hatta, M.; Sanders, J.P.

    1978-01-01

    The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models

  18. On the sequence of core-melt accidents: Fission product release, source terms and Chernobyl release

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht, H

    1986-01-01

    There is a sketch of our ideas on the course of a core melt-out accident in a PWR. There is then a survey of the most important results on fission product release, which were obtained by experiments on the SASCHA melt-out plant. The 3rd part considers questions which are important for determining source terms for the environment and the last part contains some considerations on radioactivity release from the Chernobyl reactor.

  19. FRANTIC-NRC, Accident Sequence and Event Tree Analysis for System Availability and Operation

    International Nuclear Information System (INIS)

    Ginzburg, T.

    1988-01-01

    1 - Description of problem or function: FRANTIC3 was developed to evaluate system unreliability using time-dependent techniques. The code provides two major options: to evaluate standby system unavailability or, in addition to the unavailability, to calculate the total system failure probability by including both the unavailability of the system on demand as well as the probability that it will operate for an arbitrary time period following the demand. The FRANTIC time-dependent reliability models provide a large selection of repair and testing policies applicable to standby or continuously operating systems consisting of periodically tested, monitored, and non-repairable (non-testable) components. Time- dependent and test frequency dependent failures, as well as demand stress related failure, test-caused degradation and wear-out, test associated human errors, test deficiencies, test override, unscheduled and scheduled maintenance, component renewal and replacement policies, and test strategies can be prescribed. The conditional system unavailabilities associated with the downtimes of the user specified failed component are also evaluated. Optionally, the code can perform a sensitivity study for system unavailability or total failure probability to the failure characteristics of the standby components. 2 - Method of solution: FRANTIC3 uses a set of analytical equations for component unavailabilities and failure intensities with exponential and Weibull time distributions for constant test and duration times. The FRANTIC code determines the state (test, repair, or between test) of each component at each time point and selects the appropriate component unavailability model depending on the state. Using the appropriate logical relationships among the unavailabilities of individual components, the system unavailability (or total failure probability) is also calculated at these time points. Then, the system average unavailability (or total failure probability) is calculated by integrating the instantaneous unavailability (or total failure probability) over the standby time period. A Boolean equation is specified by the user via the SYSCOM subroutine to define the system unavailability functions. The system failure occurrence rate functions needed for the system unreliability evaluation is user defined via the SYSOP subroutine

  20. On high-temperature reactor accident topology

    International Nuclear Information System (INIS)

    Fassbender, J.; Kroeger, W.; Wolters, J.

    1981-01-01

    American and German risk studies for an HTGR and independent investigations of hypothetical accident sequences led to a fundamental understanding of the topology of HTGR accident sequences. The dominating importance of core heat-up accidents was confirmed and the initiating events were identified. Complications of core heat-up accidents by air or water ingress are of minor importance for the risk, whereas the long-term development of accidents during days and weeks plays an important role for the environmental impact. The risk caused by an HTGR at a German site cannot yet be determined exactly, because no modern German HTGR design has passed a licensing procedure. Cautious estimates show that risk will appear to be substantially smaller than the LWR risk. The main reasons are the considerably reduced release of fission procucts and the slow development of core heat-up accidents leaving much time for measures which reduce the risk. (orig.) [de

  1. Modeling the Process of Event Sequence Data Generated for Working Condition Diagnosis

    Directory of Open Access Journals (Sweden)

    Jianwei Ding

    2015-01-01

    Full Text Available Condition monitoring systems are widely used to monitor the working condition of equipment, generating a vast amount and variety of telemetry data in the process. The main task of surveillance focuses on analyzing these routinely collected telemetry data to help analyze the working condition in the equipment. However, with the rapid increase in the volume of telemetry data, it is a nontrivial task to analyze all the telemetry data to understand the working condition of the equipment without any a priori knowledge. In this paper, we proposed a probabilistic generative model called working condition model (WCM, which is capable of simulating the process of event sequence data generated and depicting the working condition of equipment at runtime. With the help of WCM, we are able to analyze how the event sequence data behave in different working modes and meanwhile to detect the working mode of an event sequence (working condition diagnosis. Furthermore, we have applied WCM to illustrative applications like automated detection of an anomalous event sequence for the runtime of equipment. Our experimental results on the real data sets demonstrate the effectiveness of the model.

  2. Analyses of conditions in a large, dry PWR containment during an TMLB' accident sequence

    International Nuclear Information System (INIS)

    Sweet, D.W.; Roberts, G.J.

    1994-01-01

    The aim of the paper is to give an assessment of the conditions which would develop in the large, dry containment of a modern Westinghouse-type PWR during a severe accident where all safety systems are unavailable. The analysis is based principally on the results of calculations using the CONTAIN code, with a 4 cell model of the containment, for a station blackout (TMLB') scenario in which the vessel is assumed to fail at high pressure. In particular, the following are noted: (i) If much of the debris is in contact with water, so that decay heat can boil water directly, then the pressure rises steadily to reach the assumed containment failure point after 11/2 to 2 days. If most of the debris becomes isolated from water, for example, because of water is held up on the containment floors and in sumps and drains, the pressure rises too slowly to threaten the containment on this timescale. (ii) If a core-concrete interaction occurs, most of the associated fission product release takes place soon after relocation of molten fuel to the containment. The aerosols which transport these (and other non-gaseous fission products released earlier in the accident) in the containment agglomerate and settle. As a result, 0.1% or less of the aerosols remain airborne a day after the start of the accident. (iii) Hydrogen and carbon monoxide, which would accumulate in the containment are not expected to burn because the atmosphere would be inerted by steam. If, however, enough of the steam is condensed, for example, by recovering the containment sprays, a burn could occur but the resulting pressure spike is unlikely to threaten the containment unless a transition to detonation occurs. 6 refs., 6 tabs., 12 figs

  3. Basic principles and criteria for public health protection in the event of a nuclear accident

    International Nuclear Information System (INIS)

    Kiradzhiev, G.

    1992-01-01

    Decision making criteria for population protection in nuclear accidents are discussed, and in particular the three basic principles: 1) excluding the appearance of nonstostochastic effects that occur in the case of high individual doses; 2) weighing the risks of radiation damage if such measures are not taken; 3) optimization based on comparison of benefit and costs, using the same measures for costs of health injury to affected populations and of the protected measures to be taken. The decision making criteria developed in Bulgaria are based on international recommendations with lowered upper limit of the range for evacuation and specified doses for vulnerable groups, children and pregnant women. The organization and the specific problems of the following individual types of protective measures are described: sheltering; protection of respiratory organs; iodine prophylaxis; evacuation of the public. One major condition for ensuring protection is to provide the public with timely information on the actual situation and the necessary countermeasures. Such information should be released in a manner that allows for understanding the expediency and significance of actions to be taken. An important aspect of emergency planning consists in taking into consideration the conditions actually prevailing in the country. This is well illustrated in the principle designated as 'national level of challenge' taking into account a country's capabilities for introducing intervention levels and permissible dose levels. In the case of Bulgaria this still remains to be done in protective planning for accidents. (author)

  4. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  5. SCPRI Emergency Kit for Use in the Event of a Nuclear Accident; Le Dispositif d'Intervention Rapide du SCPRI en Cas d'Accident Nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Ervet, P.; Moroni, J. P.; Pellerin, P. [Service Central de Protection Contre les Rayonnements Ionisants, Ministere des Affaires Sociales, Le Vesinet (France)

    1969-10-15

    In the event of a nuclear accident necessitating implementation of the ORSEC radiation protection plan, the Service central de protection contre les rayonnements ionisants (Central Service for Protection against Ionizing Radiations), in conjunction with the Service national de la protection civile (National Civil Defence Service), has adopted the necessary measures for rapid evaluation of possible contamination as promptly as possible. With this aim in mind the Service has prepared emergency kits, which are permanently stored at airfields in the Paris region; these can be carried by aircraft together with two engineers from the Service, thereby enabling them to reach the site of the incident with the specialized equipment in a few hours at most. This paper describes the monitoring and sampling equipment as well as the conditions under which the kit is carried and used (it operates independently by having a built-in generating unit). It is basically designed to permit an initial assessment of the situation, to furnish local authorities with data on which to base decisions for the safety of the population, and to determine any additional measures that need to be adopted. (author) [French] Dans le cas d'un accident nucleaire impliquant la mise en application du plan ORSEC radiologique, en liaison avec le Service national de la protection civile, le Service central de protection contre les rayonnements ionisants a pris les dispositions necessaires pour faire une evaluation rapide, aussi preooce que possible, des contaminations eventuelles. Dans ce but, il a realise des cantines d'intervention qui sont deposees en permanence sur les aerodromes de la region parisienne, et peuvent etre embarquees par avion avec deux ingenieurs du service qui peuvent etre ainsi sur les lieux de l'incident, avec un materiel specialise, dans un delai qui n'excede pas quelques heures. Le memoire decrit le materiel de mesure et de prelevement, ainsi que les conditions de transport et d

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  7. Sensitivity to structure in action sequences: An infant event-related potential study.

    Science.gov (United States)

    Monroy, Claire D; Gerson, Sarah A; Domínguez-Martínez, Estefanía; Kaduk, Katharina; Hunnius, Sabine; Reid, Vincent

    2017-05-06

    Infants are sensitive to structure and patterns within continuous streams of sensory input. This sensitivity relies on statistical learning, the ability to detect predictable regularities in spatial and temporal sequences. Recent evidence has shown that infants can detect statistical regularities in action sequences they observe, but little is known about the neural process that give rise to this ability. In the current experiment, we combined electroencephalography (EEG) with eye-tracking to identify electrophysiological markers that indicate whether 8-11-month-old infants detect violations to learned regularities in action sequences, and to relate these markers to behavioral measures of anticipation during learning. In a learning phase, infants observed an actor performing a sequence featuring two deterministic pairs embedded within an otherwise random sequence. Thus, the first action of each pair was predictive of what would occur next. One of the pairs caused an action-effect, whereas the second did not. In a subsequent test phase, infants observed another sequence that included deviant pairs, violating the previously observed action pairs. Event-related potential (ERP) responses were analyzed and compared between the deviant and the original action pairs. Findings reveal that infants demonstrated a greater Negative central (Nc) ERP response to the deviant actions for the pair that caused the action-effect, which was consistent with their visual anticipations during the learning phase. Findings are discussed in terms of the neural and behavioral processes underlying perception and learning of structured action sequences. Copyright © 2017 Elsevier Ltd. All rights reserved.

  8. An estimation of the accident sequence of the LOCA groups for the PSA model of the KSNP

    International Nuclear Information System (INIS)

    Han, Seok Jung; Yang, Joon Eon

    2004-01-01

    A new trend of the probabilistic safety assessment (PSA) technology is to improve and enhance the current PSA model to be adequate for risk-informed applications (RIA). Requirements of a PSA model for the RIA are summarized as (1) reduction of the conservatism in the model utilizing all available information and (2) consideration of the specific features of a plant as designed, as operated. This is because the PSA based on conservatism and insufficient consideration of the plant-specific features resulted in a shadow effect on the assessment results. When a PSA model is used in a risk-informed application, more precise risk-information is more helpful to decision making process, so the reduction of the conservatism and the consideration of the plant-specific features in a PSA model are the most essential elements. Recently, an effort has been performed to modify the current PSA model for the Korea Standard Nuclear Power plant (KSNP) to be used in risk-informed applications. A re-estimation of the accident sequence of the loss of coolant accident (LOCA) groups for the PSA model of the KSNP has been performed

  9. Preliminary steps towards assessing aerosol retention in the break stage of a dry steam generator during severe accident SGTR sequences

    International Nuclear Information System (INIS)

    Herranz, L.E.; Lopez del Pra, C.; Sanchez Velasco, F.J.

    2006-01-01

    Severe accidents SGTR sequences are identified as major contributors to risk of PWRs. Their relevance lies in the potential radioactive release from reactor coolant system to the environment. Lack of knowledge on the source term attenuation capability of the steam generator has avoided its consideration in probabilistic safety studies and severe accident management guidelines. This paper describes a research program presently under way on the aerosol retention in the nearby of the tube breach within the secondary side of the steam generation in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST program. Experimental activities are focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development. (author)

  10. French PWR nuclear power plants: Probabilistic studies of accident sequences and related findings

    International Nuclear Information System (INIS)

    Villemeur, A.; Moroni, J.M.; Berger, J.P.; Meslin, T.

    1987-01-01

    This paper presents the major studies performed in France by EDF in the framework of probabilistic studies. It describes the part played by these studies especially as regards: the assessment of the allowed outage time in the event of a safety component unavailability, the risk assessment in the event of a total loss of system (heat sink, electric power supplies, etc.). The specific features of the French 'living' PSA, now still in progress, are also presented. (orig./HSCH)

  11. An evaluation of alternate containment concepts for severe accident sequences: Chapter 3

    International Nuclear Information System (INIS)

    Ashton, D.H.; Blazo, S.R.

    1983-01-01

    Over the past several years, numerous design concepts have been developed to enhance the ability of containments to withstand severe reactor accidents. As part of the AIF sponsored IDCOR program, a study has been completed to survey and evaluate these alternate containment design concepts. The study defines the minimum as well as optimum functional and design criteria which any such system must meet. Six concepts which satisfy these criteria are then evaluated based upon factors such as: risk reduction potential, cost, constructability and the potential detrimental effects. Based upon the results of these evaluations, a ranking of the design concepts is developed. The purpose of this paper is to present the results of the IDCOR sponsored study

  12. Containment response and radiological release for a TMLB' accident sequence in a large dry containment

    International Nuclear Information System (INIS)

    Gasser, R.D.; Bieniarz, P.P.; Tills, J.L.

    1987-01-01

    An analysis has been performed for the Bellefonte Pressurized Water Reactor (PWR) Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis, which include the effects of direct heating on containment loading and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating, which involves more than about 50% of the core, may fail the Bellefonte containment, but natural convection in the Reactor Coolant System (RCS) may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach, due to natural circulation, and after vessel breach, due to reevolution of retained fission products by fission product heating of RCS structures. (orig.)

  13. The nature of reactor accidents

    International Nuclear Information System (INIS)

    Domaratzki, Z.; Campbell, F.R.; Atchison, R.J.

    1981-01-01

    Reactor accidents are events which result in the release of radioactive material from a nuclear power plant due to the failure of one or more critical components of that plant. The failures, depending on their number and type, can result in releases whose consequences range from negligible to catastrophic. By way of examples, this paper describes four specific accidents which cover this range of consequence: failure of a reactor control system, loss of coolant, loss of coolant with impaired containment, and reactor core meltdown. For each a possible sequence of events and an estimate of the expected frequency are presented

  14. Experimental Investigation of Operation of VVER Steam Generator in Condensation Mode in the Event of the Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, Andrey [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

    2008-07-01

    For new Russian nuclear power plants with VVER-1200 reactor in the event of a beyond design basis accident, provision is made for the use of passive safety systems for necessary core cooling. These safety systems include the passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from SG arrives at the core providing its additional cooling. To investigate the condensation mode of VVER SG operation, a large scale HA2M-SG test facility was constructed. The rig incorporates: buffer tank, SG model with scale is 1:46, PHRS heat exchanger. Experiments at the test facility have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the test procedure and the basic obtained test results. (authors)

  15. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    International Nuclear Information System (INIS)

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

  16. Benchmark exercises on PWR level-1 PSA (step 3). Analyses of accident sequence and conclusions

    International Nuclear Information System (INIS)

    Niwa, Yuji; Takahashi, Hideaki.

    1996-01-01

    The results of level 1 PSA generate fluctuations due to the assumptions based on several engineering judgements set in the stages of PSA analysis. On the purpose of the investigation of uncertainties due to assumptions, three kinds of a standard problem, what we call benchmark exercise have been set. In this report, sensitivity studies (benchmark exercise) of sequence analyses are treated and conclusions are mentioned. The treatment of inter-system dependency would generate uncertainly of PSA. In addition, as a conclusion of the PSA benchmark exercise, several findings in the sequence analysis together with previous benchmark analyses in earlier INSS Journals are treated. (author)

  17. Hidden Markov event sequence models: toward unsupervised functional MRI brain mapping.

    Science.gov (United States)

    Faisan, Sylvain; Thoraval, Laurent; Armspach, Jean-Paul; Foucher, Jack R; Metz-Lutz, Marie-Noëlle; Heitz, Fabrice

    2005-01-01

    Most methods used in functional MRI (fMRI) brain mapping require restrictive assumptions about the shape and timing of the fMRI signal in activated voxels. Consequently, fMRI data may be partially and misleadingly characterized, leading to suboptimal or invalid inference. To limit these assumptions and to capture the broad range of possible activation patterns, a novel statistical fMRI brain mapping method is proposed. It relies on hidden semi-Markov event sequence models (HSMESMs), a special class of hidden Markov models (HMMs) dedicated to the modeling and analysis of event-based random processes. Activation detection is formulated in terms of time coupling between (1) the observed sequence of hemodynamic response onset (HRO) events detected in the voxel's fMRI signal and (2) the "hidden" sequence of task-induced neural activation onset (NAO) events underlying the HROs. Both event sequences are modeled within a single HSMESM. The resulting brain activation model is trained to automatically detect neural activity embedded in the input fMRI data set under analysis. The data sets considered in this article are threefold: synthetic epoch-related, real epoch-related (auditory lexical processing task), and real event-related (oddball detection task) fMRI data sets. Synthetic data: Activation detection results demonstrate the superiority of the HSMESM mapping method with respect to a standard implementation of the statistical parametric mapping (SPM) approach. They are also very close, sometimes equivalent, to those obtained with an "ideal" implementation of SPM in which the activation patterns synthesized are reused for analysis. The HSMESM method appears clearly insensitive to timing variations of the hemodynamic response and exhibits low sensitivity to fluctuations of its shape (unsustained activation during task). Real epoch-related data: HSMESM activation detection results compete with those obtained with SPM, without requiring any prior definition of the expected

  18. Leave for illness/accident or in the event of illness of a close relative - New medical certificate templates

    CERN Multimedia

    HR department

    2016-01-01

    Medical certificate templates are now available in the Admin e-guide (follow the “Forms and templates” link):    Medical certificate for illness/accident Medical certificate for a medical examination or treatment Medical certificate in the event of illness of a close relative These templates are provided for the convenience of members of the personnel and their use is recommended but not compulsory. Other forms of medical certificates issued by a medical doctor may also be submitted, provided they contain the same items of information as those given in the templates. More information on the applicable rules and on the way leave is managed at CERN can be found in the Admin e-guide web pages. Human Resources department HR.leave@cern.ch

  19. Atmospheric dispersion modelling and the use of radiological data in the event of a nuclear accident overseas

    International Nuclear Information System (INIS)

    ApSimon, H.M.; Simms, K.L.

    1988-02-01

    This report considers what radiological measurements are most useful for use in conjunction with computer simulations based on meteorological data to provide the best possible estimates of areas affected and the likely levels of contamination in the event of a nuclear accident overseas. The context is defined according to the needs at different stages in emergency procedures - before radioactivity reaches the UK, during the period of passage overhead, after passage of the material. The ability to identify localised areas where precipitation has concentrated deposition is emphasized. It is made clear that γ detectors tend to be dominated by local levels of deposited activity and are inadequate to define when radioactivity is passing overhead. Facilities for airborne monitoring are recommended. (author)

  20. Ethical aspects of technogenic catastrophes sequences on the example of the Chernobyl accident

    International Nuclear Information System (INIS)

    Mel'nov, S.B.; Sarana, Yu.V.

    2009-01-01

    It is examined such ethical aspects of technogenic catastrophes sequences on the example of Chernobyl disaster, as violation of individual right to get information about the environment condition, getting the liquidator status, maintenance of all ethical norms while holding of biomedical research on disaster victims, and forming of social-ecological stress. (authors)

  1. Invention principles and levels in the event of a nuclear accident

    International Nuclear Information System (INIS)

    Walmod-Larsen, O.

    1994-01-01

    In order to promote Nordic harmonization of the most likely protective measures to be taken in the case of large nuclear accidents, this report presents the background material needed to make common decisions on sheltering, evacuation and relocation. Brief comments only are also made on iodine prophylaxis and foodstuff restrictions. Viewing the national monetary costs per person for such measures in relation to the income per capita - and in relation to the currency exchange rates of Feb. 1994 - there are by and large no arguments to find for different intervention levels in any of the four countries, DK, NO, FI and SE. As applied α-values (the estimated monetary cost of a man-Sievert) are observed to have a large range, attempts were made to find the economic value of a health detriment. These pointed to the Willingness-To-Pay method, and a pilot project was performed in Denmark. On this basis a set of intervention levels - similar to internationally recommended levels -is proposed. Other factors influencing decisions in emergency situations are discussed. Risk perception, risk communication and psychological factors, as well as the modern decision-aiding tools capable of handling such factors are also described. (au) (47 refs.)

  2. Analysis of events resulting from an accident involving a transport aircraft carrying plutonium oxide

    International Nuclear Information System (INIS)

    Lombard, J.; Hubert, P.; Pages, P.

    1988-03-01

    This study assesses the impact on health of an aircraft accident resulting in the release into the atmosphere of the reprocessing product PuO 2 . The consequences associated with the inhalation of the initial cloud, the passage into suspension of the powder deposited on the ground and the contamination of the food chain were therefore evaluated as a function of the quantity released. It was deduced that the risk of inhalation is by far the greatest. The countermeasures likely to be implemented during emergency action were subjected to analysis. In particular, it appeared that the impact of the first cloud could not really be mitigated but that it was possible to take effective action against the other consequences. Research was undertaken to establish tolerable release quantities which could if necessary be used as acceptance criteria for packaging tests. This indicated that a release in the range 10-100 g would give rise to controllable consequences, at least in a rural environment. The calculations relating to the estimation of the acute toxicity associated with the inhalation of Plutonium and details of the emergency action plan are given in appendix

  3. RECOGNITION OF DRAINAGE TUNNELS DURING GLACIER LAKE OUTBURST EVENTS FROM TERRESTRIAL IMAGE SEQUENCES

    Directory of Open Access Journals (Sweden)

    E. Schwalbe

    2016-06-01

    Full Text Available In recent years, many glaciers all over the world have been distinctly retreating and thinning. One of the consequences of this is the increase of so called glacier lake outburst flood events (GLOFs. The mechanisms ruling such GLOF events are still not yet fully understood by glaciologists. Thus, there is a demand for data and measurements that can help to understand and model the phenomena. Thereby, a main issue is to obtain information about the location and formation of subglacial channels through which some lakes, dammed by a glacier, start to drain. The paper will show how photogrammetric image sequence analysis can be used to collect such data. For the purpose of detecting a subglacial tunnel, a camera has been installed in a pilot study to observe the area of the Colonia Glacier (Northern Patagonian Ice Field where it dams the Lake Cachet II. To verify the hypothesis, that the course of the subglacial tunnel is indicated by irregular surface motion patterns during its collapse, the camera acquired image sequences of the glacier surface during several GLOF events. Applying tracking techniques to these image sequences, surface feature motion trajectories could be obtained for a dense raster of glacier points. Since only a single camera has been used for image sequence acquisition, depth information is required to scale the trajectories. Thus, for scaling and georeferencing of the measurements a GPS-supported photogrammetric network has been measured. The obtained motion fields of the Colonia Glacier deliver information about the glacier’s behaviour before during and after a GLOF event. If the daily vertical glacier motion of the glacier is integrated over a period of several days and projected into a satellite image, the location and shape of the drainage channel underneath the glacier becomes visible. The high temporal resolution of the motion fields may also allows for an analysis of the tunnels dynamic in comparison to the changing

  4. Combinatorial events of insertion sequences and ICE in Gram-negative bacteria.

    Science.gov (United States)

    Toleman, Mark A; Walsh, Timothy R

    2011-09-01

    The emergence of antibiotic and antimicrobial resistance in Gram-negative bacteria is incremental and linked to genetic elements that function in a so-called 'one-ended transposition' manner, including ISEcp1, ISCR elements and Tn3-like transposons. The power of these elements lies in their inability to consistently recognize one of their own terminal sequences, while recognizing more genetically distant surrogate sequences. This has the effect of mobilizing the DNA sequence found adjacent to their initial location. In general, resistance in Gram-negatives is closely linked to a few one-off events. These include the capture of the class 1 integron by a Tn5090-like transposon; the formation of the 3' conserved segment (3'-CS); and the fusion of the ISCR1 element to the 3'-CS. The structures formed by these rare events have been massively amplified and disseminated in Gram-negative bacteria, but hitherto, are rarely found in Gram-positives. Such events dominate current resistance gene acquisition and are instrumental in the construction of large resistance gene islands on chromosomes and plasmids. Similar combinatorial events appear to have occurred between conjugative plasmids and phages constructing hybrid elements called integrative and conjugative elements or conjugative transposons. These elements are beginning to be closely linked to some of the more powerful resistance mechanisms such as the extended spectrum β-lactamases, metallo- and AmpC type β-lactamases. Antibiotic resistance in Gram-negative bacteria is dominated by unusual combinatorial mistakes of Insertion sequences and gene fusions which have been selected and amplified by antibiotic pressure enabling the formation of extended resistance islands. © 2011 Federation of European Microbiological Societies. Published by Blackwell Publishing Ltd. All rights reserved.

  5. An Autonomous System for Grouping Events in a Developing Aftershock Sequence

    Energy Technology Data Exchange (ETDEWEB)

    Harris, D. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dodge, D. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2011-03-22

    We describe a prototype detection framework that automatically clusters events in real time from a rapidly unfolding aftershock sequence. We use the fact that many aftershocks are repetitive, producing similar waveforms. By clustering events based on correlation measures of waveform similarity, the number of independent event instances that must be examined in detail by analysts may be reduced. Our system processes array data and acquires waveform templates with a short-term average (STA)/long-term average (LTA) detector operating on a beam directed at the P phases of the aftershock sequence. The templates are used to create correlation-type (subspace) detectors that sweep the subsequent data stream for occurrences of the same waveform pattern. Events are clustered by association with a particular detector. Hundreds of subspace detectors can run in this framework a hundred times faster than in real time. Nonetheless, to check the growth in the number of detectors, the framework pauses periodically and reclusters detections to reduce the number of event groups. These groups define new subspace detectors that replace the older generation of detectors. Because low-magnitude occurrences of a particular signal template may be missed by the STA/LTA detector, we advocate restarting the framework from the beginning of the sequence periodically to reprocess the entire data stream with the existing detectors. We tested the framework on 10 days of data from the Nevada Seismic Array (NVAR) covering the 2003 San Simeon earthquake. One hundred eighty-four automatically generated detectors produced 676 detections resulting in a potential reduction in analyst workload of up to 73%.

  6. Deepwater Horizon: Experience the Events That Led to This Accident, Follow the Investigation as They Uncover the Human Factors

    International Nuclear Information System (INIS)

    Bannerman, T.

    2016-01-01

    With the Key themes of leadership, culture, reputation and risk, process safety and the human and organizational factors inside partnership and joint ventures, this session run by AKT immerses you into the situation on board the Deepwater Horizon drilling rig in the Gulf of Mexico on the day of the disaster 20 April 2010. The sequence of events are acted out and then we follow the investigation as they uncover negligence, poor regulation, inadequate maintenance, and catastrophic decision making and what the US authorities called “a reckless disregard for safety”. This session will show how this type of workshop event has been used in different organizations, and the actors run the session to show how the facts of the disaster can be used to enhance knowledge of managers and senior leaders of factors that can trigger a major event. (author)

  7. A major sporting event does not necessarily mean an increased workload for accident and emergency departments. Euro96 Group of Accident and Emergency Departments

    OpenAIRE

    Cooke, M. W.; Allan, T. F.; Wilson, S.

    1999-01-01

    AIM: To determine whether there were any changes in attendance at accident and emergency departments that could be related to international football matches (Euro96 tournament). METHOD: Fourteen accident and emergency departments (seven adjacent to and seven distant from a Euro96 venue) provided their daily attendance figures for a nine week period: three weeks before, during, and after the tournament. The relation between daily attendance rates and Euro96 football matches was assessed ...

  8. Potential Indoor Worker Exposure From Handling Area Leakage: Example Event Sequence Frequency Analysis

    International Nuclear Information System (INIS)

    Benke, Roland R.; Adams, George R.

    2008-01-01

    potential event sequences. A hypothetical case is presented for failure of the HVAC exhaust system to provide confinement for contaminated air from otherwise normal operations. This paper presents an example calculation of frequencies for a potential event sequence involving HVAC system failure during otherwise routine wet transfer operations of spent nuclear fuel assemblies from an open container. For the simplified HVAC exhaust system model, the calculation indicated that the potential event sequence may or may not be a Category 1 event sequence, in light of current uncertainties (e.g., final HVAC system design and duration of facility operations). Categorization of potential event sequences is important because different regulatory requirements and performance objectives are specified based on the categorization of event sequences. A companion paper presents a dose calculation methodology and example calculations of indoor worker consequences for the posed example event sequence. Together, the two companion papers demonstrate capabilities for performing confirmatory calculations of frequency and consequence, which may assist the assessment of worker safety during a risk-informed regulatory review of a potential DOE license application

  9. Rare recombination events generate sequence diversity among balancer chromosomes in Drosophila melanogaster.

    Science.gov (United States)

    Miller, Danny E; Cook, Kevin R; Yeganeh Kazemi, Nazanin; Smith, Clarissa B; Cockrell, Alexandria J; Hawley, R Scott; Bergman, Casey M

    2016-03-08

    Multiply inverted balancer chromosomes that suppress exchange with their homologs are an essential part of the Drosophila melanogaster genetic toolkit. Despite their widespread use, the organization of balancer chromosomes has not been characterized at the molecular level, and the degree of sequence variation among copies of balancer chromosomes is unknown. To map inversion breakpoints and study potential diversity in descendants of a structurally identical balancer chromosome, we sequenced a panel of laboratory stocks containing the most widely used X chromosome balancer, First Multiple 7 (FM7). We mapped the locations of FM7 breakpoints to precise euchromatic coordinates and identified the flanking sequence of breakpoints in heterochromatic regions. Analysis of SNP variation revealed megabase-scale blocks of sequence divergence among currently used FM7 stocks. We present evidence that this divergence arose through rare double-crossover events that replaced a female-sterile allele of the singed gene (sn(X2)) on FM7c with a sequence from balanced chromosomes. We propose that although double-crossover events are rare in individual crosses, many FM7c chromosomes in the Bloomington Drosophila Stock Center have lost sn(X2) by this mechanism on a historical timescale. Finally, we characterize the original allele of the Bar gene (B(1)) that is carried on FM7, and validate the hypothesis that the origin and subsequent reversion of the B(1) duplication are mediated by unequal exchange. Our results reject a simple nonrecombining, clonal mode for the laboratory evolution of balancer chromosomes and have implications for how balancer chromosomes should be used in the design and interpretation of genetic experiments in Drosophila.

  10. Sequence Coding and Search System for licensee event reports: coder's manual. Volume 4

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This four volume report documents and describes SCSS in detail. Volume 3 and 4 provide a technical processor, new to SCSS, the information and methodology necessary to capture descriptive data from the LER and to codify that data into a structured format and serve as reference material for the more experienced technical processor, and contains information that is essential for the more advanced user who needs to be familiar with the intricate coding techniques in order to retrieve specific details in a sequence. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 4

  11. Sequence Coding and Search System for licensee event reports: coder's manual. Volume 3

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This four volume report documents and describes SCSS in detail. Volumes 3 and 4 provide a technical processor, new to SCSS, the information and methodology necessary to capture descriptive data from the LER and to codify that data into a structured format and serve as reference material for the more experienced technical processor, and contains information is essential for the more advanced user who needs to be familiar with the intricate coding techniques in order to retrieve specific details in a sequence. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 3

  12. The Flushtration Count Illusion: Attribute substitution tricks our interpretation of a simple visual event sequence.

    Science.gov (United States)

    Thomas, Cyril; Didierjean, André; Kuhn, Gustav

    2018-04-17

    When faced with a difficult question, people sometimes work out an answer to a related, easier question without realizing that a substitution has taken place (e.g., Kahneman, 2011, Thinking, fast and slow. New York, Farrar, Strauss, Giroux). In two experiments, we investigated whether this attribute substitution effect can also affect the interpretation of a simple visual event sequence. We used a magic trick called the 'Flushtration Count Illusion', which involves a technique used by magicians to give the illusion of having seen multiple cards with identical backs, when in fact only the back of one card (the bottom card) is repeatedly shown. In Experiment 1, we demonstrated that most participants are susceptible to the illusion, even if they have the visual and analytical reasoning capacity to correctly process the sequence. In Experiment 2, we demonstrated that participants construct a biased and simplified representation of the Flushtration Count by substituting some attributes of the event sequence. We discussed of the psychological processes underlying this attribute substitution effect. © 2018 The British Psychological Society.

  13. Sequence Coding and Search System for licensee event reports: user's guide. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    Greene, N.M.; Mays, G.T.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. This four volume report documents and describes SCSS in detail. Volume 1 is a User's Guide for searching the SCSS database. This volume contains updated material through February 1985 of the working version of ORNL/NSIC-223, Vol. 1

  14. Sequence Coding and Search System for licensee event reports: code listings. Volume 2

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2

  15. A single-chip event sequencer and related microcontroller instrumentation for atomic physics research.

    Science.gov (United States)

    Eyler, E E

    2011-01-01

    A 16-bit digital event sequencer with 50 ns resolution and 50 ns trigger jitter is implemented by using an internal 32-bit timer on a dsPIC30F4013 microcontroller, controlled by an easily modified program written in standard C. It can accommodate hundreds of output events, and adjacent events can be spaced as closely as 1.5 μs. The microcontroller has robust 5 V inputs and outputs, allowing a direct interface to common laboratory equipment and other electronics. A USB computer interface and a pair of analog ramp outputs can be added with just two additional chips. An optional display/keypad unit allows direct interaction with the sequencer without requiring an external computer. Minor additions also allow simple realizations of other complex instruments, including a precision high-voltage ramp generator for driving spectrum analyzers or piezoelectric positioners, and a low-cost proportional integral differential controller and lock-in amplifier for laser frequency stabilization with about 100 kHz bandwidth.

  16. Chernobylsk accident (Causes and Consequences)- Part 2

    International Nuclear Information System (INIS)

    Esteves, D.

    1986-09-01

    The causes and consequences of the nuclear accident at Chernobylsk-4 reactor are shortly described. The informations were provided by Russian during the specialist meeting, carried out at seat of IAEA. The Russian nuclear panorama; the site, nuclear power plant characteristics and sequence of events; the immediate measurements after accident; monitoring/radioactive releases; environmental contamination and ecological consequences; measurements of emergency; recommendations to increase the nuclear safety; and recommendations of work groups, are presented. (M.C.K.) [pt

  17. ASSET: Analysis of Sequences of Synchronous Events in Massively Parallel Spike Trains

    Science.gov (United States)

    Canova, Carlos; Denker, Michael; Gerstein, George; Helias, Moritz

    2016-01-01

    With the ability to observe the activity from large numbers of neurons simultaneously using modern recording technologies, the chance to identify sub-networks involved in coordinated processing increases. Sequences of synchronous spike events (SSEs) constitute one type of such coordinated spiking that propagates activity in a temporally precise manner. The synfire chain was proposed as one potential model for such network processing. Previous work introduced a method for visualization of SSEs in massively parallel spike trains, based on an intersection matrix that contains in each entry the degree of overlap of active neurons in two corresponding time bins. Repeated SSEs are reflected in the matrix as diagonal structures of high overlap values. The method as such, however, leaves the task of identifying these diagonal structures to visual inspection rather than to a quantitative analysis. Here we present ASSET (Analysis of Sequences of Synchronous EvenTs), an improved, fully automated method which determines diagonal structures in the intersection matrix by a robust mathematical procedure. The method consists of a sequence of steps that i) assess which entries in the matrix potentially belong to a diagonal structure, ii) cluster these entries into individual diagonal structures and iii) determine the neurons composing the associated SSEs. We employ parallel point processes generated by stochastic simulations as test data to demonstrate the performance of the method under a wide range of realistic scenarios, including different types of non-stationarity of the spiking activity and different correlation structures. Finally, the ability of the method to discover SSEs is demonstrated on complex data from large network simulations with embedded synfire chains. Thus, ASSET represents an effective and efficient tool to analyze massively parallel spike data for temporal sequences of synchronous activity. PMID:27420734

  18. A Fast Event Preprocessor and Sequencer for the Simbol-X Low Energy Detector

    Science.gov (United States)

    Schanz, T.; Tenzer, C.; Maier, D.; Kendziorra, E.; Santangelo, A.

    2009-05-01

    The Simbol-X Low Energy Detector (LED), a 128×128 pixel DEPFET (Depleted Field Effect Transistor) array, will be read out at a very high rate (8000 frames/second) and, therefore, requires a very fast on board electronics. We present an FPGA-based LED camera electronics consisting of an Event Preprocessor (EPP) for on board data preprocessing and filtering of the Simbol-X low-energy detector and a related Sequencer (SEQ) to generate the necessary signals to control the readout.

  19. A Fast Event Preprocessor and Sequencer for the Simbol-X Low Energy Detector

    International Nuclear Information System (INIS)

    Schanz, T.; Tenzer, C.; Maier, D.; Kendziorra, E.; Santangelo, A.

    2009-01-01

    The Simbol-X Low Energy Detector (LED), a 128x128 pixel DEPFET (Depleted Field Effect Transistor) array, will be read out at a very high rate (8000 frames/second) and, therefore, requires a very fast on board electronics. We present an FPGA-based LED camera electronics consisting of an Event Preprocessor (EPP) for on board data preprocessing and filtering of the Simbol-X low-energy detector and a related Sequencer (SEQ) to generate the necessary signals to control the readout.

  20. The influence of spherical cavity surface charge distribution on the sequence of partial discharge events

    International Nuclear Information System (INIS)

    Illias, Hazlee A; Chen, George; Lewin, Paul L

    2011-01-01

    In this work, a model representing partial discharge (PD) behaviour of a spherical cavity within a homogeneous dielectric material has been developed to study the influence of cavity surface charge distribution on the electric field distribution in both the cavity and the material itself. The charge accumulation on the cavity surface after a PD event and charge movement along the cavity wall under the influence of electric field magnitude and direction has been found to affect the electric field distribution in the whole cavity and in the material. This in turn affects the likelihood of any subsequent PD activity in the cavity and the whole sequence of PD events. The model parameters influencing cavity surface charge distribution can be readily identified; they are the cavity surface conductivity, the inception field and the extinction field. Comparison of measurement and simulation results has been undertaken to validate the model.

  1. The influence of spherical cavity surface charge distribution on the sequence of partial discharge events

    Energy Technology Data Exchange (ETDEWEB)

    Illias, Hazlee A [Department of Electrical Engineering, Faculty of Engineering, University of Malaya, 50603 Kuala Lumpur (Malaysia); Chen, George; Lewin, Paul L, E-mail: h.illias@um.edu.my [Tony Davies High Voltage Laboratory, School of Electronics and Computer Science, University of Southampton, Southampton, SO17 1BJ (United Kingdom)

    2011-06-22

    In this work, a model representing partial discharge (PD) behaviour of a spherical cavity within a homogeneous dielectric material has been developed to study the influence of cavity surface charge distribution on the electric field distribution in both the cavity and the material itself. The charge accumulation on the cavity surface after a PD event and charge movement along the cavity wall under the influence of electric field magnitude and direction has been found to affect the electric field distribution in the whole cavity and in the material. This in turn affects the likelihood of any subsequent PD activity in the cavity and the whole sequence of PD events. The model parameters influencing cavity surface charge distribution can be readily identified; they are the cavity surface conductivity, the inception field and the extinction field. Comparison of measurement and simulation results has been undertaken to validate the model.

  2. CNE (Embalse nuclear power plant): probabilistic safety study. Electric power supply. Events sequence

    International Nuclear Information System (INIS)

    Figueroa, N.

    1987-01-01

    The plant response to the occurrence of the starting event 'total loss of electric power supply to class IV and class III' is analyzed. This involves the study of automatical actions of safety and process systems as well as the operator actions. The probabilistic evaluation of starting event frequency is performed through fault-tree techniques. The frequency of occurrence 'loss of electric power supply to class IV (λIV = 0.56/year) and the probability of failure to demand of 'reserve' generating groups (Pd III 6.79 x 10 -3 ) contribute to the mentioned frequency. As soon as the starting event occurs, the reactor power must be reduced to 0%, the fuel must be cooled through the thermo siphon and decay heat has to be removed. The events sequence analysis leads to the conclusion that the non shutting down of the reactor with any of the shutdown systems is 'incredible' (10 -6 /year). In all cases the fuel is cooled by building the thermo siphon except when a substantial inventory loss exist due to a closure failure of some valve of pressure and inventory control system. The order of magnitude of the failure of decay heat removal through the steam generators is 4 x 10 -4 . This removal would be assured by the emergency water system. Therefore, the frequency of the sequence of possible core meltdown, when the reactor does not shut down is: λ = 5 x 10 -9 /year and for the failure of heat removal: λ = 2 x 10 -6 /year. (Author)

  3. A Macro-Observation Scheme for Abnormal Event Detection in Daily-Life Video Sequences

    Directory of Open Access Journals (Sweden)

    Chiu Wei-Yao

    2010-01-01

    Full Text Available Abstract We propose a macro-observation scheme for abnormal event detection in daily life. The proposed macro-observation representation records the time-space energy of motions of all moving objects in a scene without segmenting individual object parts. The energy history of each pixel in the scene is instantly updated with exponential weights without explicitly specifying the duration of each activity. Since possible activities in daily life are numerous and distinct from each other and not all abnormal events can be foreseen, images from a video sequence that spans sufficient repetition of normal day-to-day activities are first randomly sampled. A constrained clustering model is proposed to partition the sampled images into groups. The new observed event that has distinct distance from any of the cluster centroids is then classified as an anomaly. The proposed method has been evaluated in daily work of a laboratory and BEHAVE benchmark dataset. The experimental results reveal that it can well detect abnormal events such as burglary and fighting as long as they last for a sufficient duration of time. The proposed method can be used as a support system for the scene that requires full time monitoring personnel.

  4. Synthesis of public authorities organisation in case of emergency and in a post-event situation (following a nuclear accident or a radiological attack) in France and abroad

    International Nuclear Information System (INIS)

    Kayser, O.

    2010-01-01

    After having briefly recalled how an emergency situation (notably in case of nuclear accident or radiological attack) is taken into account in the organisation of public authorities through specific plans (PPI or plans particuliers d'intervention, intervention specific plans), this report also describes how the situation is handled by these authorities after the end of the emergency situation (i.e. when the risk of new radioactive releases is over). This post-event stage is split into two phases: a transition phase which lasts several weeks or months, and a long term consequence management phase (over months or years). The author first describes the specificities of a nuclear or radiological event (accident or attack). He recalls the global public organisation and the involved actors. For the post-event period, he indicates the various actions, describes the interdepartmental coordination and the various aspects of the program designed to manage accident consequences on the long term. He also describes the roles of permanent bodies, agencies and institutes (ASN, ASND, MSNR, IRSN, INVS, ADEME, AFSSA, Meteo France, CEA, ANDRA, AREVA, EDF, ministries). The last part describes the action of public authorities in case of a nuclear accident occurring abroad. This includes relationship with European and international bodies

  5. Damage of reactor buildings occurred at the Fukushima Daiichi accident. Focusing on sequence leading to hydrogen explosions

    International Nuclear Information System (INIS)

    Naito, Masanori

    2011-01-01

    Fukushima Daiichi accident discharged enormous radioactive materials confined inside into the environment due to hydrogen explosions occurred at reactor buildings and forced many people to live the refugee life. This article described overview of Great East Japan Earthquake, specifications of Fukushima Daiichi nuclear power plants, sequence of plant status after earthquake occurrence and computerized simulation of plant behavior of Unit 1 leading to core melt and hydrogen explosion. Simulation results with estimated and assumed conditions showed water level decreased to bottom of reactor core after 4 hrs and 15 minutes passed, core melt started after 6 hrs and 49 minutes passed, failure of core support plate after 7 hrs and 18 minutes passed and through failure of penetration at bottom of pressure vessel after 7 hrs and 25 minutes passed. Hydrogen concentration at operating floor of reactor building of Unit 1 would be 15% accumulated and the pressure would amount to about 5 bars after hydrogen explosion if reactor building did not rupture with leak-tight structure. Since reactor building was not pressure-proof structure, walls of operating floor would rupture before 5 bars attained. (T. Tanaka)

  6. Differential beta-band event-related desynchronization during categorical action sequence planning.

    Directory of Open Access Journals (Sweden)

    Hame Park

    Full Text Available A primate study reported the existence of neurons from the dorso-lateral prefrontal cortex which fired prior to executing categorical action sequences. The authors suggested these activities may represent abstract level information. Here, we aimed to find the neurophysiological representation of planning categorical action sequences at the population level in healthy humans. Previous human studies have shown beta-band event-related desynchronization (ERD during action planning in humans. Some of these studies showed different levels of ERD according to different types of action preparation. Especially, the literature suggests that variations in cognitive factors rather than physical factors (force, direction, etc modulate the level of beta-ERD. We hypothesized that the level of beta-band power will differ according to planning of different categorical sequences. We measured magnetoencephalography (MEG from 22 subjects performing 11 four-sequence actions--each consisting of one or two of three simple actions--in 3 categories; 'Paired (ooxx', 'Alternative (oxox' and 'Repetitive (oooo' ('o' and 'x' each denoting one of three simple actions. Time-frequency representations were calculated for each category during the planning period, and the corresponding beta-power time-courses were compared. We found beta-ERD during the planning period for all subjects, mostly in the contralateral fronto-parietal areas shortly after visual cue onset. Power increase (transient rebound followed ERD in 20 out of 22 subjects. Amplitudes differed among categories in 20 subjects for both ERD and transient rebound. In 18 out of 20 subjects 'Repetitive' category showed the largest ERD and rebound. The current result suggests that beta-ERD in the contralateral frontal/motor/parietal areas during planning is differentiated by the category of action sequences.

  7. A study for the sequence of events (SOE) system on the nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chae; Jeon, Jong Sun; Lee, Sun Sung; Lee, Kyung Ho; Lee, Byung Ju; Sohn, Kwang Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    It is important to identify where and why an event or a trip is occurred in the Nuclear Power Plant(NPP) and to provide proper resolution against above situation. In order to analyze the prime cause or conspicuous reason of trouble occurred after events or trips occur, the Sequence of Events(SOE) system has been adopted in Korean NPP to acquire the sequential information along where and when an event or a trip take place. The SOE system of UCN 3 and 4 plant which is included in the Plant Data Acquisition System (PDAS), shares the 3205 computer and system software with PDAS. Sharing of the computer H/w and S/W, however, requires more complicated process to provide the events or trip signals due to the inherent characteristics of the shared system. Moreover there are high potentiality of collision between synchronization signals and data transmitted to the Plant Computer System (PCS), when the synchronization signals are sent from PCS to the three SOE processors. When this collision happens the SOE system will break down, thus it is not possible to analyze the trend of events or trips. An independent SOE system composed with single processor is proposed in this paper. To begin with, the analyses on the hardware and software of SOE and PDAS system of UCN 3 and 4 were performed to justify the problems and the resolution if it exists. In order to test the new SOE system, VMEbus, VM30 CPU, change of status I/O card and OS-9 for the operating system were adopted and the analysis for this test system was done as follows; the verification should be achieved through the simulation; the simulated signals for events are given the test system as inputs and the outputs are monitored to verify whether the sequential events logging function works well or not on PC. In conclusion, this report is expected to provide the technical background for the improvement and changing of the NPP PDAS and SOE system in the future. 18 tabs., 33 figs., 26 refs. (Author) .new.

  8. A study for the sequence of events (SOE) system on the nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Byung Chae; Jeon, Jong Sun; Lee, Sun Sung; Lee, Kyung Ho; Lee, Byung Ju; Sohn, Kwang Young

    1996-06-01

    It is important to identify where and why an event or a trip is occurred in the Nuclear Power Plant(NPP) and to provide proper resolution against above situation. In order to analyze the prime cause or conspicuous reason of trouble occurred after events or trips occur, the Sequence of Events(SOE) system has been adopted in Korean NPP to acquire the sequential information along where and when an event or a trip take place. The SOE system of UCN 3 and 4 plant which is included in the Plant Data Acquisition System (PDAS), shares the 3205 computer and system software with PDAS. Sharing of the computer H/w and S/W, however, requires more complicated process to provide the events or trip signals due to the inherent characteristics of the shared system. Moreover there are high potentiality of collision between synchronization signals and data transmitted to the Plant Computer System (PCS), when the synchronization signals are sent from PCS to the three SOE processors. When this collision happens the SOE system will break down, thus it is not possible to analyze the trend of events or trips. An independent SOE system composed with single processor is proposed in this paper. To begin with, the analyses on the hardware and software of SOE and PDAS system of UCN 3 and 4 were performed to justify the problems and the resolution if it exists. In order to test the new SOE system, VMEbus, VM30 CPU, change of status I/O card and OS-9 for the operating system were adopted and the analysis for this test system was done as follows; the verification should be achieved through the simulation; the simulated signals for events are given the test system as inputs and the outputs are monitored to verify whether the sequential events logging function works well or not on PC. In conclusion, this report is expected to provide the technical background for the improvement and changing of the NPP PDAS and SOE system in the future. 18 tabs., 33 figs., 26 refs. (Author) .new

  9. The Mw=8.8 Maule earthquake aftershock sequence, event catalog and locations

    Science.gov (United States)

    Meltzer, A.; Benz, H.; Brown, L.; Russo, R. M.; Beck, S. L.; Roecker, S. W.

    2011-12-01

    The aftershock sequence of the Mw=8.8 Maule earthquake off the coast of Chile in February 2010 is one of the most well-recorded aftershock sequences from a great megathrust earthquake. Immediately following the Maule earthquake, teams of geophysicists from Chile, France, Germany, Great Britain and the United States coordinated resources to capture aftershocks and other seismic signals associated with this significant earthquake. In total, 91 broadband, 48 short period, and 25 accelerometers stations were deployed above the rupture zone of the main shock from 33-38.5°S and from the coast to the Andean range front. In order to integrate these data into a unified catalog, the USGS National Earthquake Information Center develop procedures to use their real-time seismic monitoring system (Bulletin Hydra) to detect, associate, location and compute earthquake source parameters from these stations. As a first step in the process, the USGS has built a seismic catalog of all M3.5 or larger earthquakes for the time period of the main aftershock deployment from March 2010-October 2010. The catalog includes earthquake locations, magnitudes (Ml, Mb, Mb_BB, Ms, Ms_BB, Ms_VX, Mc), associated phase readings and regional moment tensor solutions for most of the M4 or larger events. Also included in the catalog are teleseismic phases and amplitude measures and body-wave MT and CMT solutions for the larger events, typically M5.5 and larger. Tuning of automated detection and association parameters should allow a complete catalog of events to approximately M2.5 or larger for that dataset of more than 164 stations. We characterize the aftershock sequence in terms of magnitude, frequency, and location over time. Using the catalog locations and travel times as a starting point we use double difference techniques to investigate relative locations and earthquake clustering. In addition, phase data from candidate ground truth events and modeling of surface waves can be used to calibrate the

  10. Temporal and spatial predictability of an irrelevant event differently affect detection and memory of items in a visual sequence

    Directory of Open Access Journals (Sweden)

    Junji eOhyama

    2016-02-01

    Full Text Available We examined how the temporal and spatial predictability of a task-irrelevant visual event affects the detection and memory of a visual item embedded in a continuously changing sequence. Participants observed 11 sequentially presented letters, during which a task-irrelevant visual event was either present or absent. Predictabilities of spatial location and temporal position of the event were controlled in 2 × 2 conditions. In the spatially predictable conditions, the event occurred at the same location within the stimulus sequence or at another location, while, in the spatially unpredictable conditions, it occurred at random locations. In the temporally predictable conditions, the event timing was fixed relative to the order of the letters, while in the temporally unpredictable condition, it could not be predicted from the letter order. Participants performed a working memory task and a target detection reaction time task. Memory accuracy was higher for a letter simultaneously presented at the same location as the event in the temporally unpredictable conditions, irrespective of the spatial predictability of the event. On the other hand, the detection reaction times were only faster for a letter simultaneously presented at the same location as the event when the event was both temporally and spatially predictable. Thus, to facilitate ongoing detection processes, an event must be predictable both in space and time, while memory processes are enhanced by temporally unpredictable (i.e., surprising events. Evidently, temporal predictability has differential effects on detection and memory of a visual item embedded in a sequence of images.

  11. High quality maize centromere 10 sequence reveals evidence of frequent recombination events

    Directory of Open Access Journals (Sweden)

    Thomas Kai Wolfgruber

    2016-03-01

    Full Text Available The ancestral centromeres of maize contain long stretches of the tandemly arranged CentC repeat. The abundance of tandem DNA repeats and centromeric retrotransposons (CR have presented a significant challenge to completely assembling centromeres using traditional sequencing methods. Here we report a nearly complete assembly of the 1.85 Mb maize centromere 10 from inbred B73 using PacBio technology and BACs from the reference genome project. The error rates estimated from overlapping BAC sequences are 7 x 10-6 and 5 x 10-5 for mismatches and indels, respectively. The number of gaps in the region covered by the reassembly was reduced from 140 in the reference genome to three. Three expressed genes are located between 92 and 477 kb of the inferred ancestral CentC cluster, which lies within the region of highest centromeric repeat density. The improved assembly increased the count of full-length centromeric retrotransposons from 5 to 55 and revealed a 22.7 kb segmental duplication that occurred approximately 121,000 years ago. Our analysis provides evidence of frequent recombination events in the form of partial retrotransposons, deletions within retrotransposons, chimeric retrotransposons, segmental duplications including higher order CentC repeats, a deleted CentC monomer, centromere-proximal inversions, and insertion of mitochondrial sequences. Double-strand DNA break (DSB repair is the most plausible mechanism for these events and may be the major driver of centromere repeat evolution and diversity. This repair appears to be mediated by microhomology, suggesting that tandem repeats may have evolved to facilitate the repair of frequent DSBs in centromeres.

  12. Aftershock Sequences and Seismic-Like Organization of Acoustic Events Produced by a Single Propagating Crack

    Science.gov (United States)

    Alizee, D.; Bonamy, D.

    2017-12-01

    In inhomogeneous brittle solids like rocks, concrete or ceramics, one usually distinguish nominally brittle fracture, driven by the propagation of a single crack from quasibrittle one, resulting from the accumulation of many microcracks. The latter goes along with intermittent sharp noise, as e.g. revealed by the acoustic emission observed in lab scale compressive fracture experiments or at geophysical scale in the seismic activity. In both cases, statistical analyses have revealed a complex time-energy organization into aftershock sequences obeying a range of robust empirical scaling laws (the Omori-Utsu, productivity and Bath's law) that help carry out seismic hazard analysis and damage mitigation. These laws are usually conjectured to emerge from the collective dynamics of microcrack nucleation. In the experiments presented at AGU, we will show that such a statistical organization is not specific to the quasi-brittle multicracking situations, but also rules the acoustic events produced by a single crack slowly driven in an artificial rock made of sintered polymer beads. This simpler situation has advantageous properties (statistical stationarity in particular) permitting us to uncover the origins of these seismic laws: Both productivity law and Bath's law result from the scale free statistics for event energy and Omori-Utsu law results from the scale-free statistics of inter-event time. This yields predictions on how the associated parameters are related, which were analytically derived. Surprisingly, the so-obtained relations are also compatible with observations on lab scale compressive fracture experiments, suggesting that, in these complex multicracking situations also, the organization into aftershock sequences and associated seismic laws are also ruled by the propagation of individual microcrack fronts, and not by the collective, stress-mediated, microcrack nucleation. Conversely, the relations are not fulfilled in seismology signals, suggesting that

  13. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  14. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 2, Part 1C: Analysis of core damage frequency from internal events for plant operational State 5 during a refueling outage, Main report (Sections 11--14)

    International Nuclear Information System (INIS)

    Whitehead, D.; Darby, J.; Yakle, J.

    1994-06-01

    This document contains the accident sequence analysis of internally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 during a refueling outage. The report documents the methodology used during the analysis, describes the results from the application of the methodology, and compares the results with the results from two full power analyses performed on Grand Gulf

  15. Evaluation of potential severe accidents during Low Power and Shutdown Operations at Grand Gulf, Unit 1. Volume 2, Part 1B: Analysis of core damage frequency from internal events for Plant Operational State 5 during a refueling outage, Main report (Section 10)

    International Nuclear Information System (INIS)

    Whitehead, D.; Darby, J.; Yakle, J.

    1994-06-01

    This document contains the accident sequence analysis of internally initiated events for Grand Gulf, Unit 1 as it operates in the Low Power and Shutdown Plant Operational State 5 during a refueling outage. The report documents the methodology used during the analysis, describes the results from the application of the methodology, and compares the results with the results from two full power performed on Grand Gulf. This document, Volume 2, Part 1B, presents chapters Section 10 of this report, Human Reliability Analysis

  16. The 1986 Chernobyl accident; Der Unfall von Tschernobyl 1986

    Energy Technology Data Exchange (ETDEWEB)

    Kerner, Alexander; Stueck, Reinhard; Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching bei Muenchen, Koeln (Germany). Bereich Reaktorsicherheitsanalysen; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany)

    2011-02-15

    April 26, 2011 marks the 25th anniversary of the Chernobyl reactor accident, the worst incident in the history of the peaceful utilization of nuclear power. While investigations of the course of events and the causes of the accident largely present a uniform picture, descriptions still vary widely when it comes to the impact on the population and the environment. This treatment of the Chernobyl accident constitutes a summary of facts about the initiation of the accident and the sequence of events that followed. In addition, measures are described which were taken to exclude any repetition of a disaster of this kind. The health consequences and the socio-economic impact of the accident are not discussed in any detail. The first section contains an introduction and an overview of the Soviet RBMK (Chernobyl) reactor line. In section 2, fundamental characteristics of this special type of reactor, which was exclusively built in the former Soviet Union, are discussed. This information is necessary to understand the sequence of accident events and provides an answer to the frequent question whether that accident could be transferred to reactors in this country. The third section outlines the history of the accident caused ultimately by a commissioning test never performed before. The section is completed by a brief description of radiological releases and the state of the plant after the accident when entombed in the ''sarcophagus.'' The different causes are then summarized and the modifications afterwards made to RBMK reactors are outlined. (orig.)

  17. German Phase B [risk study] highlights the role of [reactor] accident management

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Phase B of the German probabilistic risk assessment study, now scheduled for publication this month, suggests that reactor accident management measures can prevent or mitigate about 90 per cent of event sequences. (author)

  18. Accident of the Fukushima-Daiichi nuclear power station. Situation two years after the event - IRSN file

    International Nuclear Information System (INIS)

    2013-03-01

    Two years after the Fukushima accident, this report proposes a review of the situation in Japan, and of the European and international actions aimed at preventing the occurrence of another nuclear accident and its radiological consequences. It is based on information available at the end of January or February 2013. After a recall of the situation in Japan and Europe in 2011 (recall of the accident, of the different simulation, calculation and information actions undertaken by the IRSN, launching of a program of additional safety assessments and of European stress tests), the report addresses the situation in Japan two years after the accident: evolution of the nuclear risk management governance, status of the Fukushima-Daiichi power station, health and environmental impact and management of the post-accidental phase, actions undertaken by the IRSN (dose assessment, cooperation in the field of severe accidents, participation to the Fukushima Dialogue). The next part details the contribution of the IRSN to the strengthening of safety and radiation protection at the international level (in relationship with international organizations: IAEA, UNSCEAR and WHO). Additional technical information is provided in appendix, as well as a report on the environmental impact of the accident, and a report on the post-accidental management of the accident

  19. Radiation protection. Measures in the event of a serious accident at a nuclear plant. Strahlenschutz. Massnahmen im Falle eines groesseren Unfalls in einer kerntechnischen Anlage

    Energy Technology Data Exchange (ETDEWEB)

    Riedel, W. (Klinikum Steglitz, Berlin (Germany))

    1990-04-01

    The report presents a fictitious scenarium conceived for the event of a release of radioactive substances in connection with a reactor accident, taking into account the geographic situation of Berlin. Depending on the dose to be expected, specific countermeasures of an organizational, administrative, and measurement-technical nature must be instituted. These include, inter alia, the formation of a group of experts, and measurements of the whole body, the skin and/or thyroid gland as a basis for the planned countermeasures. (DG).

  20. Radiation accidents

    International Nuclear Information System (INIS)

    Nenot, J.C.

    1996-01-01

    Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

  1. Report of the US Department of Energy's team analyses of the Chernobyl-4 Atomic Energy Station accident sequence

    International Nuclear Information System (INIS)

    1986-11-01

    In an effort to better understand the Chernobyl-4 accident of April 26, 1986, the US Department of Energy (DOE) formed a team of experts from the National Laboratories including Argonne National Laboratory, Brookhaven National Laboratory, Oak Ridge National Laboratory, and Pacific Northwest Laboratory. The DOE Team provided the analytical support to the US delegation for the August meeting of the International Atomic Energy Agency (IAEA), and to subsequent international meetings. The DOE Team has analyzed the accident in detail, assessed the plausibility and completeness of the information provided by the Soviets, and performed studies relevant to understanding the accident. The results of these studies are presented in this report

  2. Local sequence alignments statistics: deviations from Gumbel statistics in the rare-event tail

    Directory of Open Access Journals (Sweden)

    Burghardt Bernd

    2007-07-01

    Full Text Available Abstract Background The optimal score for ungapped local alignments of infinitely long random sequences is known to follow a Gumbel extreme value distribution. Less is known about the important case, where gaps are allowed. For this case, the distribution is only known empirically in the high-probability region, which is biologically less relevant. Results We provide a method to obtain numerically the biologically relevant rare-event tail of the distribution. The method, which has been outlined in an earlier work, is based on generating the sequences with a parametrized probability distribution, which is biased with respect to the original biological one, in the framework of Metropolis Coupled Markov Chain Monte Carlo. Here, we first present the approach in detail and evaluate the convergence of the algorithm by considering a simple test case. In the earlier work, the method was just applied to one single example case. Therefore, we consider here a large set of parameters: We study the distributions for protein alignment with different substitution matrices (BLOSUM62 and PAM250 and affine gap costs with different parameter values. In the logarithmic phase (large gap costs it was previously assumed that the Gumbel form still holds, hence the Gumbel distribution is usually used when evaluating p-values in databases. Here we show that for all cases, provided that the sequences are not too long (L > 400, a "modified" Gumbel distribution, i.e. a Gumbel distribution with an additional Gaussian factor is suitable to describe the data. We also provide a "scaling analysis" of the parameters used in the modified Gumbel distribution. Furthermore, via a comparison with BLAST parameters, we show that significance estimations change considerably when using the true distributions as presented here. Finally, we study also the distribution of the sum statistics of the k best alignments. Conclusion Our results show that the statistics of gapped and ungapped local

  3. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (3) Identification of dominant factors in transition phase of unprotected events

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Yamano, Hidemasa; Sato, Ikken

    2009-01-01

    The event progression of the transition phase in the unprotected loss of flow accident of the JSFR design concept was analyzed using the SIMMER-III code reflecting the knowledge obtained from the EAGLE experimental program. It was clarified through the parametric calculations that the fuel discharge behavior through the paths such as the inner duct of modified-FAIDUS and control-rod guide tube is playing a very important role. Effective fuel discharge through these paths prevents possibility of severe recriticality events. Important factors dominating the transition phase were identified through these parametric calculations. (author)

  4. SWR-1000 concept on control of severe accidents

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1998-01-01

    It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

  5. The nuclear accidents: Causes and consequences

    International Nuclear Information System (INIS)

    Rochd, M.

    1988-01-01

    The author discussed and compared the real causes of T.M.I. and Chernobyl accidents and cited their consequences. To better understand how these accidents occurred, a brief description of PWR type (reactor type of T.M.I.) and of RBMK type (reactor type of Chernobyl) has been presented. The author has also set out briefly the safety analysis objectives and the three barriers established to protect the public against the radiological consequences. To distinguish failures that cause severe accidents and to analyze them in details, it is necessary to classify the accidents. There are many ways to do it according to their initiator event, or to their frequency, or to their degree of gravity. The safety criteria adopted by nuclear industry have been explained. These criteria specify the limits of certain physical parameters that should not be exceeded in case of incidents or accidents. To compare the real causes of T.M.I. and Chernobyl accidents, the events that led to both have been presented. As observed the main common contributing factors in both cases are that the operators did not pay attention to warnings and signals that were available to them and that they were not trained to handle these accident sequences. The essential conclusions derived from these severe accidents are: -The improvement of operators competence contribute to reduce the accident risks; -The rapid and correct diagnosis of real conditions at each point of the accidents permits an appropriate behavior that would bring the plant to a stable state; -Competent technical teams have to intervene and to assist the operators in case of emergency; -Emergency plans and an international collaboration are necessary to limit the accident risks. 11 figs. (author)

  6. Discussion of the concept of safety indicators from the point of view of TfUX2 accident sequence for Forsmark 3

    International Nuclear Information System (INIS)

    Bujor, A.

    1991-01-01

    This paper contains general considerations on the safety indicators, with details at the system level and for the operator actions. For the system analysis, a modular analysis at a low detailed level is proposed (Module System Approach) in order to emphasize the safety related aspects at the subsystem (module) level. The operator actions are divided in ''active actions'' (actions in the control room during incident/accident situations) and ''passive actions'' (actions during tests, maintenance, repairs, etc.) and are analysed separately. In the second part, a discussion of a possible way to apply some SI to the TfUX2 accident sequence for FORSMARK-3, is done. For the analysis of the Auxiliary Feedwater Systems (AFWS) an equation is proposed to derive target values for the failure probability on demand at the train level, given the target value at the system level, including the common cause failures between the redundant trains. (author) 6 tabs., 18 refs

  7. Mapping the sequence of brain events in response to disgusting food.

    Science.gov (United States)

    Pujol, Jesus; Blanco-Hinojo, Laura; Coronas, Ramón; Esteba-Castillo, Susanna; Rigla, Mercedes; Martínez-Vilavella, Gerard; Deus, Joan; Novell, Ramón; Caixàs, Assumpta

    2018-01-01

    Warning signals indicating that a food is potentially dangerous may evoke a response that is not limited to the feeling of disgust. We investigated the sequence of brain events in response to visual representations of disgusting food using a dynamic image analysis. Functional MRI was acquired in 30 healthy subjects while they were watching a movie showing disgusting food scenes interspersed with the scenes of appetizing food. Imaging analysis included the identification of the global brain response and the generation of frame-by-frame activation maps at the temporal resolution of 2 s. Robust activations were identified in brain structures conventionally associated with the experience of disgust, but our analysis also captured a variety of other brain elements showing distinct temporal evolutions. The earliest events included transient changes in the orbitofrontal cortex and visual areas, followed by a more durable engagement of the periaqueductal gray, a pivotal element in the mediation of responses to threat. A subsequent core phase was characterized by the activation of subcortical and cortical structures directly concerned not only with the emotional dimension of disgust (e.g., amygdala-hippocampus, insula), but also with the regulation of food intake (e.g., hypothalamus). In a later phase, neural excitement extended to broad cortical areas, the thalamus and cerebellum, and finally to the default mode network that signaled the progressive termination of the evoked response. The response to disgusting food representations is not limited to the emotional domain of disgust, and may sequentially involve a variety of broadly distributed brain networks. Hum Brain Mapp 39:369-380, 2018. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

  8. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  9. ANALYSIS OF LABOUR ACCIDENTS DUE TO ROCK FALL EVENTS IN CUTTING FACE OF TUNNEL AND STUDY OF THE COUNTERMEASURES FOR SAFETY

    Science.gov (United States)

    Kikkawa, Naotaka; Itoh, Kazuya; Hori, Tomohito; Tamate, Satoshi; Toyosawa, Yasuo

    In this paper, we analysed the labour accidents which had casualties due to rock fall events in the headings of tunnel and cleared the condition of the occurrence. It was clearly revealed that the accidents mostly happened when the workers mounted the explosive and the steel arch in the headings of the mountain tunnel. In addition, the dimension of the rocks fallen were averagely 0.6m diameter, it was not so much large. Therefore, the countermeasures based on both soft and hard faces would be useful and effective, such as the displacement measurement of a cutting face of tunnel, securing the sufficient lights to observe the cutting face, boring for drainage and shotcreting in a heading of tunnel.

  10. Simulation of LOF accidents with directly electrical heated UO2 pins

    International Nuclear Information System (INIS)

    Alexas, A.

    1976-01-01

    The behavior of directly electrical heated UO 2 pins has been investigated under loss of coolant conditions. Two types of hypothetical accidents have been simulated, first, a LOF accident without power excursion (LOF accident) and second, a LOF accident with subsequent power excursion (LOF-TOP accident). A high-speed film shows the sequence of events for two characteristic experiments. In consequence of the high-speed film analysis as well as the metallographical evaluation statements are given in respect to the cladding meltdown process, the fuel melt fraction and the energy input from the beginning of a power transient to the beginning of the molten fuel ejections

  11. The OECD/NEA workshop on the indemnification of nuclear damage in the event of a nuclear accident

    International Nuclear Information System (INIS)

    Wagstaff, F.

    2002-01-01

    Since 1993, the OECD Nuclear Energy Agency (OECD/NEA) has run the International Nuclear Emergency Exercise (INEX) Program. The program serves to discuss an effective accident management approach on the basis of a simulated nuclear accident situation together with the states involved and their institutions, and also elaborate measures for its further improvement. At the present time, the INEX Program has reached Phase 3 in which, for the first time, also aspects of liability for the consequences of accidents were included. These aspects were made the subject of a workshop held after an emergency exercise. The scenario covered was based on an INES level-4 accident in the French Gravelines Nuclear Power Station situated close to the French-Belgian border. The workshop dealt with these topics, among others: the application of the Paris Convention on Third Party Liability, the Brussels Supplementary Convention, and the Vienna Convention on Civil Liability for Nuclear Damage as well as the Supplementary Compensation Convention of 1997. It was seen that there was a clear need for further discussion, especially to shed more light on the interrelationship of these treaties. (orig.) [de

  12. Precursors to potential severe core damage accidents: 1992, a status report

    International Nuclear Information System (INIS)

    1993-12-01

    This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; open-quote interesting close-quote events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports

  13. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

    A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

  14. Causal Analysis to a Subway Accident: A Comparison of STAMP and RAIB

    Directory of Open Access Journals (Sweden)

    Zhou Yao

    2018-01-01

    Full Text Available Accident investigation and analysis after the accident, vital to prevent the occurrence of similar accident and improve the safety of the system. Different methods led to a different understanding of the accident. In this paper, a subway accident was analysed with a systemic accident analysis model – STAMP (System-Theoretic Accident Modelling and Processes. The hierarchical safety control structure was obtained, and the system-level safety constraints were obtained, controllers of the physical layer were analysed one by one, and put forward the relevant safety requirements and constraints, the dynamic analysis of the structure of the safety control is carried out, and the targeted recommendations are pointed out. In comparison with the analysis results obtained by the Rail Accident Investigation Branch (RAIB. Some useful findings have been concluded. STAMP treats safety as a control problem and reduces or eliminates causes of the accident from the controlling perspective. Whereas RAIB obtains causes of the accident by analysing the sequence of events related to the accident and reasons of these events, then chooses one(or moreevent(s as the immediate cause and some of the key events as causal factors. RAIB analysis is based on the sequential event models, but STAMP analysis provides us with a holistic, dynamic way to control system to maintain safety.

  15. Radiological consequences of a bounding event sequence of Advanced Fusion Neutron Source (A-FNS)

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Makoto M., E-mail: nakamura.makoto@qst.go.jp; Ochiai, Kentaro

    2017-05-15

    Advanced Fusion Neutron Source (A-FNS) is an accelerator-based neutron source utilizing Li(d,xn) nuclear stripping reactions to simulate D-T fusion neutrons for testing and qualifying structural and functional materials of fusion reactor components, which is to be constructed at the Rokkasho site of National Institutes for Quantum and Radiological Science and Technology, Japan, in the near future. The purpose of the study reported here is to demonstrate the ultimate safety margins of A-FNS in the worst case of release of radioactive materials outside the A-FNS confinement system. For this purpose, we analyzed a ‘bounding event’ postulated in A-FNS. The postulated event sequence consists of fire of the purification system of the liquid Li loop during the maintenance, of mobilization of the tritium and {sup 7}Be, which are the impurities of the loop, and of the entire loss of confinement of the radioactive materials. We have calculated the early doses to the public due to the release of the tritium and {sup 7}Be source terms to the environment. The UFOTRI/COSYMA simulations have been performed considering the site boundary of 500 m away from the facility. The obtained results indicate that the early dose is below the level that requires the emergent public evacuation. Such results demonstrate that the A-FNS complies with the defined safety objective against its radiation hazard. The simulation results suggest that the inherent, ultimate safety characteristic found by this study may assist a licensing process for installation of A-FNS.

  16. Development of intervention levels for the protection of the public in the event of a major nuclear accident. Past, present and future

    International Nuclear Information System (INIS)

    Emmerson, B.W.

    1989-01-01

    Since the mid-1950's nuclear energy has played an increasing role in meeting the world demand for electricity production. Although during this period incidents and accidents have occurred, in most cases their effect was confined to the plant. Three accidents, however, were sufficienty serious as to involve off-site consequences for the public. The experience from each contributed significantly in the development of current emergency response criteria and planning arrangements at the national and international level. This paper summarizes these contributions as they relate to the development of intervention levels for the protection of the public in the event of an accidental release of radioactive materials to the environment. It indicates the various measures taken by those countries that were affected by the release from the Chernobyl accident and reviews the subsequent actions by relevant international organizations to provide more comprehensive guidance on applying the principles of intervention and developing derived levels, particularly those aimed at controlling the consumption of contamined foodstuffs, or their movement in international trade. Finally, it considers the prospects for developing a more harmonized intervention approach based on the guidance now being completed at the international level [fr

  17. A guide to countermeasures for implementation in the event of a nuclear accident affecting nordic food-producing areas

    International Nuclear Information System (INIS)

    Andersson, K.G.; Roed, J.; Rantavaara, A.; Rosen, K.; Salbu, B.; Skipperud, L.

    2000-08-01

    State-of-the-art information on methods for management of nuclear accidents affecting food-producing areas has been reviewed, evaluated and transposed to reflect conditions relevant to the Nordic countries. This data, describing in detail the various method-specific costs and benefits, is reported in a well-arranged format facilitating analyses in connection with decision-making. Guidance, recommendations and examples are given as to how the individual data sheets may be used in emergency preparedness planning. (au)

  18. Development of Accident Scenarios and Quantification Methodology for RAON Accelerator

    International Nuclear Information System (INIS)

    Lee, Yongjin; Jae, Moosung

    2014-01-01

    The RIsp (Rare Isotope Science Project) plans to provide neutron-rich isotopes (RIs) and stable heavy ion beams. The accelerator is defined as radiation production system according to Nuclear Safety Law. Therefore, it needs strict operate procedures and safety assurance to prevent radiation exposure. In order to satisfy this condition, there is a need for evaluating potential risk of accelerator from the design stage itself. Though some of PSA researches have been conducted for accelerator, most of them focus on not general accident sequence but simple explanation of accident. In this paper, general accident scenarios are developed by Event Tree and deduce new quantification methodology of Event Tree. In this study, some initial events, which may occur in the accelerator, are selected. Using selected initial events, the accident scenarios of accelerator facility are developed with Event Tree. These results can be used as basic data of the accelerator for future risk assessments. After analyzing the probability of each heading, it is possible to conduct quantification and evaluate the significance of the accident result. If there is a development of the accident scenario for external events, risk assessment of entire accelerator facility will be completed. To reduce the uncertainty of the Event Tree, it is possible to produce a reliable data via the presented quantification techniques

  19. Questions of jurisdiction in the event of a nuclear accident in a member state of the European union

    International Nuclear Information System (INIS)

    Galizzi, P.

    1996-01-01

    Jurisdictional problems are outlined that could be encountered by victims of a serious nuclear accident, with transboundary consequences, seeking to recover compensation (in a Member State of the European Union). The situation is only partly covered by existing treaty law and not all Member States are a party to the relevant treaties. A hypothetical case-study has been devised which supposes that a nuclear accident has occurred in the Netherlands causing damage in three selected countries. Of these, the first (Germany) is a party to the 1960 Paris Convention on Third Party Liability in the field of Nuclear Energy, the second (Hungary) is a party to the 1963 Vienna Convention on Civil Liability for Nuclear Damage, and the third (Luxembourg) is a not a party to either Convention. Answers are sought for two questions related to this hypothetical accident. Firstly, which courts have jurisdiction over private claims for damage caused in these various countries? Secondly, which law will the competent courts apply? (UK)

  20. Trans-oceanic transport of {sup 137}Cs from the Fukushima nuclear accident and impact of hypothetical Fukushima-like events of future nuclear plants in Southern China

    Energy Technology Data Exchange (ETDEWEB)

    Wai, Ka-Ming, E-mail: bhkmwai@cityu.edu.hk [Department of Geological and Mining Engineering and Sciences, Michigan Technological University, Houghton, MI (United States); Department of Physics and Material Science, City University of Hong Kong, Hong Kong (China); Yu, Peter K.N. [Department of Physics and Material Science, City University of Hong Kong, Hong Kong (China)

    2015-03-01

    A Lagrangian model was adopted to assess the potential impact of {sup 137}Cs released from hypothetical Fukushima-like accidents occurring on three potential nuclear power plant sites in Southern China in the near future (planned within 10 years) in four different seasons. The maximum surface (0–500 m) {sup 137}Cs air concentrations would be reached 10 Bq m{sup −3} near the source, comparable to the Fukushima case. In January, Southeast Asian countries would be mostly affected by the radioactive plume due to the effects of winter monsoon. In April, the impact would be mainly on Southern and Northern China. Debris of radioactive plume (∼ 1 mBq m{sup −3}) would carry out long-range transport to North America. The area of influence would be the smallest in July due to the frequent and intense wet removal events by trough of low pressure and tropical cyclone. The maximum worst-case areas of influence were 2382000, 2327000, 517000 and 1395000 km{sup 2} in January, April, July and October, respectively. Prior to the above calculations, the model was employed to simulate the trans-oceanic transport of {sup 137}Cs from the Fukushima nuclear accident. Observed and modeled {sup 137}Cs concentrations were comparable. Sensitivity runs were performed to optimize the wet scavenging parameterization. The adoption of higher-resolution (1° × 1°) meteorological fields improved the prediction. The computed large-scale plume transport pattern over the Pacific Ocean was compared with that reported in the literature. - Highlights: • A Lagrangian model was used to predict the dispersion of {sup 137}Cs from plant accident. • Observed and modeled {sup 137}Cs concentrations were comparable for the Fukushima accident. • The maximum surface concentrations could reach 10 Bq m{sup −3} for the hypothetical case. • The hypothetical radiative plumes could impact E/SE Asia and N. America.

  1. Re-assessment of road accident data-analysis policy : applying theory from involuntary, high-consequence, low-probability events like nuclear power plant meltdowns to voluntary, low-consequence, high-probability events like traffic accidents

    Science.gov (United States)

    2002-02-01

    This report examines the literature on involuntary, high-consequence, low-probability (IHL) events like nuclear power plant meltdowns to determine what can be applied to the problem of voluntary, low-consequence high-probability (VLH) events like tra...

  2. Reactivity insertion accident analysis

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

    1990-04-01

    The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

  3. Tchernobyl accident

    International Nuclear Information System (INIS)

    1986-06-01

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

  4. Visualization of Traffic Accidents

    Science.gov (United States)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

  5. Organization of public authorities in France for the event of an incident or accident involving nuclear safety: Simulation of a nuclear crisis

    International Nuclear Information System (INIS)

    Cartigny, J.; Majorel, Y.

    1986-01-01

    The French nuclear safety regulations lay down the action to be taken in the event of an incident or accident involving the types of radiological hazard that could arise in a nuclear installation or during the transport of radioactive material. The organization established for this purpose is designed to ensure that the technical measures taken by the authorities responsible for nuclear safety, radiation protection, public order and public safety are fully effective. The Interministerial Nuclear Safety Committee (Comite interministeriel de la securite nucleaire), which reports to the Prime Minister, co-ordinates the measures taken by the public authorities. The public authorities and the operators together organize exercises designed to verify the whole complex of measures foreseen in the event of an incident or accident. These exercises, which have been carried out in a systematic manner in France for some years, are based on scenarios which are as realistic as possible and enable the following objectives to be achieved: (1) analysis of the crisis apparatus (ORSECRAD plans, individual intervention plans, information conventions); (2) uncovering gaps or inadequacies; (3) arrangements for interchange of information between the various participants whose responsibilities involve them in the emergency; and (4) allowance for the information requirements of the media and the population. The information drawn from these exercises enables the various procedures to be improved step by step. (author)

  6. Supplemental analysis of accident sequences and source terms for waste treatment and storage operations and related facilities for the US Department of Energy waste management programmatic environmental impact statement

    International Nuclear Information System (INIS)

    Folga, S.; Mueller, C.; Nabelssi, B.; Kohout, E.; Mishima, J.

    1996-12-01

    This report presents supplemental information for the document Analysis of Accident Sequences and Source Terms at Waste Treatment, Storage, and Disposal Facilities for Waste Generated by US Department of Energy Waste Management Operations. Additional technical support information is supplied concerning treatment of transuranic waste by incineration and considering the Alternative Organic Treatment option for low-level mixed waste. The latest respirable airborne release fraction values published by the US Department of Energy for use in accident analysis have been used and are included as Appendix D, where respirable airborne release fraction is defined as the fraction of material exposed to accident stresses that could become airborne as a result of the accident. A set of dominant waste treatment processes and accident scenarios was selected for a screening-process analysis. A subset of results (release source terms) from this analysis is presented

  7. Event-Related Potential Correlates of Declarative and Non-Declarative Sequence Knowledge

    Science.gov (United States)

    Ferdinand, Nicola K.; Runger, Dennis; Frensch, Peter A.; Mecklinger, Axel

    2010-01-01

    The goal of the present study was to demonstrate that declarative and non-declarative knowledge acquired in an incidental sequence learning task contributes differentially to memory retrieval and leads to dissociable ERP signatures in a recognition memory task. For this purpose, participants performed a sequence learning task and were classified…

  8. Generic intervention levels for protecting the public in the event of a nuclear accident or radiological emergency

    International Nuclear Information System (INIS)

    1993-04-01

    Many international organizations are in the process of developing common safety standards for protection against ionizing radiation and for the safety of radiation sources. This document is intended to provide input to the final specification of intervention levels, at which actions of various kinds to protect members oft the public after an accident are advised. It provides the radiation protection principles underlying such intervention levels and proposes numerical values for these levels based on an analysis of some of the more directly quantifiable factors involved. Factors such as social disruption, psychological factors and political considerations are discussed, but are explicitly excluded from the derivation. Refs, figs and tabs

  9. Use of PSA to support accident management at NPPs

    International Nuclear Information System (INIS)

    Gomez Cobo, A.

    1997-01-01

    The presentation discusses the following: Overview of PSA level 2; Introduction: Framework; Accident Progression Phenomena in the Confinement/containment; Severe Accident Sequences; Examples; Results and Insights. Accident Management: Concepts; Process; Use of PSA to support Accident; Management

  10. Guide for the use of the regulations on medical surveillance to exposed workers in case of abnormal events (radiological accidents)

    International Nuclear Information System (INIS)

    1987-01-01

    According to medical surveillance, abnormal events are those extraordinary situations that may imply real or potential damage for a human being or a determined population. This guide refers to abnormal events that may imply, solely, to occupationally-exposed workers and small groups of population eventually related

  11. Technical Note: A novel leaf sequencing optimization algorithm which considers previous underdose and overdose events for MLC tracking radiotherapy

    Energy Technology Data Exchange (ETDEWEB)

    Wisotzky, Eric, E-mail: eric.wisotzky@charite.de, E-mail: eric.wisotzky@ipk.fraunhofer.de; O’Brien, Ricky; Keall, Paul J., E-mail: paul.keall@sydney.edu.au [Radiation Physics Laboratory, Sydney Medical School, University of Sydney, Sydney, NSW 2006 (Australia)

    2016-01-15

    Purpose: Multileaf collimator (MLC) tracking radiotherapy is complex as the beam pattern needs to be modified due to the planned intensity modulation as well as the real-time target motion. The target motion cannot be planned; therefore, the modified beam pattern differs from the original plan and the MLC sequence needs to be recomputed online. Current MLC tracking algorithms use a greedy heuristic in that they optimize for a given time, but ignore past errors. To overcome this problem, the authors have developed and improved an algorithm that minimizes large underdose and overdose regions. Additionally, previous underdose and overdose events are taken into account to avoid regions with high quantity of dose events. Methods: The authors improved the existing MLC motion control algorithm by introducing a cumulative underdose/overdose map. This map represents the actual projection of the planned tumor shape and logs occurring dose events at each specific regions. These events have an impact on the dose cost calculation and reduce recurrence of dose events at each region. The authors studied the improvement of the new temporal optimization algorithm in terms of the L1-norm minimization of the sum of overdose and underdose compared to not accounting for previous dose events. For evaluation, the authors simulated the delivery of 5 conformal and 14 intensity-modulated radiotherapy (IMRT)-plans with 7 3D patient measured tumor motion traces. Results: Simulations with conformal shapes showed an improvement of L1-norm up to 8.5% after 100 MLC modification steps. Experiments showed comparable improvements with the same type of treatment plans. Conclusions: A novel leaf sequencing optimization algorithm which considers previous dose events for MLC tracking radiotherapy has been developed and investigated. Reductions in underdose/overdose are observed for conformal and IMRT delivery.

  12. Historical aspects of radiation accidents

    International Nuclear Information System (INIS)

    Mettler, F.A. Jr.; Ricks, R.C.

    1990-01-01

    Radiation accidents are extremely rare events; however, the last two years have witnessed the largest radiation accidents in both the eastern and western hemispheres. It is the purpose of this chapter to review how radiation accidents are categorized, examine the temporal changes in frequency and severity, give illustrative examples of several types of radiation accidents, and finally, to describe the various registries for radiation accidents

  13. Did A Planet Survive A Post-Main Sequence Evolutionary Event?

    Science.gov (United States)

    Sorber, Rebecca; Jang-Condell, Hannah; Zimmerman, Mara

    2018-06-01

    The GL86 is star system approximately 10 pc away with a main sequence K- type ~ 0.77 M⊙ star (GL 86A) with a white dwarf ~0.49 M⊙ companion (GL86 B). The system has a ~ 18.4 AU semi-major axis, an orbital period of ~353 yrs, and an eccentricity of ~ 0.39. A 4.5 MJ planet orbits the main sequence star with a semi-major axis of 0.113 AU, an orbital period of 15.76 days, in a near circular orbit with an eccentricity of 0.046. If we assume that this planet was formed during the time when the white dwarf was a main sequence star, it would be difficult for the planet to have remained in a stable orbit during the post-main sequence evolution of GL86 B. The post-main sequence evolution with planet survival will be examined by modeling using the program Mercury (Chambers 1999). Using the model, we examine the origins of the planet: whether it formed before or after the post-main sequence evolution of GL86B. The modeling will give us insight into the dynamical evolution of, not only, the binary star system, but also the planet’s life cycle.

  14. The Fukushima accident

    International Nuclear Information System (INIS)

    Maqua, M.; Stueck, R.

    2012-01-01

    On 11 March 2011, the Tohoku earthquake and the subsequent tsunami hit the Japanese east coast, causing more than 15,000 fatalities. To this date, 3,000 people are still missing. The Fukushima Dai-ichi NPP was the nuclear installation that was most affected by the tsunami. The earthquake cut off the NPP from the national grid. About 45 minutes later, the tsunami flooded units 1-4 and led to core meltdown events with large releases for units 1, 2 and 3. Unit 4 had been in refuelling outage at that time and lost the cooling of the spent fuel pool for several days. Considerable hydrogen explosions occurred in units 1, 3 and 4. Shortly after the accident, TEPCO started to mitigate the consequences of the accident by providing external cooling to the reactors and by removing the radioactive debris from the site. Great emphasis was laid on effective radiation protection measures for the clean-up workers. Thus, up to now there has been no fatality due to the radiation caused by the Fukushima accident. The main steps of the accident sequences are described, taking into account the latest findings of investigations performed by TEPCO or on behalf of the regulatory body. The presentation focuses on the description of the status of the Fukushima Dai-ichi nuclear power plant and the future steps for cleaning-up the site. In the presentation, the major phases of the roadmap that TEPCO has developed for the clean-up are highlighted. The risks associated with the current plant status and the clean-up phases are described. Abstract the content of the manuscript in a few lines.

  15. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  16. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    2008-06-01

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

  17. Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Park, Y. W.; Park, H.G. [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly.

  18. Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

    International Nuclear Information System (INIS)

    Park, J. Y.; Park, Y. W.; Park, H.G.

    2016-01-01

    This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly

  19. Fukushima. The accident sequence and important causes. Pt. 3/3; Fukushima. Unfallablauf und wesentliche Ursachen. T. 3/3

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany). Bereich Nukleartechnik und Anlagensicherheit

    2013-07-01

    The immediate cause of the Fukushima Daiichi disaster was the earthquake that was stronger than the design basis of the NPP Fukushima. The earth quake has at least destroyed the external power supply for all six units of the power plant. It is not yet clear whether other damage has been caused in the different units. The subsequent tsunami was of dominant importance for the progress of the reactor accidents. The power plant had no appropriate protection against tsunamis of this fortitude. The connections between politics, regulatory authority and owner of the power plant did not allow an effective and independent surveillance of the activities in the power plant. The most important principles of reactor safety were not implemented in the NPP Fukushima (for instance: the heat removal from the condensation chambers was dependent on a single heat sink). The local infrastructure was not protected against severe damage from earth quakes or tsunamis, so that immediate mitigating actions were not possible.

  20. The accident at the Three Mile Island nuclear power plant

    International Nuclear Information System (INIS)

    Butragueno, J.L.

    1980-01-01

    The sequence of events in the Three Mile Island, Unit 2, accident on the March 28, 1979 is analyzed. In this plant a loss of feed-water transient became a small LOCA that caused a serious core damage. A general emergency situation was declared after uncontrolled radioactive releases were detectec. (author)

  1. Development of the severe accident risk information database management system SARD

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

    The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies

  2. Development of the severe accident risk information database management system SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

    The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies.

  3. Analysis of severe core damage accident progression for the heavy water reactor

    International Nuclear Information System (INIS)

    Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

    2010-01-01

    In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

  4. Causal factors in accidents of high-speed craft and conventional ocean-going vessels

    International Nuclear Information System (INIS)

    Antao, Pedro; Guedes Soares, C.

    2008-01-01

    An analysis of 40 ocean-going commercial vessel accidents is compared with the study of a similar number of high-speed crafts (HSCs) accidents, using in both cases a methodology that highlights the sequence of events leading to the accident and identifies the associated latent or causal factors. The main objective of this study was to identify and understand the difference in the pattern of causal factors associated with HSC accidents, as compared with the more traditional ocean-going ships. From the analysis one can see that the HSC accidents are mainly related to bridge personnel and operations, where the human element is the key factor identified as being responsible for the majority of the accidents. When compared with ocean-going commercial vessels, it is clear that navigational equipment and procedures have a larger preponderance in terms of the occurrence of accidents of HSC and particular attention should be given to these issues

  5. The Chernobyl accident consequences

    International Nuclear Information System (INIS)

    2001-04-01

    Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

  6. Radiation, accidents, society

    International Nuclear Information System (INIS)

    1988-01-01

    This book is meant to be used as a reference book for information officers at the event of a nuclear accident. The main part is edited in alphabetical order to facilitate use under stress. The book gives a short review of the health risks of radiation, and descriptions of accidents that have occured. The index words that have been chosen for the main part of the book have been selected due to experiences in connection with incidents and accidents. (L.E.)

  7. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  8. Revisiting the Canterbury earthquake sequence after the 14 February 2016 Mw 5.7 event

    NARCIS (Netherlands)

    Herman, Matthew W.; Furlong, Kevin P.

    2016-01-01

    On 14 February 2016, an Mw 5.7 (GNS Science moment magnitude) earthquake ruptured offshore east of Christchurch, New Zealand. This earthquake occurred in an area that had previously experienced significant seismicity from 2010 to 2012 during the Canterbury earthquake sequence, starting with the 2010

  9. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  10. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J. [and others

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.

  11. Seepage patterns of Diuron in a ditch bed during a sequence of flood events

    International Nuclear Information System (INIS)

    Dages, C.; Samouëlian, A.; Negro, S.; Storck, V.; Huttel, O.; Voltz, M.

    2015-01-01

    Although ditches limit surface water contamination, groundwater recharge through ditches in Mediterranean catchments may result in groundwater contamination. We analysed the dynamics of pesticide percolation in ditches by conducting an original lab experiment that mimicked the successive percolation processes that occur during a flood season. Nine successive percolation events were operated on an undisturbed soil column collected from a ditch bed. The infiltrating water was doped with 14 C-Diuron at concentrations that were chosen to decrease between the events so as to correspond to values observed during actual flood events. The water and solute fluxes were monitored during each event, and the final extractable and non-extractable Diuron residues in the column were determined. Two main observations were made. First, a high leaching potential was observed through the ditch bed over a succession of infiltrating flood events, with 58.9% of the infiltrated Diuron and its metabolites leaching. Second, compared with the contamination of surface water circulating in the ditches, the contamination of seepage water exhibited smaller peak values and persisted much longer because of the desorption of Diuron residues stored in the ditch bed. Thus, ditches serve as buffering zones between surface and groundwater. However, compared with field plots, ditches appear to be a preferential location for the percolation of pesticides into groundwater at the catchment scale. - Highlights: • Diuron percolation in a ditch bed during flood events was mimicked in a column setup. • Diuron percolation can represent up to 50% of the infiltrated Diuron. • The ditch bed exhibits a high buffering capacity due to its high sorption properties. • Contamination period of percolation water lasts longer than that of infiltrating water. • Diuron residues stored in ditch bed move deeper than in field topsoils.

  12. Seepage patterns of Diuron in a ditch bed during a sequence of flood events

    Energy Technology Data Exchange (ETDEWEB)

    Dages, C., E-mail: cecile.dages@supagro.inra.fr; Samouëlian, A.; Negro, S.; Storck, V.; Huttel, O.; Voltz, M.

    2015-12-15

    Although ditches limit surface water contamination, groundwater recharge through ditches in Mediterranean catchments may result in groundwater contamination. We analysed the dynamics of pesticide percolation in ditches by conducting an original lab experiment that mimicked the successive percolation processes that occur during a flood season. Nine successive percolation events were operated on an undisturbed soil column collected from a ditch bed. The infiltrating water was doped with {sup 14}C-Diuron at concentrations that were chosen to decrease between the events so as to correspond to values observed during actual flood events. The water and solute fluxes were monitored during each event, and the final extractable and non-extractable Diuron residues in the column were determined. Two main observations were made. First, a high leaching potential was observed through the ditch bed over a succession of infiltrating flood events, with 58.9% of the infiltrated Diuron and its metabolites leaching. Second, compared with the contamination of surface water circulating in the ditches, the contamination of seepage water exhibited smaller peak values and persisted much longer because of the desorption of Diuron residues stored in the ditch bed. Thus, ditches serve as buffering zones between surface and groundwater. However, compared with field plots, ditches appear to be a preferential location for the percolation of pesticides into groundwater at the catchment scale. - Highlights: • Diuron percolation in a ditch bed during flood events was mimicked in a column setup. • Diuron percolation can represent up to 50% of the infiltrated Diuron. • The ditch bed exhibits a high buffering capacity due to its high sorption properties. • Contamination period of percolation water lasts longer than that of infiltrating water. • Diuron residues stored in ditch bed move deeper than in field topsoils.

  13. Experimental analysis of the behaviour of iodine in the event of hypothetical accidents. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Richter, F.; Rippel, R.; Proebstle, G.; Fernholz, O.

    1986-01-01

    Experiments have been performed simulating hypothetical core-melt accidents in order to determine droplet-bound transport of radio-nuclides. Different measurement methods have been applied to evaluate steam moisture and droplet size distribution, the carry-over factor of a tracer substance, and, to some extent, droplet velocity, under atmospheric sump water boiling conditions. Part flow analysis yields carry-over factor values on the order of magnitude 10 -5 . Thus it is smaller than would be expected from visual measurements of steam moisture in the main flow, a result which is due to droplet velocity characteristics which limit the carry-over through openings. Results distinctly show that steam moisture (10 -3 up to 7x10 -5 , depending on the distance from the sump) and the droplet size (4-57 μm) can only be used as a source term. In order to evaluate the quantity released from a leakage, a supplementary investigation of droplet carry-over mechanisms will be required. (orig.) [de

  14. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  15. Timing and sequencing of events marking the transition to adulthood in two informal settlements in Nairobi, Kenya.

    Science.gov (United States)

    Beguy, Donatien; Kabiru, Caroline W; Zulu, Eliya M; Ezeh, Alex C

    2011-06-01

    Young people living in poor urban informal settlements face unique challenges as they transition to adulthood. This exploratory paper uses retrospective information from the baseline survey of a 3-year prospective study to examine the timing and sequencing of four key markers (first sex, marriage, birth, and independent housing) of the transition to adulthood among 3,944 adolescents in two informal settlements in Nairobi city, Kenya. Event history analysis techniques are employed to examine the timing of the events. Results indicate that there is no significant gender difference with regard to first sexual debut among adolescents. For many boys and girls, the first sexual experience occurs outside of marriage or other union. For males, the sequencing of entry begins with entry into first sex, followed by independent housing. Conversely, for females, the sequencing begins with first sex and then parenthood. Apart from sexual debut, the patterns of entry into union and parenthood do not differ much from what was observed for Nairobi as a whole. The space constraints that typify the two slums may have influenced the pattern of leaving home observed. We discuss these and other findings in light of their implications for young people's health and well-being in resource-poor settings in urban areas.

  16. Children's Representation and Imitation of Events: How Goal Organization Influences 3-Year-Old Children's Memory for Action Sequences.

    Science.gov (United States)

    Loucks, Jeff; Mutschler, Christina; Meltzoff, Andrew N

    2017-09-01

    Children's imitation of adults plays a prominent role in human cognitive development. However, few studies have investigated how children represent the complex structure of observed actions which underlies their imitation. We integrate theories of action segmentation, memory, and imitation to investigate whether children's event representation is organized according to veridical serial order or a higher level goal structure. Children were randomly assigned to learn novel event sequences either through interactive hands-on experience (Study 1) or via storybook (Study 2). Results demonstrate that children's representation of observed actions is organized according to higher level goals, even at the cost of representing the veridical temporal ordering of the sequence. We argue that prioritizing goal structure enhances event memory, and that this mental organization is a key mechanism of social-cognitive development in real-world, dynamic environments. It supports cultural learning and imitation in ecologically valid settings when social agents are multitasking and not demonstrating one isolated goal at a time. Copyright © 2016 Cognitive Science Society, Inc.

  17. Criticality accident in Argentina

    International Nuclear Information System (INIS)

    Oliveira, A.R. de.

    1984-01-01

    A recent criticality type accident, ocurred in Argetina, is commented. Considerations about the nature of the facility where this accident took place, its genesis, type of operation carried out on the day of the event, and the medical aspects involved are done. (Author) [pt

  18. HLA DNA sequence variation among human populations: molecular signatures of demographic and selective events.

    Directory of Open Access Journals (Sweden)

    Stéphane Buhler

    2011-02-01

    Full Text Available Molecular differences between HLA alleles vary up to 57 nucleotides within the peptide binding coding region of human Major Histocompatibility Complex (MHC genes, but it is still unclear whether this variation results from a stochastic process or from selective constraints related to functional differences among HLA molecules. Although HLA alleles are generally treated as equidistant molecular units in population genetic studies, DNA sequence diversity among populations is also crucial to interpret the observed HLA polymorphism. In this study, we used a large dataset of 2,062 DNA sequences defined for the different HLA alleles to analyze nucleotide diversity of seven HLA genes in 23,500 individuals of about 200 populations spread worldwide. We first analyzed the HLA molecular structure and diversity of these populations in relation to geographic variation and we further investigated possible departures from selective neutrality through Tajima's tests and mismatch distributions. All results were compared to those obtained by classical approaches applied to HLA allele frequencies.Our study shows that the global patterns of HLA nucleotide diversity among populations are significantly correlated to geography, although in some specific cases the molecular information reveals unexpected genetic relationships. At all loci except HLA-DPB1, populations have accumulated a high proportion of very divergent alleles, suggesting an advantage of heterozygotes expressing molecularly distant HLA molecules (asymmetric overdominant selection model. However, both different intensities of selection and unequal levels of gene conversion may explain the heterogeneous mismatch distributions observed among the loci. Also, distinctive patterns of sequence divergence observed at the HLA-DPB1 locus suggest current neutrality but old selective pressures on this gene. We conclude that HLA DNA sequences advantageously complement HLA allele frequencies as a source of data used

  19. Public health and regulatory concerns in the event of 131 milk contamination from fallout or a nuclear facility accident

    International Nuclear Information System (INIS)

    Davies, Sherwood

    1978-01-01

    Deposition of 131 I on pasture land in high density milk producing areas may result in large volumes of contaminated milk entering the food chain. The potential for radiation exposure of the population through milk may extend to relatively large areas. The magnitude of exposure through ingestion of milk may be 300 to 700 times greater than through inhalation. Measurable amounts of 131 I can be identified in milk within 12-24 hours after a contaminating event with peak milk concentrations occurring within a few days. Implementation of needed protective actions within a day following a contaminating event can minimize milk contamination and avoid unnecessary population exposure. Regulatory agencies should consider the potential for milk contamination, point of control, i.e., dairy farm-milk plant-man, time framework for initiating needed protective actions, effectiveness of control procedures, sampling and laboratory resources, interpretation of anomalous data and preplanned enforcement actions. (author)

  20. Analysis of Three Mile Island Unit 2 accident

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    NSAC is conducting a detailed review of this accident and of the lessons to be learned. So far it has concentrated primarily on events during the sixteen hours following initiation of the accident. A sequence of events has been developed and is being verified and annotated by comparing oral and written statements with instrumentation records, data logs, operator logs, and inferences which can be made from these records. This report is being developed with the expectation that, while not completed or fully verified, it may be useful at this time. Supplements may be issued later as the analyses which are still under way are completed

  1. Accident scenarios triggered by lightning strike on atmospheric storage tanks

    International Nuclear Information System (INIS)

    Necci, Amos; Argenti, Francesca; Landucci, Gabriele; Cozzani, Valerio

    2014-01-01

    Severe Natech accidents may be triggered by lightning strike affecting storage tanks containing relevant inventories of hazardous materials. The present study focused on the identification of event sequences and accident scenarios following lightning impact on atmospheric tanks. Reference event trees, validated using past accident analysis, are provided to describe the specific accident chains identified, accounting for reference protection and mitigation safety barriers usually adopted in current industrial practice. An overall methodology was outlined to allow the calculation of the expected frequencies of final scenarios following lightning impact on atmospheric storage tanks, taking into account the expected performance of available safety barriers. The methodology was applied to a case study in order to better understand the data that may be obtained and their importance in the framework of quantitative risk assessment (QRA) and of the risk management of industrial facilities with respect to external hazards due to natural events. - Highlights: • Event sequences following lightning impact on atmospheric tanks were identified. • Reference event trees including standard safety barriers were obtained. • Safety barriers applied in industrial practice were assessed to quantify event trees. • Frequencies of final scenarios following lightning impact on tanks were calculated. • Natech scenarios caused by lightning have an important influence on risk profiles

  2. A fast one-chip event-preprocessor and sequencer for the Simbol-X Low Energy Detector

    Science.gov (United States)

    Schanz, T.; Tenzer, C.; Maier, D.; Kendziorra, E.; Santangelo, A.

    2010-12-01

    We present an FPGA-based digital camera electronics consisting of an Event-Preprocessor (EPP) for on-board data preprocessing and a related Sequencer (SEQ) to generate the necessary signals to control the readout of the detector. The device has been originally designed for the Simbol-X low energy detector (LED). The EPP operates on 64×64 pixel images and has a real-time processing capability of more than 8000 frames per second. The already working releases of the EPP and the SEQ are now combined into one Digital-Camera-Controller-Chip (D3C).

  3. A fast one-chip event-preprocessor and sequencer for the Simbol-X Low Energy Detector

    Energy Technology Data Exchange (ETDEWEB)

    Schanz, T., E-mail: schanz@astro.uni-tuebingen.d [Kepler Center for Astro- and Particlephysics, Institut fuer Astronomie und Astrophysik Tuebingen, Sand 1, 72076 Tuebingen (Germany); Tenzer, C., E-mail: tenzer@astro.uni-tuebingen.d [Kepler Center for Astro- and Particlephysics, Institut fuer Astronomie und Astrophysik Tuebingen, Sand 1, 72076 Tuebingen (Germany); Maier, D.; Kendziorra, E.; Santangelo, A. [Kepler Center for Astro- and Particlephysics, Institut fuer Astronomie und Astrophysik Tuebingen, Sand 1, 72076 Tuebingen (Germany)

    2010-12-11

    We present an FPGA-based digital camera electronics consisting of an Event-Preprocessor (EPP) for on-board data preprocessing and a related Sequencer (SEQ) to generate the necessary signals to control the readout of the detector. The device has been originally designed for the Simbol-X low energy detector (LED). The EPP operates on 64x64 pixel images and has a real-time processing capability of more than 8000 frames per second. The already working releases of the EPP and the SEQ are now combined into one Digital-Camera-Controller-Chip (D3C).

  4. A fast one-chip event-preprocessor and sequencer for the Simbol-X Low Energy Detector

    International Nuclear Information System (INIS)

    Schanz, T.; Tenzer, C.; Maier, D.; Kendziorra, E.; Santangelo, A.

    2010-01-01

    We present an FPGA-based digital camera electronics consisting of an Event-Preprocessor (EPP) for on-board data preprocessing and a related Sequencer (SEQ) to generate the necessary signals to control the readout of the detector. The device has been originally designed for the Simbol-X low energy detector (LED). The EPP operates on 64x64 pixel images and has a real-time processing capability of more than 8000 frames per second. The already working releases of the EPP and the SEQ are now combined into one Digital-Camera-Controller-Chip (D3C).

  5. Events

    Directory of Open Access Journals (Sweden)

    Igor V. Karyakin

    2016-02-01

    Full Text Available The 9th ARRCN Symposium 2015 was held during 21st–25th October 2015 at the Novotel Hotel, Chumphon, Thailand, one of the most favored travel destinations in Asia. The 10th ARRCN Symposium 2017 will be held during October 2017 in the Davao, Philippines. International Symposium on the Montagu's Harrier (Circus pygargus «The Montagu's Harrier in Europe. Status. Threats. Protection», organized by the environmental organization «Landesbund für Vogelschutz in Bayern e.V.» (LBV was held on November 20-22, 2015 in Germany. The location of this event was the city of Wurzburg in Bavaria.

  6. Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP; Branduolines saugos ir avariju eigos Ignalinos AE tyrimu apibendrinimas

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, A; Uspuras, E [Lithuanian Energy Institute, Kaunas (Lithuania)

    2001-07-01

    The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

  7. Automatically Identifying Fusion Events between GLUT4 Storage Vesicles and the Plasma Membrane in TIRF Microscopy Image Sequences

    Directory of Open Access Journals (Sweden)

    Jian Wu

    2015-01-01

    Full Text Available Quantitative analysis of the dynamic behavior about membrane-bound secretory vesicles has proven to be important in biological research. This paper proposes a novel approach to automatically identify the elusive fusion events between VAMP2-pHluorin labeled GLUT4 storage vesicles (GSVs and the plasma membrane. The differentiation is implemented to detect the initiation of fusion events by modified forward subtraction of consecutive frames in the TIRFM image sequence. Spatially connected pixels in difference images brighter than a specified adaptive threshold are grouped into a distinct fusion spot. The vesicles are located at the intensity-weighted centroid of their fusion spots. To reveal the true in vivo nature of a fusion event, 2D Gaussian fitting for the fusion spot is used to derive the intensity-weighted centroid and the spot size during the fusion process. The fusion event and its termination can be determined according to the change of spot size. The method is evaluated on real experiment data with ground truth annotated by expert cell biologists. The evaluation results show that it can achieve relatively high accuracy comparing favorably to the manual analysis, yet at a small fraction of time.

  8. Structural aspects of the Chernobyl accident

    International Nuclear Information System (INIS)

    Murray, R.C.; Cummings, G.E.

    1988-01-01

    On April 26, 1986 the world's worst nuclear power plant accident occurred at the Unit 4 of the Chernobyl Nuclear Power Station in the USSR. This paper presents a discussion of the design of the Chernobyl Power Plant, the sequence of events that led to the accident and the damage caused by the resulting explosion. The structural design features that contributed to the accident and resulting damage will be highlighted. Photographs and sketches obtained from various worldwide news agencies will be shown to try and gain a perspective of the extent of the damage. The aftermath, clean-up, and current situation will be discussed and the important lessons learned for the structural engineer will be presented. 15 refs., 10 figs

  9. Event-sequence time series analysis in ground-based gamma-ray astronomy

    International Nuclear Information System (INIS)

    Barres de Almeida, U.; Chadwick, P.; Daniel, M.; Nolan, S.; McComb, L.

    2008-01-01

    The recent, extreme episodes of variability detected from Blazars by the leading atmospheric Cerenkov experiments motivate the development and application of specialized statistical techniques that enable the study of this rich data set to its furthest extent. The identification of the shortest variability timescales supported by the data and the actual variability structure observed in the light curves of these sources are some of the fundamental aspects being studied, that answers can bring new developments on the understanding of the physics of these objects and on the mechanisms of production of VHE gamma-rays in the Universe. Some of our efforts in studying the time variability of VHE sources involve the application of dynamic programming algorithms to the problem of detecting change-points in a Poisson sequence. In this particular paper we concentrate on the more primary issue of the applicability of counting statistics to the analysis of time-series on VHE gamma-ray astronomy.

  10. Learning and Recognition of a Non-conscious Sequence of Events in Human Primary Visual Cortex.

    Science.gov (United States)

    Rosenthal, Clive R; Andrews, Samantha K; Antoniades, Chrystalina A; Kennard, Christopher; Soto, David

    2016-03-21

    Human primary visual cortex (V1) has long been associated with learning simple low-level visual discriminations [1] and is classically considered outside of neural systems that support high-level cognitive behavior in contexts that differ from the original conditions of learning, such as recognition memory [2, 3]. Here, we used a novel fMRI-based dichoptic masking protocol-designed to induce activity in V1, without modulation from visual awareness-to test whether human V1 is implicated in human observers rapidly learning and then later (15-20 min) recognizing a non-conscious and complex (second-order) visuospatial sequence. Learning was associated with a change in V1 activity, as part of a temporo-occipital and basal ganglia network, which is at variance with the cortico-cerebellar network identified in prior studies of "implicit" sequence learning that involved motor responses and visible stimuli (e.g., [4]). Recognition memory was associated with V1 activity, as part of a temporo-occipital network involving the hippocampus, under conditions that were not imputable to mechanisms associated with conscious retrieval. Notably, the V1 responses during learning and recognition separately predicted non-conscious recognition memory, and functional coupling between V1 and the hippocampus was enhanced for old retrieval cues. The results provide a basis for novel hypotheses about the signals that can drive recognition memory, because these data (1) identify human V1 with a memory network that can code complex associative serial visuospatial information and support later non-conscious recognition memory-guided behavior (cf. [5]) and (2) align with mouse models of experience-dependent V1 plasticity in learning and memory [6]. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. Whale phylogeny and rapid radiation events revealed using novel retroposed elements and their flanking sequences

    Directory of Open Access Journals (Sweden)

    Zhou Kaiya

    2011-10-01

    Full Text Available Abstract Background A diversity of hypotheses have been proposed based on both morphological and molecular data to reveal phylogenetic relationships within the order Cetacea (dolphins, porpoises, and whales, and great progress has been made in the past two decades. However, there is still some controversy concerning relationships among certain cetacean taxa such as river dolphins and delphinoid species, which needs to be further addressed with more markers in an effort to address unresolved portions of the phylogeny. Results An analysis of additional SINE insertions and SINE-flanking sequences supported the monophyly of the order Cetacea as well as Odontocete, Delphinoidea (Delphinidae + Phocoenidae + Mondontidae, and Delphinidae. A sister relationship between Delphinidae and Phocoenidae + Mondontidae was supported, and members of classical river dolphins and the genera Tursiops and Stenella were found to be paraphyletic. Estimates of divergence times revealed rapid divergences of basal Odontocete lineages in the Oligocene and Early Miocene, and a recent rapid diversification of Delphinidae in the Middle-Late Miocene and Pliocene within a narrow time frame. Conclusions Several novel SINEs were found to differentiate Delphinidae from the other two families (Monodontidae and Phocoenidae, whereas the sister grouping of the latter two families with exclusion of Delphinidae was further revealed using the SINE-flanking sequences. Interestingly, some anomalous PCR amplification patterns of SINE insertions were detected, which can be explained as the result of potential ancestral SINE polymorphisms and incomplete lineage sorting. Although a few loci were potentially anomalous, this study demonstrated that the SINE-based approach is a powerful tool in phylogenetic studies. Identifying additional SINE elements that resolve the relationships in the superfamily Delphinoidea and family Delphinidae will be important steps forward in completely resolving

  12. Whale phylogeny and rapid radiation events revealed using novel retroposed elements and their flanking sequences.

    Science.gov (United States)

    Chen, Zhuo; Xu, Shixia; Zhou, Kaiya; Yang, Guang

    2011-10-27

    A diversity of hypotheses have been proposed based on both morphological and molecular data to reveal phylogenetic relationships within the order Cetacea (dolphins, porpoises, and whales), and great progress has been made in the past two decades. However, there is still some controversy concerning relationships among certain cetacean taxa such as river dolphins and delphinoid species, which needs to be further addressed with more markers in an effort to address unresolved portions of the phylogeny. An analysis of additional SINE insertions and SINE-flanking sequences supported the monophyly of the order Cetacea as well as Odontocete, Delphinoidea (Delphinidae + Phocoenidae + Mondontidae), and Delphinidae. A sister relationship between Delphinidae and Phocoenidae + Mondontidae was supported, and members of classical river dolphins and the genera Tursiops and Stenella were found to be paraphyletic. Estimates of divergence times revealed rapid divergences of basal Odontocete lineages in the Oligocene and Early Miocene, and a recent rapid diversification of Delphinidae in the Middle-Late Miocene and Pliocene within a narrow time frame. Several novel SINEs were found to differentiate Delphinidae from the other two families (Monodontidae and Phocoenidae), whereas the sister grouping of the latter two families with exclusion of Delphinidae was further revealed using the SINE-flanking sequences. Interestingly, some anomalous PCR amplification patterns of SINE insertions were detected, which can be explained as the result of potential ancestral SINE polymorphisms and incomplete lineage sorting. Although a few loci were potentially anomalous, this study demonstrated that the SINE-based approach is a powerful tool in phylogenetic studies. Identifying additional SINE elements that resolve the relationships in the superfamily Delphinoidea and family Delphinidae will be important steps forward in completely resolving cetacean phylogenetic relationships in the future.

  13. The 2008 Wells, Nevada earthquake sequence: Source constraints using calibrated multiple event relocation and InSAR

    Science.gov (United States)

    Nealy, Jennifer; Benz, Harley M.; Hayes, Gavin; Berman, Eric; Barnhart, William

    2017-01-01

    The 2008 Wells, NV earthquake represents the largest domestic event in the conterminous U.S. outside of California since the October 1983 Borah Peak earthquake in southern Idaho. We present an improved catalog, magnitude complete to 1.6, of the foreshock-aftershock sequence, supplementing the current U.S. Geological Survey (USGS) Preliminary Determination of Epicenters (PDE) catalog with 1,928 well-located events. In order to create this catalog, both subspace and kurtosis detectors are used to obtain an initial set of earthquakes and associated locations. The latter are then calibrated through the implementation of the hypocentroidal decomposition method and relocated using the BayesLoc relocation technique. We additionally perform a finite fault slip analysis of the mainshock using InSAR observations. By combining the relocated sequence with the finite fault analysis, we show that the aftershocks occur primarily updip and along the southwestern edge of the zone of maximum slip. The aftershock locations illuminate areas of post-mainshock strain increase; aftershock depths, ranging from 5 to 16 km, are consistent with InSAR imaging, which shows that the Wells earthquake was a buried source with no observable near-surface offset.

  14. Sequence of pathogenic events in cynomolgus macaques infected with aerosolized monkeypox virus.

    Science.gov (United States)

    Tree, J A; Hall, G; Pearson, G; Rayner, E; Graham, V A; Steeds, K; Bewley, K R; Hatch, G J; Dennis, M; Taylor, I; Roberts, A D; Funnell, S G P; Vipond, J

    2015-04-01

    To evaluate new vaccines when human efficacy studies are not possible, the FDA's "Animal Rule" requires well-characterized models of infection. Thus, in the present study, the early pathogenic events of monkeypox infection in nonhuman primates, a surrogate for variola virus infection, were characterized. Cynomolgus macaques were exposed to aerosolized monkeypox virus (10(5) PFU). Clinical observations, viral loads, immune responses, and pathological changes were examined on days 2, 4, 6, 8, 10, and 12 postchallenge. Viral DNA (vDNA) was detected in the lungs on day 2 postchallenge, and viral antigen was detected, by immunostaining, in the epithelium of bronchi, bronchioles, and alveolar walls. Lesions comprised rare foci of dysplastic and sloughed cells in respiratory bronchioles. By day 4, vDNA was detected in the throat, tonsil, and spleen, and monkeypox antigen was detected in the lung, hilar and submandibular lymph nodes, spleen, and colon. Lung lesions comprised focal epithelial necrosis and inflammation. Body temperature peaked on day 6, pox lesions appeared on the skin, and lesions, with positive immunostaining, were present in the lung, tonsil, spleen, lymph nodes, and colon. By day 8, vDNA was present in 9/13 tissues. Blood concentrations of interleukin 1ra (IL-1ra), IL-6, and gamma interferon (IFN-γ) increased markedly. By day 10, circulating IgG antibody concentrations increased, and on day 12, animals showed early signs of recovery. These results define early events occurring in an inhalational macaque monkeypox infection model, supporting its use as a surrogate model for human smallpox. Bioterrorism poses a major threat to public health, as the deliberate release of infectious agents, such smallpox or a related virus, monkeypox, would have catastrophic consequences. The development and testing of new medical countermeasures, e.g., vaccines, are thus priorities; however, tests for efficacy in humans cannot be performed because it would be unethical and

  15. Review the number of accidents in Tehran over a two-year period and prediction of the number of events based on a time-series model

    Science.gov (United States)

    Teymuri, Ghulam Heidar; Sadeghian, Marzieh; Kangavari, Mehdi; Asghari, Mehdi; Madrese, Elham; Abbasinia, Marzieh; Ahmadnezhad, Iman; Gholizadeh, Yavar

    2013-01-01

    Background: One of the significant dangers that threaten people’s lives is the increased risk of accidents. Annually, more than 1.3 million people die around the world as a result of accidents, and it has been estimated that approximately 300 deaths occur daily due to traffic accidents in the world with more than 50% of that number being people who were not even passengers in the cars. The aim of this study was to examine traffic accidents in Tehran and forecast the number of future accidents using a time-series model. Methods: The study was a cross-sectional study that was conducted in 2011. The sample population was all traffic accidents that caused death and physical injuries in Tehran in 2010 and 2011, as registered in the Tehran Emergency ward. The present study used Minitab 15 software to provide a description of accidents in Tehran for the specified time period as well as those that occurred during April 2012. Results: The results indicated that the average number of daily traffic accidents in Tehran in 2010 was 187 with a standard deviation of 83.6. In 2011, there was an average of 180 daily traffic accidents with a standard deviation of 39.5. One-way analysis of variance indicated that the average number of accidents in the city was different for different months of the year (P accidents occurred in March, July, August, and September. Thus, more accidents occurred in the summer than in the other seasons. The number of accidents was predicted based on an auto-regressive, moving average (ARMA) for April 2012. The number of accidents displayed a seasonal trend. The prediction of the number of accidents in the city during April of 2012 indicated that a total of 4,459 accidents would occur with mean of 149 accidents per day during these three months. Conclusion: The number of accidents in Tehran displayed a seasonal trend, and the number of accidents was different for different seasons of the year. PMID:26120405

  16. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 1: Sections 1-9

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

    1995-04-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  17. The process of suicide: the sequence of disruptive events in the lives of suicide victims.

    Science.gov (United States)

    Humphrey, J A; French, L; Niswander, G D; Casey, T M

    1974-06-01

    Certain disruptions in social relations or loss of social roles tend to precede the act of suicide. In general, this study has attempted to demonstrate the sequential ordering of these social losses and otherdisruptive events. It has been shown that loss of roles irn childhood and early adolescence, parents, home, sibilings, close relatives, and student role tends to be followed by chaotic marriages, loss of occupation role(s), and loss of one's own health either physical or mentalor, in some cases, both. Also parental roles tend to be lost through the death of a child. Take away one's social roles and you take away his humanity (See Palmer, 1970). An individual with nowhere to turn and having lost all, tends to be highly vulnerable to suicide.

  18. Sequence of eruptive events in the Vesuvio area recorded in shallow-water Ionian Sea sediments

    Directory of Open Access Journals (Sweden)

    C. Taricco

    2008-01-01

    Full Text Available The dating of the cores we drilled from the Gallipoli terrace in the Gulf of Taranto (Ionian Sea, previously obtained by tephroanalysis, is checked by applying a method to objectively recognize volcanic events. This automatic statistical procedure allows identifying pulse-like features in a series and evaluating quantitatively the confidence level at which the significant peaks are detected. We applied it to the 2000-years-long pyroxenes series of the GT89-3 core, on which the dating is based. The method confirms the dating previously performed by detecting at a high confidence level the peaks originally used and indicates a few possible undocumented eruptions. Moreover, a spectral analysis, focussed on the long-term variability of the pyroxenes series and performed by several advanced methods, reveals that the volcanic pulses are superimposed to a millennial trend and a 400 years oscillation.

  19. Sequence of eruptive events in the Vesuvio area recorded in shallow-water Ionian Sea sediments

    Science.gov (United States)

    Taricco, C.; Alessio, S.; Vivaldo, G.

    2008-01-01

    The dating of the cores we drilled from the Gallipoli terrace in the Gulf of Taranto (Ionian Sea), previously obtained by tephroanalysis, is checked by applying a method to objectively recognize volcanic events. This automatic statistical procedure allows identifying pulse-like features in a series and evaluating quantitatively the confidence level at which the significant peaks are detected. We applied it to the 2000-years-long pyroxenes series of the GT89-3 core, on which the dating is based. The method confirms the dating previously performed by detecting at a high confidence level the peaks originally used and indicates a few possible undocumented eruptions. Moreover, a spectral analysis, focussed on the long-term variability of the pyroxenes series and performed by several advanced methods, reveals that the volcanic pulses are superimposed to a millennial trend and a 400 years oscillation.

  20. Chernobyl accident

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1995-01-01

    The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

  1. The Complete Genome Sequence of Herpesvirus Papio 2 (Cercopithecine Herpesvirus 16) Shows Evidence of Recombination Events among Various Progenitor Herpesviruses†

    Science.gov (United States)

    Tyler, Shaun D.; Severini, Alberto

    2006-01-01

    We have sequenced the entire genome of herpesvirus papio 2 (HVP-2; Cercopithecine herpesvirus 16) strain X313, a baboon herpesvirus with close homology to other primate alphaherpesviruses, such as SA8, monkey B virus, and herpes simplex virus (HSV) type 1 and type 2. The genome of HVP-2 is 156,487 bp in length, with an overall GC content of 76.5%. The genome organization is identical to that of the other members of the genus Simplexvirus, with a long and a short unique region, each bordered by inverted repeats which end with an “a” sequence. All of the open reading frames detected in this genome were homologous and colinear with those of SA8 and B virus. The HSV gene RL1 (γ134.5; neurovirulence factor) is not present in HVP-2, as is the case for SA8 and B virus. The HVP-2 genome is 85% homologous to its closest relative, SA8. However, segment-by-segment bootstrap analysis of the genome revealed at least two regions that display closer homology to the corresponding sequences of B virus. The first region comprises the UL41 to UL44 genes, and the second region is located within the UL36 gene. We hypothesize that this localized and defined shift in homology is due to recombination events between an SA8-like progenitor of HVP-2 and a herpesvirus species more closely related to the B virus. Since some of the genes involved in these putative recombination events are determinants of virulence, a comparative analysis of their function may provide insight into the pathogenic mechanism of simplexviruses. PMID:16414998

  2. The complete genome sequence of herpesvirus papio 2 (Cercopithecine herpesvirus 16) shows evidence of recombination events among various progenitor herpesviruses.

    Science.gov (United States)

    Tyler, Shaun D; Severini, Alberto

    2006-02-01

    We have sequenced the entire genome of herpesvirus papio 2 (HVP-2; Cercopithecine herpesvirus 16) strain X313, a baboon herpesvirus with close homology to other primate alphaherpesviruses, such as SA8, monkey B virus, and herpes simplex virus (HSV) type 1 and type 2. The genome of HVP-2 is 156,487 bp in length, with an overall GC content of 76.5%. The genome organization is identical to that of the other members of the genus Simplexvirus, with a long and a short unique region, each bordered by inverted repeats which end with an "a" sequence. All of the open reading frames detected in this genome were homologous and colinear with those of SA8 and B virus. The HSV gene RL1 (gamma(1)34.5; neurovirulence factor) is not present in HVP-2, as is the case for SA8 and B virus. The HVP-2 genome is 85% homologous to its closest relative, SA8. However, segment-by-segment bootstrap analysis of the genome revealed at least two regions that display closer homology to the corresponding sequences of B virus. The first region comprises the UL41 to UL44 genes, and the second region is located within the UL36 gene. We hypothesize that this localized and defined shift in homology is due to recombination events between an SA8-like progenitor of HVP-2 and a herpesvirus species more closely related to the B virus. Since some of the genes involved in these putative recombination events are determinants of virulence, a comparative analysis of their function may provide insight into the pathogenic mechanism of simplexviruses.

  3. Accidents in nuclear ships

    Energy Technology Data Exchange (ETDEWEB)

    Oelgaard, P L [Risoe National Lab., Roskilde (Denmark); [Technical Univ. of Denmark, Lyngby (Denmark)

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10{sup -3} per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au).

  4. Accidents in nuclear ships

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10 -3 per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au)

  5. Overview of core disruptive accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.

    1977-01-01

    An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

  6. SCENARIO OF AN ACCIDENT OF SOIL DAMS IN CASE OF WATER SPILL OVER A DAM CREST BY USING FAULT TREE ANALYSIS

    OpenAIRE

    Kuznetsov Dmitriy Viktorovich

    2016-01-01

    The scenario of a hydrodynamic accident of water flow over a crest of a soil dam is considered by the method of fault tree analysis, for which the basic reasons and controlled diagnostic indicators of an accident have been defined. Logical operators “AND”/”OR” were used for creation of a sequence of logically connected events, leading to an undesired event in the scenario of accident. The scenario of the accident was plotted in case of three basic reasons - an excessive settling of a dam cres...

  7. Precursors to potential severe core damage accidents: 1992, a status report; Volume 18: Appendices B, C, D, E, F, and G

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-01

    This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; {open_quote}interesting{close_quote} events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports.

  8. Application of forensic image analysis in accident investigations.

    Science.gov (United States)

    Verolme, Ellen; Mieremet, Arjan

    2017-09-01

    Forensic investigations are primarily meant to obtain objective answers that can be used for criminal prosecution. Accident analyses are usually performed to learn from incidents and to prevent similar events from occurring in the future. Although the primary goal may be different, the steps in which information is gathered, interpreted and weighed are similar in both types of investigations, implying that forensic techniques can be of use in accident investigations as well. The use in accident investigations usually means that more information can be obtained from the available information than when used in criminal investigations, since the latter require a higher evidence level. In this paper, we demonstrate the applicability of forensic techniques for accident investigations by presenting a number of cases from one specific field of expertise: image analysis. With the rapid spread of digital devices and new media, a wealth of image material and other digital information has become available for accident investigators. We show that much information can be distilled from footage by using forensic image analysis techniques. These applications show that image analysis provides information that is crucial for obtaining the sequence of events and the two- and three-dimensional geometry of an accident. Since accident investigation focuses primarily on learning from accidents and prevention of future accidents, and less on the blame that is crucial for criminal investigations, the field of application of these forensic tools may be broader than would be the case in purely legal sense. This is an important notion for future accident investigations. Copyright © 2017 Elsevier B.V. All rights reserved.

  9. Performance of Earthquake Early Warning Systems during the Major Events of the 2016-2017 Central Italy Seismic Sequence.

    Science.gov (United States)

    Festa, G.; Picozzi, M.; Alessandro, C.; Colombelli, S.; Cattaneo, M.; Chiaraluce, L.; Elia, L.; Martino, C.; Marzorati, S.; Supino, M.; Zollo, A.

    2017-12-01

    Earthquake early warning systems (EEWS) are systems nowadays contributing to the seismic risk mitigation actions, both in terms of losses and societal resilience, by issuing an alert promptly after the earthquake origin and before the ground shaking impacts the targets to be protected. EEWS systems can be grouped in two main classes: network based and stand-alone systems. Network based EEWS make use of dense seismic networks surrounding the fault (e.g. Near Fault Observatory; NFO) generating the event. The rapid processing of the P-wave early portion allows for the location and magnitude estimation of the event then used to predict the shaking through ground motion prediction equations. Stand-alone systems instead analyze the early P-wave signal to predict the ground shaking carried by the late S or surface waves, through empirically calibrated scaling relationships, at the recording site itself. We compared the network-based (PRESTo, PRobabilistic and Evolutionary early warning SysTem, www.prestoews.org, Satriano et al., 2011) and the stand-alone (SAVE, on-Site-Alert-leVEl, Caruso et al., 2017) systems, by analyzing their performance during the 2016-2017 Central Italy sequence. We analyzed 9 earthquakes having magnitude 5.0 security actions. PRESTo also evaluated the accuracy of location and magnitude. Both systems well predict the ground shaking nearby the event source, with a success rate around 90% within the potential damage zone. The lead-time is significantly larger for the network based system, increasing to more than 10s at 40 km from the event epicentre. The stand-alone system better performs in the near-source region showing a positive albeit small lead-time (operational in Italy, based on the available acceleration networks, by improving the capability of reducing the lead-time related to data telemetry.

  10. Danngarrd-Oscar events recorded in a terrestrial sequence in central British Columbia, Canada

    Science.gov (United States)

    Ward, B. C.; Geertsema, M.; Telka, A.; Mathewes, R.

    2012-12-01

    Danngarrd-Oscar events recorded in the GISP2 Greenland Ice Core. The increasingly dry and cold conditions indicated by the macrofossil assemblage likely reflect the growth of ice in the Coast Mountains that would reduce the availability of moisture to the Interior Plateau from Pacific air masses. This is confirmed by reconstruction of the growth of the Cordilleran Ice Sheet during the Late Wisconsinan based on published radiocarbon dates.

  11. Complete nucleotide sequence of CTX-M-15-plasmids from clinical Escherichia coli isolates: insertional events of transposons and insertion sequences.

    Directory of Open Access Journals (Sweden)

    Annemieke Smet

    Full Text Available BACKGROUND: CTX-M-producing Escherichia coli strains are regarded as major global pathogens. METHODOLOGY/PRINCIPAL FINDINGS: The nucleotide sequence of three plasmids (pEC_B24: 73801-bp; pEC_L8: 118525-bp and pEC_L46: 144871-bp from Escherichia coli isolates obtained from patients with urinary tract infections and one plasmid (pEC_Bactec: 92970-bp from an Escherichia coli strain isolated from the joint of a horse with arthritis were determined. Plasmid pEC_Bactec belongs to the IncI1 group and carries two resistance genes: bla(TEM-1 and bla(CTX-M-15. It shares more than 90% homology with a previously published bla(CTX-M-plasmid from E. coli of human origin. Plasmid pEC_B24 belongs to the IncFII group whereas plasmids pEC_L8 and pEC_L46 represent a fusion of two replicons of type FII and FIA. On the pEC_B24 backbone, two resistance genes, bla(TEM-1 and bla(CTX-M-15, were found. Six resistance genes, bla(TEM-1, bla(CTX-M-15, bla(OXA-1, aac6'-lb-cr, tetA and catB4, were detected on the pEC_L8 backbone. The same antimicrobial drug resistance genes, with the exception of tetA, were also identified on the pEC_L46 backbone. Genome analysis of all 4 plasmids studied provides evidence of a seemingly frequent transposition event of the bla(CTX-M-15-ISEcp1 element. This element seems to have a preferred insertion site at the tnpA gene of a bla(TEM-carrying Tn3-like transposon, the latter itself being inserted by a transposition event. The IS26-composite transposon, which contains the bla(OXA-1, aac6'-lb-cr and catB4 genes, was inserted into plasmids pEC_L8 and pEC_L46 by homologous recombination rather than a transposition event. Results obtained for pEC_L46 indicated that IS26 also plays an important role in structural rearrangements of the plasmid backbone and seems to facilitate the mobilisation of fragments from other plasmids. CONCLUSIONS: Collectively, these data suggests that IS26 together with ISEcp1 could play a critical role in the evolution of

  12. Genetic plasticity of the Shigella virulence plasmid is mediated by intra- and inter-molecular events between insertion sequences.

    Science.gov (United States)

    Pilla, Giulia; McVicker, Gareth; Tang, Christoph M

    2017-09-01

    Acquisition of a single copy, large virulence plasmid, pINV, led to the emergence of Shigella spp. from Escherichia coli. The plasmid encodes a Type III secretion system (T3SS) on a 30 kb pathogenicity island (PAI), and is maintained in a bacterial population through a series of toxin:antitoxin (TA) systems which mediate post-segregational killing (PSK). The T3SS imposes a significant cost on the bacterium, and strains which have lost the plasmid and/or genes encoding the T3SS grow faster than wild-type strains in the laboratory, and fail to bind the indicator dye Congo Red (CR). Our aim was to define the molecular events in Shigella flexneri that cause loss of Type III secretion (T3S), and to examine whether TA systems exert positional effects on pINV. During growth at 37°C, we found that deletions of regions of the plasmid including the PAI lead to the emergence of CR-negative colonies; deletions occur through intra-molecular recombination events between insertion sequences (ISs) flanking the PAI. Furthermore, by repositioning MvpAT (which belongs to the VapBC family of TA systems) near the PAI, we demonstrate that the location of this TA system alters the rearrangements that lead to loss of T3S, indicating that MvpAT acts both globally (by reducing loss of pINV through PSK) as well as locally (by preventing loss of adjacent sequences). During growth at environmental temperatures, we show for the first time that pINV spontaneously integrates into different sites in the chromosome, and this is mediated by inter-molecular events involving IS1294. Integration leads to reduced PAI gene expression and impaired secretion through the T3SS, while excision of pINV from the chromosome restores T3SS function. Therefore, pINV integration provides a reversible mechanism for Shigella to circumvent the metabolic burden imposed by pINV. Intra- and inter-molecular events between ISs, which are abundant in Shigella spp., mediate plasticity of S. flexneri pINV.

  13. Trending analysis of precursor events

    International Nuclear Information System (INIS)

    Watanabe, Norio

    1998-01-01

    The Accident Sequence Precursor (ASP) Program of United States Nuclear Regulatory Commission (U.S.NRC) identifies and categorizes operational events at nuclear power plants in terms of the potential for core damage. The ASP analysis has been performed on yearly basis and the results have been published in the annual reports. This paper describes the trends in initiating events and dominant sequences for 459 precursors identified in the ASP Program during the 1969-94 period and also discusses a comparison with dominant sequences predicted in the past Probabilistic Risk Assessment (PRA) studies. These trends were examined for three time periods, 1969-81, 1984-87 and 1988-94. Although the different models had been used in the ASP analyses for these three periods, the distribution of precursors by dominant sequences show similar trends to each other. For example, the sequences involving loss of both main and auxiliary feedwater were identified in many PWR events and those involving loss of both high and low coolant injection were found in many BWR events. Also, it was found that these dominant sequences were comparable to those determined to be dominant in the predictions by the past PRAs. As well, a list of the 459 precursors identified are provided in Appendix, indicating initiating event types, unavailable systems, dominant sequences, conditional core damage probabilities, and so on. (author)

  14. First Responders and Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  15. 22 CFR 102.8 - Reporting accidents.

    Science.gov (United States)

    2010-04-01

    ... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Reporting accidents. 102.8 Section 102.8... Accidents Abroad § 102.8 Reporting accidents. (a) To airline and Civil Aeronautics Administration... probably be the first to be informed of the accident, in which event he will be expected to report the...

  16. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report

  17. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Main report (Chapters 7--12). Volume 2, Part 1B

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specific shutdown accidents would be useful

  18. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices E (Sections E.1--E.8). Volume 2, Part 3A

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful

  19. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Main report (Chapters 1--6). Volume 2, Part 1A

    International Nuclear Information System (INIS)

    Chu, T.L.; Musicki, Z.; Kohut, P.

    1992-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown written specifically for shutdown accidents would be useful. This document presents Chapters 1--6 of the report

  20. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1994-06-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report.

  1. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  2. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations. Volume 1: Sections 1-9

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report

  3. Lessons from the Fukushima nuclear power accident

    International Nuclear Information System (INIS)

    Hatamura, Yotaro

    2013-01-01

    Through the investigation of the Fukushima Nuclear Power Accident as the chairman of the related Government's Committee, many things had been considered. Essence of the accident could be not only what occurred in the Fukushima nuclear power station, but also dispersed radioactive materials forced many residents to move and not to be returned. Such events as indication errors of water level meter occurring in severe accident could no be thought and remote mechanical operation of valves under high radiation environment were not prepared. Contamination by radioactive clouds caused the evacuation of residents for a long period. Lessons learned from the accident were described such as; (1) the verification of the road to failure connecting selected accident sequence and road to success with another supposed choice, (2) considering what might occur and then what should be needed on the contrary, (3) nuclear power, if should be continued, should be used with the premise of its hazards, and (4) advise to nuclear engineer for adequate information dissemination and technical explanation to the public and keeping nuclear technologies alive. (T. Tanaka)

  4. A High-yield Fall Risk and Adverse Events Screening Questions From the Stopping Elderly Accidents, Death, and Injuries (STEADI) Guideline for Older Emergency Department Fall Patients.

    Science.gov (United States)

    Sri-On, Jiraporn; Tirrell, Gregory Philip; Kamsom, Anucha; Marill, Keith A; Shankar, Kalpana Narayan; Liu, Shan W

    2018-03-25

    The objectives were to examine whether responses to the Stopping Elderly Accidents, Death, and Injuries (STEADI) questions responses predicted adverse events after an older adult emergency department (ED) fall visits and to identify factors associated with such recurrent fall. We conducted a prospective study at two urban, teaching hospitals. We included patients aged ≥ 65 years who presented to the ED for an accidental fall. Data were gathered for fall-relevant comorbidities, high-risk medications for falls, and the responses to 12 questions from the STEADI guideline recommendation. Our outcomes were the number of 6-month adverse events that were defined as mortality, ED revisit, subsequent hospitalization, recurrent falls, and a composite outcome. There were 548 (86.3%) patients who completed follow-up and 243 (44.3%) patients experienced an adverse event after a fall within 6 months. In multivariate analysis, seven questions from the STEADI guideline predicted various outcomes. The question "Had previous fall" predicted recurrent falls (odds ratio [OR] = 2.45, 95% confidence interval [CI] = 1.52 to 3.97), the question "Feels unsteady when walking sometimes" (OR = 2.34, 95% CI = 1.44 to 3.81), and "Lost some feeling in their feet" predicted recurrent falls. In addition to recurrent falls risk, the supplemental questions "Use or have been advised to use a cane or walker," "Take medication that sometimes makes them feel light-headed or more tired than usual," "Take medication to help sleep or improve mood," and "Have to rush to a toilet" predicted other outcomes. A STEADI score of ≥4 did not predict adverse outcomes although seven individual questions from the STEADI guidelines were associated with increased adverse outcomes within 6 months. These may be organized into three categories (previous falls, physical activity, and high-risk medications) and may assist emergency physicians to evaluate and refer high-risk fall patients for a comprehensive

  5. 3-D Dynamic rupture simulation for the 2016 Kumamoto, Japan, earthquake sequence: Foreshocks and M6 dynamically triggered event

    Science.gov (United States)

    Ando, R.; Aoki, Y.; Uchide, T.; Imanishi, K.; Matsumoto, S.; Nishimura, T.

    2016-12-01

    A couple of interesting earthquake rupture phenomena were observed associated with the sequence of the 2016 Kumamoto, Japan, earthquake sequence. The sequence includes the April 15, 2016, Mw 7.0, mainshock, which was preceded by multiple M6-class foreshock. The mainshock mainly broke the Futagawa fault segment striking NE-SW direction extending over 50km, and it further triggered a M6-class earthquake beyond the distance more than 50km to the northeast (Uchide et al., 2016, submitted), where an active volcano is situated. Compiling the data of seismic analysis and InSAR, we presumed this dynamic triggering event occurred on an active fault known as Yufuin fault (Ando et al., 2016, JPGU general assembly). It is also reported that the coseismic slip was significantly large at a shallow portion of Futagawa Fault near Aso volcano. Since the seismogenic depth becomes significantly shallower in these two areas, we presume the geothermal anomaly play a role as well as the elasto-dynamic processes associated with the coseismic rupture. In this study, we conducted a set of fully dynamic simulations of the earthquake rupture process by assuming the inferred 3D fault geometry and the regional stress field obtained referring the stress tensor inversion. As a result, we showed that the dynamic rupture process was mainly controlled by the irregularity of the fault geometry subjected to the gently varying regional stress field. The foreshocks ruptures have been arrested at the juncture of the branch faults. We also show that the dynamic triggering of M-6 class earthquakes occurred along the Yufuin fault segment (located 50 km NE) because of the strong stress transient up to a few hundreds of kPa due to the rupture directivity effect of the M-7 event. It is also shown that the geothermal condition may lead to the susceptible condition of the dynamic triggering by considering the plastic shear zone on the down dip extension of the Yufuin segment, situated in the vicinity of an

  6. Strategic crisis and risk communication during a prolonged natural hazard event: lessons learned from the Canterbury earthquake sequence

    Science.gov (United States)

    Wein, A. M.; Potter, S.; Becker, J.; Doyle, E. E.; Jones, J. L.

    2015-12-01

    While communication products are developed for monitoring and forecasting hazard events, less thought may have been given to crisis and risk communication plans. During larger (and rarer) events responsible science agencies may find themselves facing new and intensified demands for information and unprepared for effectively resourcing communications. In a study of the communication of aftershock information during the 2010-12 Canterbury Earthquake Sequence (New Zealand), issues are identified and implications for communication strategy noted. Communication issues during the responses included reliability and timeliness of communication channels for immediate and short decision time frames; access to scientists by those who needed information; unfamiliar emergency management frameworks; information needs of multiple audiences, audience readiness to use the information; and how best to convey empathy during traumatic events and refer to other information sources about what to do and how to cope. Other science communication challenges included meeting an increased demand for earthquake education, getting attention on aftershock forecasts; responding to rumor management; supporting uptake of information by critical infrastructure and government and for the application of scientific information in complex societal decisions; dealing with repetitive information requests; addressing diverse needs of multiple audiences for scientific information; and coordinating communications within and outside the science domain. For a science agency, a communication strategy would consider training scientists in communication, establishing relationships with university scientists and other disaster communication roles, coordinating messages, prioritizing audiences, deliberating forecasts with community leaders, identifying user needs and familiarizing them with the products ahead of time, and practicing the delivery and use of information via scenario planning and exercises.

  7. Analysis of Three Mile Island-Unit 2 accident

    International Nuclear Information System (INIS)

    1980-03-01

    The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute has analyzed the Three Mile Island-2 accident. Early results of this analysis were a brief narrative summary, issued in mid-May 1979 and an initial version of this report issued later in 1979 as noted in the Foreword. The present report is a revised version of the 1979 report, containing summaries, a highly detailed sequence of events, a comparison of that sequence of events with those from other sources, 25 appendices, references and a list of abbreviations and acronyms. A matrix of equipment and system actions is included as a folded insert

  8. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation

    International Nuclear Information System (INIS)

    Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

    2010-01-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  9. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  10. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  11. The Chernobyl accident

    International Nuclear Information System (INIS)

    Berg, J.O.; Christensen, G.; Lingjaerde, R.; Smidt Olsen, H.; Wethe, P.I.

    1986-10-01

    In connection with the Chernobyl accident the report gives a description of the technical features of importance to the accident, the course of events, and the estimated health hazards in the local environment. Dissimilarities in western and Sovjet reactor safety philosophy are dealt with, as well as conceivable concequences in relation to technology and research in western nuclear power programmes. Results of activity level measurements of air and foodstuff, made in Norway by Institute for Energy Technology, are given

  12. Chernobyl accident and Denmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. The event at the accident site, the release and dispersal of radioactive substances into the atmosphere and over Europe, is described. A discussion of the Danish organisation for nuclear emergencies, how it was activated and adapted to the actual situation, is given. A comprehensive description of the radiological contamination in Denmark following the accident and the estimated health effects, is presented. The situation in other European countries is mentioned. (author)

  13. Probabilistic risk assessment using event tables and the BNL [Brookhaven National Laboratory] event-tree analyzer

    International Nuclear Information System (INIS)

    Fullwood, R.R.; Shier, W.G.

    1989-01-01

    Probabilistic risk analysis (PRA) is being used to study design alternatives for the advanced neutron source research reactor being designed at Oak Ridge National Laboratory for operation in the 1990s. Major communication paths between the designers and the safety analysts are accident discussions supported by event tables, event-tree graphics, and accident sequence probabilities. The BETA code used in conjunction with a word processor provides this linkage. This paper describes the process, features of the BETA, how it works, and some examples of usage

  14. Assessment of accident risks in the CRBRP. Volume 2. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-03-01

    Appendices to Volume I include core-related accident-sequence definition, CRBRP risk-assessment sequence-probability determinations, failure-probability data, accident scenario evaluation, radioactive material release analysis, ex-core accident analysis, safety philosophy and design features, calculation of reactor accident consequences, sensitivity study, and risk from fires.

  15. Severe accident considerations in Canadian nuclear power reactors

    International Nuclear Information System (INIS)

    Omar, A.M.; Measures, M.P.; Scott, C.K.; Lewis, M.J.

    1990-08-01

    This paper describes a current study on severe accidents being sponsored by the Atomic Energy Control Board (AECB) and provides background on other related Canadian work. Scoping calculations are performed in Phase I of the AECB study to establish the relative consequences of several permutations resulting from six postulated initiating events, nine containment states, and a selection of meteorological conditions and health effects mitigating criteria. In Phase II of the study, selected accidents sequences would be analyzed in detail using models suitable for the design features of the Canadian nuclear power reactors

  16. Development of A Methodology for Assessing Various Accident Management Strategies Using Decision Tree Models

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Nam Yeong; Kim, Jin Tae; Jae, Moo Sung [Hanyang University, Seoul (Korea, Republic of); Jerng, Dong Wook [Chung-Ang University, Seoul (Korea, Republic of)

    2016-05-15

    The purpose of ASP (Accident Sequence Precursor) analysis is to evaluate operational accidents in full power and low power operation by using PRA (Probabilistic Risk Assessment) technologies. The awareness of the importance of ASP analysis has been on rise. The methodology for ASP analysis has been developed in Korea, KINS (Korea Institute of Nuclear Safety) has managed KINS-ASP program since it was developed. In this study, we applied ASP analysis into operational accidents in full power and low power operation to quantify CCDP (Conditional Core Damage Probability). To reflect these 2 cases into PRA model, we modified fault trees and event trees of the existing PRA model. Also, we suggest the ASP regulatory system in the conclusion. In this study, we reviewed previous studies for ASP analysis. Based on it, we applied it into operational accidents in full power and low power operation. CCDP of these 2 cases are 1.195E-06 and 2.261E-03. Unlike other countries, there is no regulatory basis of ASP analysis in Korea. ASP analysis could detect the risk by assessing the existing operational accidents. ASP analysis can improve the safety of nuclear power plant by detecting, reviewing the operational accidents, and finally removing potential risk. Operator have to notify regulatory institute of operational accident before operator takes recovery work for the accident. After follow-up accident, they have to check precursors in data base to find similar accident.

  17. Identification of the operating crew's information needs for accident management

    International Nuclear Information System (INIS)

    Nelson, W.R.; Hanson, D.J.; Ward, L.W.; Solberg, D.E.

    1988-01-01

    While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe-accident research results to enhance the capability of the plant operating crew to effectively manage severe accidents. One element of this program includes identification of the information needed by the operating crew in severe-accident situations. This paper discusses a method developed for identifying these information needs and its application. The methodology has been applied to a generic reactor design representing a PWR with a large dry containment. The information needs were identified by systematically determining what information is needed to assess the health of the critical functions, identify the presence of challenges, select strategies, and assess the effectiveness of these strategies. This method allows the systematic identification of information needs for a broad range of severe-accident scenarios and can be validated by exercising the functional models for any specific event sequence

  18. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

  19. Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jae Ho; Kim, Hyun Gil; In, Wang Kee; Kim, Weon Ju; Koo, Yang Hyum [KAERI, Daejeon (Korea, Republic of); Lee, Seung Jae [KEPCONF, Daejeon (Korea, Republic of)

    2016-05-15

    Research on accident tolerant fuels (ATFs) is aimed at developing innovative fuels, which can mitigate or prevent the consequences of accidents. In Korea, innovative concepts are being developed to improve fuel safety and reliability of LWRs during accident events and normal operations. ATF technologies will be developed and commercialized through a sequence of long-lead and extensive activities. The interim milestone for new fuel program is that we would be ready for an irradiation test in commercial reactor by 2021. This presentation deals with the status of ATF development in KOREA and plan to implement new fuel technology successfully in commercial nuclear power plants.

  20. Some critical remarks on a sequence of events interpreted to possibly originate from a decay chain of an element 120 isotope

    Energy Technology Data Exchange (ETDEWEB)

    Hessberger, F.P. [GSI - Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Helmholtz-Institut Mainz, Mainz (Germany); Ackermann, D. [GANIL, Caen (France)

    2017-06-15

    A sequence of three events observed in an irradiation of {sup 248}Cm with {sup 54}Cr at the velocity filter SHIP of the GSI - Helmholtzzentrum fuer Schwerionenforschung GmbH, 64291 Darmstadt, Germany, had been interpreted as a decay chain consisting of three α particles. On the basis of measured energies, a possible assignment to the decay of an isotope of element 120 was discussed, although it was stated that a definite assignment could not be made. A critical analysis of the data, however, shows that the reported events do not have the properties of a decay chain consisting of three α particles and (probably being terminated by) a spontaneous fission event, but that this is rather a random sequence of events. (orig.)

  1. Accident management approach in Armenia

    International Nuclear Information System (INIS)

    Ghazaryan, K.

    1999-01-01

    In this lecture the accident management approach in Armenian NPP (ANPP) Unit 2 is described. List of BDBAs had been developed by OKB Gydropress in 1994. 13 accident sequences were included in this list. The relevant analyses had been performed in VNIIAES and the 'Guidelines on operator actions for beyond design basis accident (BDBA) management at ANPP Unit 2' had been prepared. These instructions are discussed

  2. Accident management information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs

  3. Accident management information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs.

  4. Technical basis document for external events

    International Nuclear Information System (INIS)

    OBERG, B.D.

    2003-01-01

    This document supports the Tank Farms Documented Safety Analysis and presents the technical basis for the FR-equencies of externally initiated accidents. The consequences of externally initiated events are discussed in other documents that correspond to the accident that was caused by the external event. The external events include aircraft crash, vehicle accident, range fire, and rail accident

  5. Accident Assessment

    International Nuclear Information System (INIS)

    Tripputi, Ivo; Lund, Ingemar

    2002-01-01

    There is a general feeling that decommissioning is an activity involving limited risks, compared to NPP operation, and in particular risks involving the general public. This is technically confirmed by licensing analysis and evaluations, where, once the spent fuel has been removed from the plant, the radioactivity inventory available to be released to the environment is very limited. Decommissioning activities performed so far in the world have also confirmed the first assumptions and no specific issue has been identified, in this field, to justify a completely new approach. Commercial interests in international harmonization, which could drive an in-depth discussion about the bases of this approach, are weak at the moment. However, there are several reasons why a discussion in an international framework about the Safety Case for decommissioning (and, in particular, about Accident Assessment) may be considered necessary and important, and why it may show some specific and peculiar aspects. An effort for a comprehensive and systematic D and D accident safety assessment of the decommissioning process is justified. It is necessary also to explore in a holistic way the aspects of industrial safety, and develop tools for the decision-making process optimization. The expected results are the implementation of appropriate and optimized protective measures in any event and of adequate on/off-site emergency plans for optimal public and workers protection. The experience from other decommissioning projects and large-scale industrial activities is essential to balance provisions and an Operating Experience review process (specific for decommissioning) should help to focus on real issues

  6. Accident management for severe accidents

    International Nuclear Information System (INIS)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

  7. Biological effects of ionizing radiations. Radiological accident from Goiania, GO, Brazil

    International Nuclear Information System (INIS)

    Okuno, Emico

    2013-01-01

    This article presents the fundaments of radiation physics, the natural and artificial sources, biological effects, radiation protection. We also examine the sequence of events that resulted in Goiania accident with a source of caesium-137 from abandoned radiotherapy equipment and its terrible consequences. (author)

  8. Cyclic Sequences, Events and Evolution of the Sino-Korean Plate,with a Discussion on the Evolution of Molar-tooth Carbonates,Phosphorites and Source Rocks

    Institute of Scientific and Technical Information of China (English)

    MENG Xianghua; GE Ming

    2003-01-01

    This paper gives an account of the research that the authors conducted on the cyclic sequences, events and evolutionary history from Proterozoic to Meso-Cenozoic in the Sino-Korean plate based on the principle of the Cosmos-Earth System. The authors divided this plate into 20 super-cyclic or super-mega-cyclic periods and more than 100 Oort periods. The research focused on important sea flooding events, uplift interruption events, tilting movement events, molar-tooth carbonate events, thermal events, polarity reversal events, karst events, volcanic explosion events and storm events, as well as types of resource areas and paleotectonic evolution. By means of the isochronous theory of the Cosmos-Earth System periodicity and based on long-excentricity and periodicity, the authors elaborately studied the paleogeographic evolution of the aulacogen of the Sino-Korean plate, the oolitic beach platform formation, the development of foreland basin and continental rift valley basin, and reconstructed the evolution of tectonic paleogeography and stratigraphic framework in the Sino-Korean plate in terms of evolutionary maps. Finally, the authors gave a profound discussion on the formation and development of molar-tooth carbonates, phosphorites and source rocks.

  9. Studies of potential severe accidents in Finnish nuclear power plants. Quarterly report 3. quarter 1987

    International Nuclear Information System (INIS)

    Aro, Ilari.

    1989-07-01

    This thesis is based on six publications dealing with severe accident studies in Finnish nuclear power plants. Main emphasis has been put on general technical bases and methodologies applied in severe accident evaluation in Finland. As an example of the use of the analysis and evaluation methods, the analysis of one representative accident sequence, t otal loss of AC power , has been presented for both Finnish power plant types. This accident sequence is required to be analyzed in the Finnish safety guide YVL 2.2 which deals with transient and accident analyses as a basis of technical solutions at nuclear powr plants. Two different analysis methods, MAAP 3.0 and MARCH 3/STCP have been used for receiving as complete a picture as possible of the flow of events and for verifying the models to some extent. Besides the use of the two different models, the method of sensitivity analysis has been used for evaluating the effects of some important technical parameters on the accident flow. Finally, conclusions of the applicability of the two methods for analyzing severe accident sequences in Finnish plants have been discussed

  10. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  11. Persistence of airline accidents.

    Science.gov (United States)

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

  12. Limitations of systemic accident analysis methods

    Directory of Open Access Journals (Sweden)

    Casandra Venera BALAN

    2016-12-01

    Full Text Available In terms of system theory, the description of complex accidents is not limited to the analysis of the sequence of events / individual conditions, but highlights nonlinear functional characteristics and frames human or technical performance in relation to normal functioning of the system, in safety conditions. Thus, the research of the system entities as a whole is no longer an abstraction of a concrete situation, but an exceeding of the theoretical limits set by analysis based on linear methods. Despite the issues outlined above, the hypothesis that there isn’t a complete method for accident analysis is supported by the nonlinearity of the considered function or restrictions, imposing a broad vision of the elements introduced in the analysis, so it can identify elements corresponding to nominal parameters or trigger factors.

  13. Containment performance evaluation for the GESSAR-II plant for seismic initiating events

    International Nuclear Information System (INIS)

    Shiu, K.K.; Chu, T.; Ludewig, H.; Pratt, W.T.

    1986-01-01

    As a part of the overall effort undertaken by Brookhaven National Laboratory (BNL) to review the GESSAR-II probabilistic risk assessment, an independent containment performance evaluation was performed using the containment event tree approach. This evaluation focused principally on those accident sequences which are initiated by seismic events. This paper reports the findings of this study. 1 ref

  14. Quantification of sequence exchange events between PMS2 and PMS2CL provides a basis for improved mutation scanning of Lynch syndrome patients.

    NARCIS (Netherlands)

    Klift, H.M. van der; Tops, C.M.; Bik, E.C.; Boogaard, M.W.; Borgstein, A.M.; Hansson, K.B.; Ausems, M.G.E.M.; Gomez Garcia, E.; Green, A.; Hes, F.J.; Izatt, L.; Hest, L.P. van; Alonso, A.M.; Vriends, A.H.; Wagner, A.; Zelst-Stams, W.A.G. van; Vasen, H.F.; Morreau, H.; Devilee, P.; Wijnen, J.T.

    2010-01-01

    Heterozygous mutations in PMS2 are involved in Lynch syndrome, whereas biallelic mutations are found in Constitutional mismatch repair-deficiency syndrome patients. Mutation detection is complicated by the occurrence of sequence exchange events between the duplicated regions of PMS2 and PMS2CL. We

  15. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of excedeence of damage by integrated Safety Analysis Methodology

    International Nuclear Information System (INIS)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-01-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Excedence Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  16. Probabilistic risk assessment for the Los Alamos Meson Physics Facility worst-case design-basis accident

    International Nuclear Information System (INIS)

    Sharirli, M.; Butner, J.M.; Rand, J.L.; Macek, R.J.; McKinney, S.J.; Roush, M.L.

    1992-01-01

    This paper presents results from a Los Alamos National Laboratory Engineering and Safety Analysis Group assessment of the worse-case design-basis accident associated with the Clinton P. Anderson Meson Physics Facility (LAMPF)/Weapons Neutron Research (WNR) Facility. The primary goal of the analysis was to quantify the accident sequences that result in personnel radiation exposure in the WNR Experimental Hall following the worst-case design-basis accident, a complete spill of the LAMPF accelerator 1L beam. This study also provides information regarding the roles of hardware systems and operators in these sequences, and insights regarding the areas where improvements can increase facility-operation safety. Results also include confidence ranges to incorporate combined effects of uncertainties in probability estimates and importance measures to determine how variations in individual events affect the frequencies in accident sequences

  17. Safety and risk questions following the nuclear incidents and accidents in Japan. Summary final report

    International Nuclear Information System (INIS)

    Mildenberger, Oliver

    2015-03-01

    After the nuclear accidents in Japan, GRS has carried out in-depth investigations of the events. On the one hand, the accident sequences in the affected units have been analysed from various viewpoints. On the other hand, the transferability of the findings to German plants has been examined to possibly make recommendations for safety improvements. The accident sequences at Fukushima Daiichi have been traced with as much detail as possible based on all available information. Additional insights have been drawn from thermohydraulic analyses with the GRS code system ATHLET-CD/COCOSYS focusing on the events in units 2 and 3, e.g. with regard to core damage and the state of the containments in the first days of the accident sequence. In-depth investigations have also been carried out on topics such as natural external hazards, electrical power supply or organizational measures. In addition, methodological studies on further topics related with the accidents have been performed. Through a detailed analysis of the relevant data from the events in Japan, the basis for an in-depth examination of the transferability to German plants was created. It was found that an implementation of most of the insights gained from the investigations had already been initiated as part of the GRS information notice 2012/02. Further findings have been communicated to the federal government and introduced into other relevant bodies, e.g. the Nuclear Safety Standards Committee (KTA) or the Reactor Safety Commission (RSK).

  18. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  19. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    International Nuclear Information System (INIS)

    Watanabe, Norio; Tamaki, Hitoshi

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  20. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio [Planning and Analysis Division, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  1. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  2. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  3. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  4. Development and application of techniques to assist in the establishment of intervention levels for the introduction of countermeasures in the event of an accident

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.; Konstantinov, Y.O.; Demin, V.F.

    1993-02-01

    International and CIS (Commenwealth of independent states) guidance on intervention after nuclear or radiological accidents are reviewed, both the guidance existing at the time of the Chernobyl accident and the present guidance from ICRP, IAEA, OECD/NEA, WHO, CAC and CEC. The new thinking on intervention policy in the CIS is briefly addressed. The basic justification/optimisation principles for intervention as recommended by the international organisations have been applied for the longer term protective measures of relocation and foodstuff restrictions to derive intervention levels based on averted doses and monetary costs. Numerical values of the Intervention Levels as well as Operational Intervention Levels are presented. (au) (23 refs.)

  5. Medical procedures in the event of nuclear power plant accidents. Guidelines for: Medical consultants for emergency response commander; physicians in emergency care centres; physicians in outpatient and inpatient care

    International Nuclear Information System (INIS)

    Genkel, Simone

    2008-01-01

    The author of the contribution under consideration reports on medical procedures in the event of nuclear power plant accidents. This contribution consists of the following sections: protective measures, tasks of radiation protection physicians, emergency care centres. It has been pointed out that differentiation of the hospitals is acquired which accept radiation accident patients. However, only a small number of hospitals will be able to professionally treat patients with suspected gastrointestinal or pronounced (muco)cutaneous type of hospitals with haemotological-oncological departments. Thus they should be able to treat patients who have been exposed to radiation doses between 1 and 6 Gy without any difficulties. Even larger is the number of hospitals which can accept patients who were exposed to a radiation dose of less than 1 Gy, but suffer from other complicating diseases (injuries, general diseases)

  6. Radiological accidents: education for prevention and confrontation

    International Nuclear Information System (INIS)

    Cardenas Herrera, Juan; Fernandez Gomez, Isis Maria

    2008-01-01

    The purpose of this work is to train and inform on radiological accidents as a preventive measure to improve the people life quality. Radiological accidents are part of the events of technological origin which are composed of nuclear and radiological accidents. As a notable figure is determined that there have been 423 radiological accidents from 1944 to 2005 and among the causes prevail industrial accidents, by irradiations, medical accidents and of laboratories, among others. Latin American countries such as Argentina, Brazil, Mexico and Peru are some where most accidents have occurred by radioactivity. The radiological accidents can have sociological, environmental, economic, social and political consequences. In addition, there are scenarios of potential nuclear accidents and in them the potential human consequences. Also, the importance of the organization and planning in a nuclear emergency is highlighted. Finally, the experience that Cuba has lived on the subject of radiological accidents is described [es

  7. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  8. Using Pennsylvania's Three Mile Island Accident as a Case Study to Analyze Newspaper Coverage: A Diary of Events and Suggestions for Teaching Strategies.

    Science.gov (United States)

    Susskind, Jacob L.

    1983-01-01

    Methods for studying the coverage of the same current news story in several newspapers are outlined. Secondary school students critically examine news reporting, detect false or propagandistic reports, and learn to weigh and judge evidence. An example using the Three Mile Island nuclear accident is provided. (KC)

  9. North Wales Group report on the effects of the Chernobyl accident

    International Nuclear Information System (INIS)

    1987-11-01

    A report is presented by the North Wales Group concerning the sequence of events affecting North Wales and the identification of the residual problems following contamination from the Chernobyl accident. The first part of the report attempts to establish a time scale for radiation restrictions applicable in North Wales and the size of the areas which are involved. Part two deals with national arrangements to handle incidents like Chernobyl and examines the wider field of international arrangements. A review is given of events as seen by the affected community following the Chernobyl accident. (U.K.)

  10. Safety related studies on the accident behaviour of the HTR-100

    International Nuclear Information System (INIS)

    Wolters, J.; Mertens, J.; Altes, J.; Bongartz, R.; Breitbach, G.; David, P.H.; Degen, G.; Ehrlich, H.G.; Escherich, K.H.; Frank, E.; Hennings, W.; Jahn, W.; Koschmieder, R.; Marx, J.; Meister, G.; Moormann, R.; Rehm, W.; Verfondern, K.

    1991-10-01

    The aim of investigations was to verify the safety concept of the plant for balance and to quantify the radiological risk to be expected in operating an HTR-100 double unit system. Moreover, aspects of the investment risk were considered. The spectrum of initiating events ranged from so-called transients to leaks in the primary circuit and steam generator and even included earthquakes. Some of the event trees derived were highly complex and extensive due to the situation of the steam generator above the core and with regard to the double unit plant concept with increased possibilities of accident control, but also with respect to potential accident propagation. Correspondingly sophisticated analyses were required to identify risk-relevant event sequences. Environmental exposure for all risk-relevant accidents is so low that accident consequence calculations do not reveal any lethal radiation doses and practically no stochastic fatal injuries. These calculations neither assumed acute protective measures nor long-term resettlement or decontamination. The radiological risk caused by an HTR-100 plant is therefore to be classified as very low. The initiating events selected as representative and the event sequences studied in detail cover the risk-relevant event spectrum well into the hypothetical range. (orig./HP) [de

  11. Genome Sequence Analysis of New Isolates of the Winona Strain of Plum pox virus and the First Definitive Evidence of Intrastrain Recombination Events.

    Science.gov (United States)

    James, Delano; Sanderson, Dan; Varga, Aniko; Sheveleva, Anna; Chirkov, Sergei

    2016-04-01

    Plum pox virus (PPV) is genetically diverse with nine different strains identified. Mutations, indel events, and interstrain recombination events are known to contribute to the genetic diversity of PPV. This is the first report of intrastrain recombination events that contribute to PPV's genetic diversity. Fourteen isolates of the PPV strain Winona (W) were analyzed including nine new strain W isolates sequenced completely in this study. Isolates of other strains of PPV with more than one isolate with the complete genome sequence available in GenBank were included also in this study for comparison and analysis. Five intrastrain recombination events were detected among the PPV W isolates, one among PPV C strain isolates, and one among PPV M strain isolates. Four (29%) of the PPV W isolates analyzed are recombinants; one of which (P2-1) is a mosaic, with three recombination events identified. A new interstrain recombinant event was identified between a strain M isolate and a strain Rec isolate, a known recombinant. In silico recombination studies and pairwise distance analyses of PPV strain D isolates indicate that a threshold of genetic diversity exists for the detectability of recombination events, in the range of approximately 0.78×10(-2) to 1.33×10(-2) mean pairwise distance. RDP4 analyses indicate that in the case of PPV Rec isolates there may be a recombinant breakpoint distinct from the obvious transition point of strain sequences. Evidence was obtained that indicates that the frequency of PPV recombination is underestimated, which may be true for other RNA viruses where low genetic diversity exists.

  12. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  13. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

  14. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  15. Guidance on accidents involving radioactivity

    International Nuclear Information System (INIS)

    1989-01-01

    This annex contains advice to Health Authorities on their response to accidents involving radioactivity. The guidance is in six parts:-(1) planning the response required to nuclear accidents overseas, (2) planning the response required to UK nuclear accidents a) emergency plans for nuclear installations b) nuclear powered satellites, (3) the handling of casualties contaminated with radioactive substances, (4) background information for dealing with queries from the public in the event of an accident, (5) the national arrangements for incident involving radioactivity (NAIR), (6) administrative arrangements. (author)

  16. The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

    2005-01-01

    The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

  17. Quantification of sequence exchange events between PMS2 and PMS2CL provides a basis for improved mutation scanning of Lynch syndrome patients.

    Science.gov (United States)

    van der Klift, Heleen M; Tops, Carli M J; Bik, Elsa C; Boogaard, Merel W; Borgstein, Anne-Marijke; Hansson, Kerstin B M; Ausems, Margreet G E M; Gomez Garcia, Encarna; Green, Andrew; Hes, Frederik J; Izatt, Louise; van Hest, Liselotte P; Alonso, Angel M; Vriends, Annette H J T; Wagner, Anja; van Zelst-Stams, Wendy A G; Vasen, Hans F A; Morreau, Hans; Devilee, Peter; Wijnen, Juul T

    2010-05-01

    Heterozygous mutations in PMS2 are involved in Lynch syndrome, whereas biallelic mutations are found in Constitutional mismatch repair-deficiency syndrome patients. Mutation detection is complicated by the occurrence of sequence exchange events between the duplicated regions of PMS2 and PMS2CL. We investigated the frequency of such events with a nonspecific polymerase chain reaction (PCR) strategy, co-amplifying both PMS2 and PMS2CL sequences. This allowed us to score ratios between gene and pseudogene-specific nucleotides at 29 PSV sites from exon 11 to the end of the gene. We found sequence transfer at all investigated PSVs from intron 12 to the 3' end of the gene in 4 to 52% of DNA samples. Overall, sequence exchange between PMS2 and PMS2CL was observed in 69% (83/120) of individuals. We demonstrate that mutation scanning with PMS2-specific PCR primers and MLPA probes, designed on PSVs, in the 3' duplicated region is unreliable, and present an RNA-based mutation detection strategy to improve reliability. Using this strategy, we found 19 different putative pathogenic PMS2 mutations. Four of these (21%) are lying in the region with frequent sequence transfer and are missed or called incorrectly as homozygous with several PSV-based mutation detection methods. (c) 2010 Wiley-Liss, Inc.

  18. Overview of AEOD's program for trending reactor operational events

    International Nuclear Information System (INIS)

    Baranowsky, P.W.; O'Reilly, P.D.; Rasmuson, D.M.; Houghton, J.R.

    1994-01-01

    This paper presents an overview of the trending program being performed by AEOD. The major elements of the program include: (1) system and component reliability trending and analysis, (2) special data collection and analysis (e.g., IPE and PRA component failure data, common cause failure event data), (3) risk assessment of safety issues based on actual operating experience, (4) Accident Sequence Precursor (ASP) Program, and (5) trending US industry risk. AEOD plans to maintain up-to-date safety data trends for selected high risk or high regulatory profile components, systems, accident initiators, accident sequences, and regulatory issues. AEOD will also make greater use of PRA insights and perform limited probabilistic safety assessments to evaluate the safety significance of qualitative results. Examples of a system study and an issue evaluation are presented, as well as a summary of the common cause failure event database

  19. Exercise-Induced Rhabdomyolysis and Stress-Induced Malignant Hyperthermia Events, Association with Malignant Hyperthermia Susceptibility, and RYR1 Gene Sequence Variations

    Directory of Open Access Journals (Sweden)

    Antonella Carsana

    2013-01-01

    Full Text Available Exertional rhabdomyolysis (ER and stress-induced malignant hyperthermia (MH events are syndromes that primarily afflict military recruits in basic training and athletes. Events similar to those occurring in ER and in stress-induced MH events are triggered after exposure to anesthetic agents in MH-susceptible (MHS patients. MH is an autosomal dominant hypermetabolic condition that occurs in genetically predisposed subjects during general anesthesia, induced by commonly used volatile anesthetics and/or the neuromuscular blocking agent succinylcholine. Triggering agents cause an altered intracellular calcium regulation. Mutations in RYR1 gene have been found in about 70% of MH families. The RYR1 gene encodes the skeletal muscle calcium release channel of the sarcoplasmic reticulum, commonly known as ryanodine receptor type 1 (RYR1. The present work reviews the documented cases of ER or of stress-induced MH events in which RYR1 sequence variations, associated or possibly associated to MHS status, have been identified.

  20. Seismic sequence stratigraphy of Miocene deposits related to eustatic, tectonic and climatic events, Cap Bon Peninsula, northeastern Tunisia

    Science.gov (United States)

    Gharsalli, Ramzi; Zouaghi, Taher; Soussi, Mohamed; Chebbi, Riadh; Khomsi, Sami; Bédir, Mourad

    2013-09-01

    The Cap Bon Peninsula, belonging to northeastern Tunisia, is located in the Maghrebian Alpine foreland and in the North of the Pelagian block. By its paleoposition, during the Cenozoic, in the edge of the southern Tethyan margin, this peninsula constitutes a geological entity that fossilized the eustatic, tectonic and climatic interactions. Surface and subsurface study carried out in the Cap Bon onshore area and surrounding offshore of Hammamet interests the Miocene deposits from the Langhian-to-Messinian interval time. Related to the basin and the platform positions, sequence and seismic stratigraphy studies have been conducted to identify seven third-order seismic sequences in subsurface (SM1-SM7), six depositional sequences on the Zinnia-1 petroleum well (SDM1-SDM6), and five depositional sequences on the El Oudiane section of the Jebel Abderrahmane (SDM1-SDM5). Each sequence shows a succession of high-frequency systems tract and parasequences. These sequences are separated by remarkable sequence boundaries and maximum flooding surfaces (SB and MFS) that have been correlated to the eustatic cycles and supercycles of the Global Sea Level Chart of Haq et al. (1987). The sequences have been also correlated with Sequence Chronostratigraphic Chart of Hardenbol et al. (1998), related to European basins, allows us to arise some major differences in number and in size. The major discontinuities, which limit the sequences resulted from the interplay between tectonic and climatic phenomena. It thus appears very judicious to bring back these chronological surfaces to eustatic and/or local tectonic activity and global eustatic and climatic controls.

  1. Expert software for accident identification

    International Nuclear Information System (INIS)

    Dobnikar, M.; Nemec, T.; Muehleisen, A.

    2003-01-01

    Each type of an accident in a Nuclear Power Plant (NPP) causes immediately after the start of the accident variations of physical parameters that are typical for that type of the accident thus enabling its identification. Examples of these parameter are: decrease of reactor coolant system pressure, increase of radiation level in the containment, increase of pressure in the containment. An expert software enabling a fast preliminary identification of the type of the accident in Krsko NPP has been developed. As input data selected typical parameters from Emergency Response Data System (ERDS) of the Krsko NPP are used. Based on these parameters the expert software identifies the type of the accident and also provides the user with appropriate references (past analyses and other documentation of such an accident). The expert software is to be used as a support tool by an expert team that forms in case of an emergency at Slovenian Nuclear Safety Administration (SNSA) with the task to determine the cause of the accident, its most probable scenario and the source term. The expert software should provide initial identification of the event, while the final one is still to be made after appropriate assessment of the event by the expert group considering possibility of non-typical events, multiple causes, initial conditions, influences of operators' actions etc. The expert software can be also used as an educational/training tool and even as a simple database of available accident analyses. (author)

  2. The radiological accident in Tammiku

    International Nuclear Information System (INIS)

    1998-01-01

    course of the accident, the remedial actions taken, and the lessons learnt from the sequence of events. It does not include discussions on theoretical aspects of the use and appropriateness of different methods for dose reconstruction

  3. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

    The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

  4. Severe accidents in nuclear reactors

    International Nuclear Information System (INIS)

    Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

    2004-01-01

    The likelihood of accidents leading to core meltdown in nuclear reactors is low. The consequences of such an event are but so severe that developing and implementing of adequate measures for preventing or diminishing the consequences of such events are of paramount importance. The analysis of major accidents requires sophisticated computation codes but necessary are also relevant experiments for checking the accuracy of the predictions and capability of these codes. In this paper an overview of the severe accidents worldwide with definitions, computation codes and relating experiments is presented. The experimental research activity of severe accidents was conducted in INR Pitesti since 2003, when the Institute jointed the SARNET Excellence Network. The INR activity within SARNET consists in studying scenarios of severe accidents by means of ASTEC and RELAP/SCDAP codes and conducting bench-scale experiments

  5. Intervention principles and levels in the event of a nuclear accident. Final report on the Nordic Nuclear Safety Research Project BER-3

    International Nuclear Information System (INIS)

    Walmod-Larsen, O.

    1994-04-01

    The aim of the Nordic BER-3 project has been to harmonize the Nordic intervention levels after a nuclear accident. The paper deals with the findings and recommendations to be presented to the Nordic authorities as background material for common decisions on the most likely protective actions. In the report sheltering, evaluation and relocation are treated in detail. Iodine prophylaxis and foodstuff restrictions are briefly commented on. The basis for this work is the internationally accepted basic principles for interventions

  6. Return on experience by the Marcoule-Gard CLI in the management of a post-accident situation following a nuclear or radiological event

    International Nuclear Information System (INIS)

    Charre, J.P.

    2010-01-01

    After having recalled the legal framework which strengthens the roles and responsibilities of local authorities with respect to natural or technological risks, this short report describes the activity of the Marcoule-Gard CLI (Commission locale d'information, information local commission) in the management of a post-accident situation during a nuclear crisis. This commission notably participates to several work groups, and to some exercises (two examples - a simulated crisis and a real one - are briefly reported and analyzed)

  7. A study on the development of framework and supporting tools for severe accident management

    International Nuclear Information System (INIS)

    Chang, Hyun Sop

    1996-02-01

    Through the extensive research on severe accidents, knowledge on severe accident phenomenology has constantly increased. Based upon such advance, probabilistic risk studies have been performed for some domestic plants to identify plant-specific vulnerabilities to severe accidents. Severe accident management is a program devised to cover such vulnerabilities, and leads to possible resolution of severe accident issues. This study aims at establishing severe accident management framework for domestic nuclear power plants where severe accident management program is not yet established. Emphasis is given to in-vessel and ex-vessel accident management strategies and instrumentation availability for severe accident management. Among the various strategies investigated, primary system depressurization is found to be the most effective means to prevent high pressure core melt scenarios. During low pressure core melt sequences, cooling of in-vessel molten corium through reactor cavity flooding is found to be effective. To prevent containment failure, containment filtered venting is found to be an effective measure to cope with long-term and gradual overpressurization, together with appropriate hydrogen control measure. Investigation of the availability of Yonggwang 3 and 4 instruments shows that most of instruments essential to severe accident management lose their desired functions during the early phase of severe accident progression, primarily due to the environmental condition exceeded ranges of instruments. To prevent instrument failure, a wider range of instruments are recommended to be used for some severe accident management strategies such as reactor cavity flooding. Severe accidents are generally known to accompany a number of complex phenomena and, therefore, it is very beneficial when severe accident management personnel is aided by appropriately designed supporting systems. In this study, a support system for severe accident management personnel is developed

  8. Mutual emergency assistance in the event of accident during transport of radioactive materials within the member states of the European Community

    International Nuclear Information System (INIS)

    Selling, H.A.

    1984-01-01

    The study consist of a compilation of information on the relevant emergency response plans that are at present in existence in the ten countries of the European Community. Consideration is given to the development of proposals for facilitating co-operation between the emergency services in different countries, particularly with regard to accidents that might occur near national boundaries or in countries in which all the necessary resources might not be available. The particular items of interest covered in this study are: compilation of information on existing organizational emergency response arrangements within each Member State relating to accidents in the transport of radioactive materials by all modes, including road, rail, inland waterways, air and compilation of information on existing arrangements for receiving or providing assistance from or to other Member States. Identification of any avoidable incompatibilities on an international scale. Recommendations for improving the existing arrangements and for encouraging the development of adequate systems of mutual emergency notification, liaison and assistance as required by the circumstances, recommendations should be compatible with the broader framework of emergency response for all types of accidents developed within Member States and envisaged in the IAEA system for mutual emergency assistance

  9. Organization of action to be taken in the event of a nuclear accident in the nuclear power stations of Electricite de France

    International Nuclear Information System (INIS)

    Martin, J.J.

    1977-01-01

    Depending on the magnitude of the accident in a nuclear power station, the organization of action provides for calling in only the members of the teams on emergency duty if the accident is localized or for putting into effect a structured emergency plan if the accident goes much beyond the confines of the monitored zone. The emergency plan calls for establishing four command posts, the functions of which are specified. Radiological data are centralized at a monitoring command post whose task is to analyse the situation for the management command post. The latter takes decisions which apply to the interior of the site and provides information to the civil defense services at the prefecture level. Any action to be taken outside the site lies within the competence of the prefecture services, which can draw on the resources of the Ministry of Health and the Commissariat a l'energie atomique. Analysis of some incidents in EDF power stations, which are described briefly, has shown the need to facilitate the work of the responsible officials by the preparation of simple charts or schemes which can be used for making an estimate quickly but on the higher side of the potential dose commitment for the public. (author)

  10. [Good practice in occupational health services--Certification of stroke as an accident at work. Need for secondary prevention in people returning to work after acute cerebrovascular events].

    Science.gov (United States)

    Marcinkiewicz, Andrzej; Walusiak-Skorupa, Jolanta

    2015-01-01

    The classification of an acute vascular episode, both heart infarct and stroke, as an accident at work poses difficulties not only for post accidental teams, but also to occupational health professionals, experts and judges at labor and social insurance courts. This article presents the case of a 41-year-old office worker, whose job involved client services. While attending a very aggressive customer she developed solid stress that resulted in symptoms of the central nervous system (headache, speech disturbances). During her hospitalisation at the neurological unit ischemic stroke with transient mixed type aphasia was diagnosed. Magnetic resonance imaging (MRI) scan of the head revealed subacute ischemia. After an analysis of the accident circumstances, the employer's post accidental team decided that ischemic stroke had been an accident at work, because it was a sudden incident due to an external cause inducing work-related traumatic stroke. As a primary cause tough stress and emotional strain due to the situation developed while attending the customer were acknowledged. During control medical check up after 5 months the patient was found to be fit for work, so she could return to work. However, it should be noted that such a check up examination of subjects returning to work after stroke must be holistic, including the evaluation of job predispositions and health education aimed at secondary prevention of heart and vascular diseases with special reference to their risk factors. This work is available in Open Access model and licensed under a CC BY-NC 3.0 PL license.

  11. Good practice in occupational health services – Certification of stroke as an accident at work. Need for secondary prevention in people returning to work after acute cerebrovascular events

    Directory of Open Access Journals (Sweden)

    Andrzej Marcinkiewicz

    2015-08-01

    Full Text Available The classification of an acute vascular episode, both heart infarct and stroke, as an accident at work poses difficulties not only for post accidental teams, but also to occupational health professionals, experts and judges at labor and social insurance courts. This article presents the case of a 41-year-old office worker, whose job involved client services. While attending a very aggressive customer she developed solid stress that resulted in symptoms of the central nervous system (headache, speech disturbances. During her hospitalisation at the neurological unit ischemic stroke with transient mixed type aphasia was diagnosed. Magnetic resonance imaging (MRI scan of the head revealed subacute ischemia. After an analysis of the accident circumstances, the employer’s post accidental team decided that ischemic stroke had been an accident at work, because it was a sudden incident due to an external cause inducing work-related traumatic stroke. As a primary cause tough stress and emotional strain due to the situation developed while attending the customer were acknowledged. During control medical check up after 5 months the patient was found to be fit for work, so she could return to work. However, it should be noted that such a check up examination of subjects returning to work after stroke must be holistic, including the evaluation of job predispositions and health education aimed at secondary prevention of heart and vascular diseases with special reference to their risk factors. Med Pr 2015;66(4:595–599

  12. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    accidents and, secondly, the physical phenomena, studies and analyses described in Chapters 5 to 8. Chapter 5 is devoted to describing the physical phenomena liable to occur during a core melt accident, in the reactor vessel and the reactor containment. It also presents the sequence of events and the methods for mitigating their impact. For each of the subjects covered, a summary of the physical phenomena involved is followed by a description of the past, present and planned experiments designed to study these phenomena, along with their modelling, the validation of which is based on the test results. The chapter then describes the computer codes that couple all of the models and provide the best current state of knowledge of the phenomena. Lastly, this knowledge is reviewed while taking into account the gaps and uncertainties, and the outlook for the future is presented, notably regarding experimental programmes and the development of modelling and numerical simulation tools. Chapter 6 focuses on the behaviour of the containment enclosures during a core melt accident. After summarising the potential leakage paths of radioactive substances through the different containments in the case of the accidents chosen in the design phase, it presents the studies of the mechanical behaviour of the different containments under the loadings that can result from the hazards linked with the phenomena described in Chapter 5. Chapter 6 also discusses the risks of containment building bypass in a core melt accident situation. Chapter 7 presents the lessons learned regarding the phenomenology of core melt accidents and the improvement of nuclear reactor safety. Lastly, Chapter 8 presents a review of development and validation efforts regarding the main computer codes dealing with 'severe accidents', which draw on and build upon the knowledge mainly acquired through the research programmes: ASTEC (IRSN and GRS), MAAP-4 (FAI (US)) and used by EDF and by utilities in many other

  13. TITAN: a computer program for accident occurrence frequency analyses by component Monte Carlo simulation

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Tamaki, Hitoshi; Kanai, Shigeru

    2000-04-01

    In a plant system consisting of complex equipments and components for a reprocessing facility, there might be grace time between an initiating event and a resultant serious accident, allowing operating personnel to take remedial actions, thus, terminating the ongoing accident sequence. A component Monte Carlo simulation computer program TITAN has been developed to analyze such a complex reliability model including the grace time without any difficulty to obtain an accident occurrence frequency. Firstly, basic methods for the component Monte Carlo simulation is introduced to obtain an accident occurrence frequency, and then, the basic performance such as precision, convergence, and parallelization of calculation, is shown through calculation of a prototype accident sequence model. As an example to illustrate applicability to a real scale plant model, a red oil explosion in a German reprocessing plant model is simulated to show that TITAN can give an accident occurrence frequency with relatively good accuracy. Moreover, results of uncertainty analyses by TITAN are rendered to show another performance, and a proposal is made for introducing of a new input-data format to adapt the component Monte Carlo simulation. The present paper describes the calculational method, performance, applicability to a real scale, and new proposal for the TITAN code. In the Appendixes, a conventional analytical method is shown to avoid complex and laborious calculation to obtain a strict solution of accident occurrence frequency, compared with Monte Carlo method. The user's manual and the list/structure of program are also contained in the Appendixes to facilitate TITAN computer program usage. (author)

  14. Accident at Harrisburg

    International Nuclear Information System (INIS)

    1979-05-01

    The course of events during the accident on 28 March 1979 at Three Mile Island-2 Reactor at Harrisburg, Pennsylvania, is described in detail. The effects (in the environment and within the safety containment) are described. The following points are then discussed: the possibility of a comparable accident occurring in the nuclear power stations in the German Federal Republic; the possibility of any point having been overlooked in the design of nuclear power stations in the Federal Republic; whether previous risk analyses are still valid; and how near the Three Mile Island reactor was to a core meltdown. Some conclusions are drawn. (U.K.)

  15. Event and fault tree model for reliability analysis of the greek research reactor

    International Nuclear Information System (INIS)

    Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes

    2013-01-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

  16. Event and fault tree model for reliability analysis of the greek research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: atalbuquerque@ien.gov.br, E-mail: btony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

  17. Event-specific qualitative and quantitative PCR detection of the GMO carnation (Dianthus caryophyllus) variety Moonlite based upon the 5'-transgene integration sequence.

    Science.gov (United States)

    Li, P; Jia, J W; Jiang, L X; Zhu, H; Bai, L; Wang, J B; Tang, X M; Pan, A H

    2012-04-27

    To ensure the implementation of genetically modified organism (GMO)-labeling regulations, an event-specific detection method was developed based on the junction sequence of an exogenous integrant in the transgenic carnation variety Moonlite. The 5'-transgene integration sequence was isolated by thermal asymmetric interlaced PCR. Based upon the 5'-transgene integration sequence, the event-specific primers and TaqMan probe were designed to amplify the fragments, which spanned the exogenous DNA and carnation genomic DNA. Qualitative and quantitative PCR assays were developed employing the designed primers and probe. The detection limit of the qualitative PCR assay was 0.05% for Moonlite in 100 ng total carnation genomic DNA, corresponding to about 79 copies of the carnation haploid genome; the limit of detection and quantification of the quantitative PCR assay were estimated to be 38 and 190 copies of haploid carnation genomic DNA, respectively. Carnation samples with different contents of genetically modified components were quantified and the bias between the observed and true values of three samples were lower than the acceptance criterion (GMO detection method. These results indicated that these event-specific methods would be useful for the identification and quantification of the GMO carnation Moonlite.

  18. Normal accidents

    International Nuclear Information System (INIS)

    Perrow, C.

    1989-01-01

    The author has chosen numerous concrete examples to illustrate the hazardousness inherent in high-risk technologies. Starting with the TMI reactor accident in 1979, he shows that it is not only the nuclear energy sector that bears the risk of 'normal accidents', but also quite a number of other technologies and industrial sectors, or research fields. The author refers to the petrochemical industry, shipping, air traffic, large dams, mining activities, and genetic engineering, showing that due to the complexity of the systems and their manifold, rapidly interacting processes, accidents happen that cannot be thoroughly calculated, and hence are unavoidable. (orig./HP) [de

  19. Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

    International Nuclear Information System (INIS)

    Park, Soo Young; Ahn, Kwang Il

    2012-01-01

    Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

  20. Genomic Investigation Reveals Highly Conserved, Mosaic, Recombination Events Associated with Capsular Switching among Invasive Neisseria meningitidis Serogroup W Sequence Type (ST)-11 Strains.

    Science.gov (United States)

    Mustapha, Mustapha M; Marsh, Jane W; Krauland, Mary G; Fernandez, Jorge O; de Lemos, Ana Paula S; Dunning Hotopp, Julie C; Wang, Xin; Mayer, Leonard W; Lawrence, Jeffrey G; Hiller, N Luisa; Harrison, Lee H

    2016-07-03

    Neisseria meningitidis is an important cause of meningococcal disease globally. Sequence type (ST)-11 clonal complex (cc11) is a hypervirulent meningococcal lineage historically associated with serogroup C capsule and is believed to have acquired the W capsule through a C to W capsular switching event. We studied the sequence of capsule gene cluster (cps) and adjoining genomic regions of 524 invasive W cc11 strains isolated globally. We identified recombination breakpoints corresponding to two distinct recombination events within W cc11: A 8.4-kb recombinant region likely acquired from W cc22 including the sialic acid/glycosyl-transferase gene, csw resulted in a C→W change in capsular phenotype and a 13.7-kb recombinant segment likely acquired from Y cc23 lineage includes 4.5 kb of cps genes and 8.2 kb downstream of the cps cluster resulting in allelic changes in capsule translocation genes. A vast majority of W cc11 strains (497/524, 94.8%) retain both recombination events as evidenced by sharing identical or very closely related capsular allelic profiles. These data suggest that the W cc11 capsular switch involved two separate recombination events and that current global W cc11 meningococcal disease is caused by strains bearing this mosaic capsular switch. © The Author 2016. Published by Oxford University Press on behalf of the Society for Molecular Biology and Evolution.

  1. Biological effects of ionizing radiations. Radiological accident from Goiania, GO, Brazil; Efeitos biologicos das radiacoes ionizantes. Acidente radiologico de Goiania

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Emico, E-mail: emico.okuno@if.usp.br [Instituto de Fisica da Universidade de Sao Paulo (IF-USP), SP (Brazil)

    2013-01-15

    This article presents the fundaments of radiation physics, the natural and artificial sources, biological effects, radiation protection. We also examine the sequence of events that resulted in Goiania accident with a source of caesium-137 from abandoned radiotherapy equipment and its terrible consequences. (author)

  2. BiodosEPR-2006 Meeting: Acute dosimetry consensus committee recommendations on biodosimetry applications in events involving uses of radiation by terrorists and radiation accidents

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, George A. [U.S. Department of Health and Human Services, Office of Preparedness and Emergency Operations, 200 Independence Avenue, SW, Room 403B-1, Washington, DC 20201 (United States); Swartz, Harold M. [Dept. of Radiology and Physiology Dept., Dartmouth Medical School, HB 7785, Vail 702, Rubin 601, Hanover, NH 03755 (United States); Amundson, Sally A. [Center for Radiological Research, Columbia University Medical Center, 630 W. 168th Street, VC11-215, New York, NY 10032 (United States); Blakely, William F. [Armed Forces Radiobiology Research Inst., 8901 Wisconsin Avenue, Bethesda, MD 20889-5603 (United States)], E-mail: blakely@afrri.usuhs.mil; Buddemeier, Brooke [Science and Technology, U.S. Department of Homeland Security, Washington, DC 20528 (United States); Gallez, Bernard [Biomedical Magnetic Resonance Unit and Lab. of Medicinal Chemistry and Radiopharmacy, Univ. Catholique de Louvain, Brussels (Belgium); Dainiak, Nicholas [Dept. of Medicine, Bridgeport Hospital, 267 Grant Street, Bridgeport, CT 06610 (United States); Goans, Ronald E. [MJW Corporation, 1422 Eagle Bend Drive, Clinton, TN 37716-4029 (United States); Hayes, Robert B. [Remote Sensing Lab., MS RSL-47, P.O. Box 98421, Las Vegas, NV 89193 (United States); Lowry, Patrick C. [Radiation Emergency Assistance Center/Training Site (REAC/TS), Oak Ridge Associated Universities, P.O. Box 117, Oak Ridge, TN 37831-0117 (United States); Noska, Michael A. [Food and Drug Administration, FDA/CDRH, 1350 Piccard Drive, HFZ-240, Rockville, MD 20850 (United States); Okunieff, Paul [Dept. of Radiation Oncology (Box 647), Univ. of Rochester, 601 Elmwood Avenue, Rochester, NY 14642 (United States); Salner, Andrew L. [Helen and Harry Gray Cancer Center, Hartford Hospital, 80 Seymour Street, Hartford, CT 06102 (United States); Schauer, David A. [National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 400, Bethesda, MD 20814-3095 (United States)] (and others)

    2007-07-15

    In the aftermath of a radiological terrorism incident or mass-casualty radiation accident, first responders and receivers require prior guidance and pre-positioned resources for assessment, triage and medical management of affected individuals [NCRP, 2005. Key elements of preparing emergency responders for nuclear and radiological terrorism. NCRP Commentary No. 19, Bethesda, Maryland, USA]. Several recent articles [Dainiak, N., Waselenko, J.K., Armitage, J.O., MacVittie, T.J., Farese, A.M., 2003. The hematologist and radiation casualties. Hematology (Am. Soc. Hematol. Educ. Program) 473-496; Waselenko, J.K., MacVittie, T.J., Blakely, W.F., Pesik, N., Wiley, A.L., Dickerson, W.E., Tsu, H., Confer, D.L., Coleman, C.N., Seed, T., Lowry, P., Armitage, J.O., Dainiak, N., Strategic National Stockpile Radiation Working Group, 2004. Medical management of the acute radiation syndrome: recommendations of the Strategic National Stockpile Radiation Working Group. Ann. Intern. Med. 140(12), 1037-1051; Blakely, W.F., Salter, C.A., Prasanna, P.G., 2005. Early-response biological dosimetry-recommended countermeasure enhancements for mass-casualty radiological incidents and terrorism. Health Phys. 89(5), 494-504; Goans, R.E., Waselenko, J.K., 2005. Medical management of radiation casualties. Health Phys. 89(5), 505-512; Swartz, H.M., Iwasaki, A., Walczak, T., Demidenko, E., Salikhov, I., Lesniewski, P., Starewicz, P., Schauer, D., Romanyukha, A., 2005. Measurements of clinically significant doses of ionizing radiation using non-invasive in vivo EPR spectroscopy of teeth in situ. Appl. Radiat. Isot. 62, 293-299; . Acute radiation injury: contingency planning for triage, supportive care, and transplantation. Biol. Blood Marrow Transplant. 12(6), 672-682], national [. Management of persons accidentally contaminated with radionuclides. NCRP Report No. 65, Bethesda, Maryland, USA; . Management of terrorist events involving radioactive material. NCRP Report No. 138, Bethesda, Maryland

  3. BiodosEPR-2006 Meeting: Acute dosimetry consensus committee recommendations on biodosimetry applications in events involving uses of radiation by terrorists and radiation accidents

    International Nuclear Information System (INIS)

    Alexander, George A.; Swartz, Harold M.; Amundson, Sally A.; Blakely, William F.; Buddemeier, Brooke; Gallez, Bernard; Dainiak, Nicholas; Goans, Ronald E.; Hayes, Robert B.; Lowry, Patrick C.; Noska, Michael A.; Okunieff, Paul; Salner, Andrew L.; Schauer, David A.

    2007-01-01

    In the aftermath of a radiological terrorism incident or mass-casualty radiation accident, first responders and receivers require prior guidance and pre-positioned resources for assessment, triage and medical management of affected individuals [NCRP, 2005. Key elements of preparing emergency responders for nuclear and radiological terrorism. NCRP Commentary No. 19, Bethesda, Maryland, USA]. Several recent articles [Dainiak, N., Waselenko, J.K., Armitage, J.O., MacVittie, T.J., Farese, A.M., 2003. The hematologist and radiation casualties. Hematology (Am. Soc. Hematol. Educ. Program) 473-496; Waselenko, J.K., MacVittie, T.J., Blakely, W.F., Pesik, N., Wiley, A.L., Dickerson, W.E., Tsu, H., Confer, D.L., Coleman, C.N., Seed, T., Lowry, P., Armitage, J.O., Dainiak, N., Strategic National Stockpile Radiation Working Group, 2004. Medical management of the acute radiation syndrome: recommendations of the Strategic National Stockpile Radiation Working Group. Ann. Intern. Med. 140(12), 1037-1051; Blakely, W.F., Salter, C.A., Prasanna, P.G., 2005. Early-response biological dosimetry-recommended countermeasure enhancements for mass-casualty radiological incidents and terrorism. Health Phys. 89(5), 494-504; Goans, R.E., Waselenko, J.K., 2005. Medical management of radiation casualties. Health Phys. 89(5), 505-512; Swartz, H.M., Iwasaki, A., Walczak, T., Demidenko, E., Salikhov, I., Lesniewski, P., Starewicz, P., Schauer, D., Romanyukha, A., 2005. Measurements of clinically significant doses of ionizing radiation using non-invasive in vivo EPR spectroscopy of teeth in situ. Appl. Radiat. Isot. 62, 293-299; . Acute radiation injury: contingency planning for triage, supportive care, and transplantation. Biol. Blood Marrow Transplant. 12(6), 672-682], national [. Management of persons accidentally contaminated with radionuclides. NCRP Report No. 65, Bethesda, Maryland, USA; . Management of terrorist events involving radioactive material. NCRP Report No. 138, Bethesda, Maryland

  4. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  5. Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents

    International Nuclear Information System (INIS)

    Houghton, W.J.

    1980-06-01

    A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors

  6. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  7. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  8. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  9. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1998-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  10. Use of NUREG-1150 and IPEs in accident management

    International Nuclear Information System (INIS)

    Mauersberger

    1992-01-01

    The fundamental objective of the accident management program is to assure, in the event of a severe accident at a nuclear plant, that the effectiveness of personnel and equipment is maximized in preventing or mitigating the consequences of the accident. This document studies the use of NUREG-1150 and IPEs in accident management. Figs

  11. 10 CFR 76.85 - Assessment of accidents.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy... Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences to... postulated accidents which include internal and external events and natural phenomena in order to ensure...

  12. 46 CFR 97.30-5 - Accidents to machinery.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Accidents to machinery. 97.30-5 Section 97.30-5 Shipping... Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use of...

  13. 46 CFR 78.33-5 - Accidents to machinery.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Accidents to machinery. 78.33-5 Section 78.33-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PASSENGER VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-5 Accidents to machinery. (a) In the event of an accident...

  14. 46 CFR 196.30-5 - Accidents to machinery.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Accidents to machinery. 196.30-5 Section 196.30-5... Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-5 Accidents to machinery. (a) In the event of an accident to a boiler, unfired pressure vessel, or machinery tending to render the further use of...

  15. 50 CFR 25.72 - Reporting of accidents.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 6 2010-10-01 2010-10-01 false Reporting of accidents. 25.72 Section 25... Reporting of accidents. Accidents involving damage to property, injury to the public or injury to wildlife..., but in no event later than 24 hours after the accident, by the persons involved, to the refuge manager...

  16. Flanking sequence determination and event-specific detection of genetically modified wheat B73-6-1.

    Science.gov (United States)

    Xu, Junyi; Cao, Jijuan; Cao, Dongmei; Zhao, Tongtong; Huang, Xin; Zhang, Piqiao; Luan, Fengxia

    2013-05-01

    In order to establish a specific identification method for genetically modified (GM) wheat, exogenous insert DNA and flanking sequence between exogenous fragment and recombinant chromosome of GM wheat B73-6-1 were successfully acquired by means of conventional polymerase chain reaction (PCR) and thermal asymmetric interlaced (TAIL)-PCR strategies. Newly acquired exogenous fragment covered the full-length sequence of transformed genes such as transformed plasmid and corresponding functional genes including marker uidA, herbicide-resistant bar, ubiquitin promoter, and high-molecular-weight gluten subunit. The flanking sequence between insert DNA revealed high similarity with Triticum turgidum A gene (GenBank: AY494981.1). A specific PCR detection method for GM wheat B73-6-1 was established on the basis of primers designed according to the flanking sequence. This specific PCR method was validated by GM wheat, GM corn, GM soybean, GM rice, and non-GM wheat. The specifically amplified target band was observed only in GM wheat B73-6-1. This method is of high specificity, high reproducibility, rapid identification, and excellent accuracy for the identification of GM wheat B73-6-1.

  17. Application of probabilistic methods to accident analysis at waste management facilities

    International Nuclear Information System (INIS)

    Banz, I.

    1986-01-01

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

  18. Study on severe accidents and countermeasures for WWER-1000 reactors using the integral code ASTEC

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Altstadt, E.; Kliem, S.; Reinke, N.

    2011-01-01

    The research field focussing on the investigations and the analyses of severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. On the one side, it is the aim of the severe accident research to understand the complex phenomena during the in- and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective severe accident management measures. This paper is focused on the possibilities for accident management measures in case of severe accidents. The reactor pressure vessel is the last barrier to keep the molten materials inside the reactor, and thus to prevent higher loads to the containment. To assess the behaviour of a nuclear power plant during transient or accident conditions, computer codes are widely used, which have to be validated against experiments or benchmarked against other codes. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The results have shown that in case of unavailability of major active safety systems the reactor pressure vessel would ultimately fail. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accident, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed. The presented investigations contribute to the validation of the European integral severe accidents code ASTEC for WWER-1000 type of reactors

  19. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

  20. In the event of a nuclear accident in France: the IRSN makes its expertise available for the management of the post-accidental consequences

    International Nuclear Information System (INIS)

    Cessac, B.; Herviou, K.

    2013-01-01

    The lack of organisation, methodologies and strategies to respond to the post-accidental consequences of a nuclear accident in France has been point out in 2004. In 2005, a Post-accident Management Steering Committee (CODIR-PA) has been established by the French Nuclear Safety Authority (ASN) to define the first elements of an updated French doctrine on the management of situations following a nuclear accident or a radiological emergency. The process involves several partners among them IRSN which provides a technical and scientific support. Developments made by the IRSN for more than twenty years in the field of sciences of the environment, of the man health and of metrology have indeed brought a strong basis to support current reflexions. In this context, IRSN contributes to make some proposals in order to elaborate a zoning for the implementation of protective action in post-accidental situations. This zoning is focused on actions regarding locally produced foodstuffs which may be contaminated as well as on the issue of relocation of population when the ambient level of radioactivity does not allow anymore people to stay in the area. Moreover, in case of emergency situation resulting in contaminated areas, the IRSN makes as well its expertise available to public authorities by suggesting optimized strategies to manage the effects in the environment and on the population. In the prolongation of its action during the emergency phase, the IRSN must continue to mobilize its human capacities and its technical means to answer the specific needs which are posed at the time of the post-accidental phase. The first elements of doctrines of management of post-accidental situations emitted by the CODIR-PA bring a new visibility to the IRSN in its future developments, in particular on the practical and operational aspects of the expertise during an emergency. The incident that occurred in France in July 2008 has shown the central position of the IRSN as a support to the

  1. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

    2014-01-01

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  2. Dose assessment in radiological accidents

    International Nuclear Information System (INIS)

    Donkor, S.

    2013-04-01

    The applications of ionizing radiation bring many benefits to humankind, ranging from power generation to uses in medicine, industry and agriculture. Facilities that use radiation source require special care in the design and operation of equipment to prevent radiation injury to workers or to the public. Despite considerable development of radiation safety, radiation accidents do happen. The purpose of this study is therefore to discuss how to assess doses to people who will be exposed to a range of internal and external radiation sources in the event of radiological accidents. This will go a long way to complement their medical assessment thereby helping to plan their treatment. Three radiological accidents were reviewed to learn about the causes of those accidents and the recommendations that were put in place to prevent recurrence of such accidents. Various types of dose assessment methods were discussed.(au)

  3. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  4. Nuclear ship accidents

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1993-05-01

    In this report available information on 28 nuclear ship accident and incidents is considered. Of these 5 deals with U.S. ships and 23 with USSR ships. The ships are in almost all cases nuclear submarines. Only events that involve the nuclear propulsion plants, radiation exposures, fires/explosions and sea water leaks into the submarines are considered. Comments are made on each of the events, and at the end of the report an attempt is made to point out the weaknesses of the submarine designs which have resulted in the accidents. It is emphasized that much of the available information is of a rather dubious nature. consequently some of the assessments made may not be correct. (au)

  5. The effectiveness of using pictures in teaching young children about burn injury accidents.

    Science.gov (United States)

    Liu, Hsueh-Fen; Lin, Fang-Suey; Chang, Chien-Ju

    2015-11-01

    This study utilized the "story grammar" approach (Stein and Glenn, 1979) to analyze the within-corpus differences in recounting of sixty 6- and 7-year-old children, specifically whether illustrations (5-factor accident sequence) were or were not resorted to as a means to assist their narration of a home accident in which a child received a burn injury from hot soup. Our investigation revealed that the message presentation strategy "combining oral and pictures" better helped young children to memorize the story content (sequence of events leading to the burn injury) than "oral only." Specifically, the content of "the dangerous objects that caused the injury", "the unsafe actions that people involved took", and "how the people involved felt about the severity of the accident" differed significantly between the two groups. Copyright © 2015 Elsevier Ltd and The Ergonomics Society. All rights reserved.

  6. Derived intervention levels for application in controlling radiation doses to the public in the event of a nuclear accident or radiological emergency

    International Nuclear Information System (INIS)

    1986-01-01

    This document sets out the principles and procedures for estimating derived intervention levels (DILs) and illustrates the application of these procedures to the estimation of DILs for a range of nuclides and exposure pathways, as aids to decisions on the application of protective measures for the public in the early and intermediate phases of a nuclear accident. The levels are derived subject to a number of assumptions about the intervention level of dose, the characteristics of the released material, the habits of the exposed individuals and local environmental conditions. Some guidance is given on the sensitivity of the estimated DILs to plausible variations in the above assumptions. The detailed procedures described in the document for estimating DILs and illustrations of their application are limited to accidental releases to the atmosphere

  7. Severe accident management guidelines

    International Nuclear Information System (INIS)

    Uhle, Jennifer

    2014-01-01

    The events at Fukushima Daiichi have highlighted the importance of Severe Accident Management Guidelines (SAMGs). As the world has learned from the catastrophe and countries are considering changes to their nuclear regulatory programs, the content of SAMGs and their regulatory control are being evaluated. This presentation highlights several factors that are being addressed in the United States as rulemaking is underway pertaining to SAMGs. The question of how to be prepared for the unexpected is discussed with specific insights gleaned from Fukushima. (author)

  8. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Rassow, J.

    1986-01-01

    The documentation aims at giving a clearly arranged account of facts, interrelations and comparative evaluations of general interest. It deals with the course of events, atmospheric dispersion and fallout of the substances released and discusses the basic principles of the metering of radioactive radiation, the calculation of body doses and comparative evaluations with the radioactive exposure and risks involved by other sources. The author intends to contribute to an objective discussion about the Chernobyl reactor accident and nuclear energy as such. (DG) [de

  9. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 5: Analysis of core damage frequency from seismic events for plant operational state 5 during a refueling outage

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

    1994-08-01

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Sandia National Laboratories studying a boiling water reactor (Grand Gulf), and the other at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1). Both the Sandia and Brookhaven projects have examined only accidents initiated by internal plant faults---so-called ''internal initiators.'' This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling outage conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Grand Gulf. All of the many systems modeling assumptions, component non-seismic failure rates, and human effort rates that were used in the internal-initiator study at Grand Gulf have been adopted here, so that the results of the study can be as comparable as possible. Both the Sandia study and this study examine only one shutdown plant operating state (POS) at Grand Gulf, namely POS 5 representing cold shutdown during a refueling outage. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POS 5. The results of the analysis are that the core-damage frequency for earthquake-initiated accidents during refueling outages in POS 5 is found to be quite low in absolute terms, less than 10 -7 /year

  10. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1. Volume 5: Analysis of core damage frequency from seismic events during mid-loop operations

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

    1994-08-01

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1) and the other at Sandia National Laboratories studying a boiling water reactor (Grand Gulf). Both the Brookhaven and Sandia projects have examined only accidents initiated by internal plant faults--so-called ''internal initiators.'' This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Surry Unit 1. All of the many systems modeling assumptions, component non-seismic failure rates, and human error rates that were used in the internal-initiator study at Surry have been adopted here, so that the results of the two studies can be as comparable as possible. Both the Brookhaven study and this study examine only two shutdown plant operating states (POSs) during refueling outages at Surry, called POS 6 and POS 10, which represent mid-loop operation before and after refueling, respectively. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POSs 6 and 10. The results of the analysis are that the core-damage frequency of earthquake-initiated accidents during refueling outages in POS 6 and POS 10 is found to be low in absolute terms, less than 10 -6 /year

  11. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station

    International Nuclear Information System (INIS)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

    1993-01-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences</