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Sample records for accident sequence event

  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. 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.

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. 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.

  9. 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

  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. 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. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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)

  9. 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

  10. 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

  11. 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)

  12. 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.

  13. 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...

  14. 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)

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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. ...

  14. 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

  15. 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

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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.

  6. 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

  7. 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)

  8. 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

  9. 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.)

  10. 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)

  11. 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)

  12. 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

  13. LOSP-initiated event tree analysis for BWR

    International Nuclear Information System (INIS)

    Watanabe, Norio; Kondo, Masaaki; Uno, Kiyotaka; Chigusa, Takeshi; Harami, Taikan

    1989-03-01

    As a preliminary study of 'Japanese Model Plant PSA', a LOSP (loss of off-site power)-initiated Event Tree Analysis for a Japanese typical BWR was carried out solely based on the open documents such as 'Safety Analysis Report'. The objectives of this analysis are as follows; - to delineate core-melt accident sequences initiated by LOSP, - to evaluate the importance of core-melt accident sequences in terms of occurrence frequency, and - to develop a foundation of plant information and analytical procedures for efficiently performing further 'Japanese Model Plant PSA'. This report describes the procedure and results of the LOSP-initiated Event Tree Analysis. In this analysis, two types of event trees, Functional Event Tree and Systemic Event Tree, were developed to delineate core-melt accident sequences and to quantify their frequencies. Front-line System Event Tree was prepared as well to provide core-melt sequence delineation for accident progression analysis of Level 2 PSA which will be followed in a future. Applying U.S. operational experience data such as component failure rates and a LOSP frequency, we obtained the following results; - The total frequency of core-melt accident sequences initiated by LOSP is estimated at 5 x 10 -4 per reactor-year. - The dominant sequences are 'Loss of Decay Heat Removal' and 'Loss of Emergency Electric Power Supply', which account for more than 90% of the total core-melt frequency. In this analysis, a higher value of 0.13/R·Y was used for the LOSP frequency than experiences in Japan and any recovery action was not considered. In fact, however, there has been no experience of LOSP event in Japanese nuclear power plants so far and it is also expected that offsite power and/or PCS would be recovered before core melt. Considering Japanese operating experience and recovery factors will reduce the total core-melt frequency to less than 10 -6 per reactor-year. (J.P.N.)

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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

  19. 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)

  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. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. 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.

  11. 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

  12. 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

  13. 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...

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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)

  19. 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

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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)

  5. 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)

  6. 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.).

  7. 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

  8. 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)

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. Reference accident (Core disruption accident - safety analysis detailed report no. 11)

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    The PEC safety analysis led to the conclusion that all credible sequences (incident sequences characterized by a frequency of occurrence above 10/sup minus 7/ events per year) are limited to the design basis conditions of components of the plant protection systems, and that none of them leads to a release of mechanical energy or to an extensive damage of the core and primary containment structures event in the case of failure to scram. Nevertheless, as is done in other countries for similar reactors, some events beyond the limits of credibility were considered for the PEC reactor. These were defined on a absolutely hypothetical basis that involves severe core disruption and dynamic loading of primary containment boundary. A series of containments, each having a different role, was designed to mitigate the radiological effects of a postulated core disruptive accident. The final aim was to demonstrate that residual heat can be removed and that the release of radioactivity to the environment is within acceptable limits.

  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. 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

  19. 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

  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. 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

  2. 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

  3. 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

  4. Analysis of Three Mile Island - Unit 2 accident

    International Nuclear Information System (INIS)

    1979-07-01

    The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute is analyzing the Three Mile Island-2 accident. An early result of this analysis was a brief narrative summary, issued in mid May 1979. The present report contains a revised version of that narrative summary, a highly detailed sequence of events, a standard reference list, a list of abbreviations and acronyms, and several appendices. The appendices serve either to describe plant features which are pertinent to the understanding of the sequence of events, or indicate how certain inferences and conclusions in the report were reached. Supplementing the appendices contained herein, additional appendices are in preparation; these will be issued when available (e.g., the appendices Hydrogen Phenomena and Operator Actions during Initial Transient will follow later). Also in preparation is a matrix of equipment and systems actions during the accident. This report together with future supplements and a separate Core Damage Assessment report, will embody the principal results of that phase of NSAC work which is devoted to learning and understanding what happened during the accident. Subsequent phases will concentrate on causes, lessons learned and generic remedial or preventive measures which may be appropriate

  5. Analysis of Three Mile Island-Unit 2 accident

    International Nuclear Information System (INIS)

    1979-07-01

    The Nuclear Safety Analysis Center (NSAC) of the Electic Power Research Institute is analyzing the Three Mile Island-2 accident. An early result of this analysis was a brief narrative summary, issued in mid-May 1979. The present report contains a revised version of that narrative summary, a highly detailed sequence of events, a standard reference list, a list of abbreviations and acronyms, and several appendices. The appendices serve either to describe plant features which are pertinent to the understanding of the sequence of events, or indicate how certain inferences and conclusions in the report were reached. Supplementing the appendices contained herein, additional appendices are in preparation; these will be issued when available (e.g., the appendices Hydrogen Phenomena and Operator Actions duing Initial Transient will follow later). Also in preparation is a matrix of equipment and systems actions during the accident. This report together with future supplements and a separate Core Damage Assessment report, will embody the principal results of that phase of NSAC's work which is devoted to learning and understanding what happened during the accident. Subsequent phases will concentrate on causes, lessons learned and generic remedial or preventive measures which may be appropriate

  6. 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

  7. 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)

  8. 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

  9. 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).

  10. 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.)

  11. 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

  12. 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

  13. 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)

  14. 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...

  15. '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.

  16. Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1990-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. Large LOCA accident analysis for AP1000 under earthquake

    International Nuclear Information System (INIS)

    Yu, Yu; Lv, Xuefeng; Niu, Fenglei

    2015-01-01

    Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

  4. 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

  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. 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

  7. 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

  8. Introduction of operator actions in the event trees

    International Nuclear Information System (INIS)

    Bars, G.; Lanore, J.M.; Villeroux, C.

    1984-11-01

    In the PRA in progress in France for a 900 MW PWR plant, an effort is done for introducing operator actions during accident sequences. A first approach of this complex problem relies on an extensive use of existing methods an knowledge in diverse fields. Identification of actions is based on the operating procedures, and in particular on the existence of special emergency procedures which define the optimal actions during severe accidents. This approach implies the introduction in the event trees of the notion of procedure failure. Quantification of the corresponding probabilities leads to several problems including physics of the sequences, systems availability and human behaviour for decision making and actions. This treatment is illustrated by the example of the small break event tree

  9. 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)

  10. 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)

  11. 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.

  12. 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.

  13. 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

  14. 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 for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  15. 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.

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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

  2. Overview of results and perspectives from the Shoreham major common-cause initiating events study

    International Nuclear Information System (INIS)

    Joksimovich, V.; Orvis, D.D.; Paccione, R.J.

    1986-01-01

    This study represents the continuation of a large effort by LILCO to fully understand the potential hazards posed by future operation of the Shoreham Nuclear Power Stations (SNPS). The Shoreham Probabilistic Risk Assessment, a level 3 PRA without external events, provided a characterization of the accident sequences that could leave the core in a condition in which it would be vulnerable to severe damage if further mitigating actions were not taken. It estimated the frequency and magnitude of the potential radioactivity releases associated with such sequences. The study was limited to accident sequences initiated by so called internal events to the plant including a loss of offsite power. It also characterized the public risk associated with those accident sequences. The ''Major Common-Cause Initiating Events Study'' (MCCI) for the Shoreham plant was performed to obtain insights into the plant's susceptibility to, and inherent defenses against, certain MCCIs. Major common-cause initiating events are occurrences which have the potential to initiate a plant transient or LOCA and, also, damage one or more plant systems needed to mitigate the effects of a transient or LOCA. The scope of the MCCI study included detailed analyses of seismic events and fires through the severe core damage and bounding analyses of aircraft crashes, windstorms, turbine missiles and release of hazardous materials near the plant

  3. 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

  4. 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...

  5. 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

  6. 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

  7. 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.

  8. A study on the implementation effect of accident management strategies on safety

    International Nuclear Information System (INIS)

    Jae, Moo Sung; Kim, Dong Ha; Jin, Young Ho

    1996-01-01

    This paper presents a new approach for assessing accident management strategies using containment event trees(CETs) developed during an individual plant examination (IPE) for a reference plant (CE type, 950 MWe PWR). Various accident management strategies to reduce risk have been proposed through IPE. Three strategies for the station blackout sequence are used as an example: 1) reactor cavity flooding only, 2) primary system depressurization only, and 3) doing both. These strategies are assumed to be initiated at about the time of core uncovery. The station blackout (SBO) sequence is selected in this paper since it is identified as one of the most threatening sequences to safety of the reference plant. The effectiveness and adverse effects of each accident management strategy are considered synthetically in the CETs. A best estimate assessment for the developed CETs using data obtained from NUREG-1150, other PRA results, and the MAAP code calculations is performed. The strategies are ranked with respect to minimizing the frequencies of various containment failure modes. The proposed approach is demonstrated to be very flexible in that it can be applied to any kind of accident management strategy for any sequence. 9 refs., 3 figs., 2 tabs. (author)

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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

  15. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.

    1985-01-01

    SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

  16. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

    1986-01-01

    The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

  17. 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

  18. 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

  19. 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.

  20. 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)

  1. 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).

  2. 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)

  3. Operator modeling of a loss-of-pumping accident using MicroSAINT

    International Nuclear Information System (INIS)

    Olsen, L.M.

    1992-01-01

    The Savannah River Laboratory (SRL) human factors group has been developing methods for analyzing nuclear reactor operator actions during hypothetical design-basis accident scenarios. The SRL reactors operate at a lower temperature and pressure than power reactors resulting in accident sequences that differ from those of power reactors. Current methodology development is focused on modeling control room operator response times dictated by system event times specified in the Savannah River Site Reactor Safety Analysis Report (SAR). The modeling methods must be flexible enough to incorporate changes to hardware, procedures, or postulated system event times and permit timely evaluation. The initial model developed was for the loss-of-pumping accident (LOPA) because a significant number of operator actions are required to respond to this postulated event. Human factors engineers had been researching and testing a network modeling simulation language called MicroSAINT to simulate operators' personal and interpersonal actions relative to operating system events. The LOPA operator modeling project demonstrated the versatility and flexibility of MicroSAINT for modeling control room crew interactions

  4. 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

  5. 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.

  6. Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

  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. 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.

  9. 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)

  10. 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)

  11. 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

  12. 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...

  13. 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.

  14. 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

  15. 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

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. Identification of Initiating Events for PGSFR

    International Nuclear Information System (INIS)

    Kim, Jintae; Jae, Moosung

    2016-01-01

    The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150

  3. Identification of Initiating Events for PGSFR

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

    The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150.

  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. 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

  6. 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)

  7. 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)

  8. 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)

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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,…

  14. 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.

  15. 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

  16. 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.

  17. Procedure for conducting probabilistic safety assessment: level 1 full power internal event analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Won Dae; Lee, Y. H.; Hwang, M. J. [and others

    2003-07-01

    This report provides guidance on conducting a Level I PSA for internal events in NPPs, which is based on the method and procedure that was used in the PSA for the design of Korea Standard Nuclear Plants (KSNPs). Level I PSA is to delineate the accident sequences leading to core damage and to estimate their frequencies. It has been directly used for assessing and modifying the system safety and reliability as a key and base part of PSA. Also, Level I PSA provides insights into design weakness and into ways of preventing core damage, which in most cases is the precursor to accidents leading to major accidents. So Level I PSA has been used as the essential technical bases for risk-informed application in NPPs. The report consists six major procedural steps for Level I PSA; familiarization of plant, initiating event analysis, event tree analysis, system fault tree analysis, reliability data analysis, and accident sequence quantification. The report is intended to assist technical persons performing Level I PSA for NPPs. A particular aim is to promote a standardized framework, terminology and form of documentation for PSAs. On the other hand, this report would be useful for the managers or regulatory persons related to risk-informed regulation, and also for conducting PSA for other industries.

  18. 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

  19. 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

  20. 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)

  1. 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.

  2. 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.

  3. SCENARIO OF AN ACCIDENT OF SOIL DAMS IN CASE OF WATER SPILL OVER A DAM CREST BY USING FAULT TREE ANALYSIS

    Directory of Open Access Journals (Sweden)

    Kuznetsov Dmitriy Viktorovich

    2016-04-01

    Full Text Available 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 crest, an excess flood, an inoperable spillway, taking into account the sequence of the events’ development and with observance of the necessary conditions leading to an accident. “Technical” reasons were observed in the present scenario, force majeure events were not considered. The provided scenario of the accident consists of two branches of events’ development: the left one that depends on an upstream level, and the right one that depends on settling of a dam crest. In each of the considered events an accident “the water spill over a crest of a soil dam” is possible only in case of execution of two different conditions at the same time, i.e. in case of an appropriate upstream level and the appropriate mark of a crest of a soil dam. The conditions of the accident are defined by diagnostic indices - the upstream level and settling of a dam crest, which at the same time are safety criteria of the hydraulic structure for soil dams. They allow defining the technical condition of the construction. Four possible technical conditions are suggested for the definition of technical statuses - normative, operable, limited operable, abnormal. Criteria of safety are the boundaries of the state: for loading and impact - it is the upstream level, for geometrical compliance of the construction - it is a dam crest mark.

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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)

  11. Report on Fukushima Daiichi NPP precursor events

    International Nuclear Information System (INIS)

    2014-01-01

    also been analysed regarding their initiators, their effective barriers and their main lessons learned. The question related to the effective barriers in significant events can be answered with respect to the different event sequences and the specific NPP designs. Regarding the question related to the potential improvements for the international systems on operating experiences, it can be stated that significant efforts have been made - nationally and internationally - to derive specific and generic recommendations for further improvement of NPPs. The conclusion of this report is that the existing operating experience feedback systems provide a good tool to prevent recurrence of events. Operating experiences considering also the risk significance provide a great source of potential improvements that have demonstrated their usefulness in the course of real events. There have been major works done on new event features, e.g. after the Barsebaeck event and the Forsmark- 1 event. After the TMI-2 accident and the Chernobyl accident - as well as after the Fukushima Daiichi NPP accident - further international approaches have been started far beyond the continuous work of WGOE

  12. 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)

  13. 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.

  14. Implementation of accident management programmes in nuclear power plants

    International Nuclear Information System (INIS)

    2004-01-01

    good practices and developments in Member States and is intended as reference material for NPPs, as well as an information source for other organizations such as regulatory bodies. It is a follow-up to the IAEA report on Accident Management Programmes in Nuclear Power Plants, published in 1994, and reflects the considerable progress made since that time. The objective of this report is to provide a description of the elements to be addressed by the team responsible for developing and implementing a plant specific AMP at an NPP. Although it is intended primarily for use by NPP operators, utilities and their technical support organizations, it can also facilitate preparation of the relevant national regulatory requirements. Important event sequences that may lead to severe accidents shall be identified using a combination of probabilistic methods, deterministic methods and sound engineering judgement. These event sequences shall then be reviewed against a set of criteria aimed at determining which severe accidents should be addressed in the design. Potential design or procedural changes that could either reduce the likelihood of these selected events, or mitigate their consequences, should these selected events occur, shall be evaluated, and shall be implemented if reasonably practicable. Consideration shall be given to the plant full design capabilities, including the possible use of some systems (i.e. safety and non-safety systems) beyond their originally intended function and anticipated operating conditions, and the use of additional temporary systems to return the plant to a controlled state and/or to mitigate the consequences of a severe accident, provided that it can be shown that the systems are able to function in the environmental conditions to be expected. For multiunit plants, consideration shall be given to the use of available means and/or support from other units, provided that the safe operation of the other units is not compromised. Accident management

  15. 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)

  16. Learning lessons from Natech accidents - the eNATECH accident database

    Science.gov (United States)

    Krausmann, Elisabeth; Girgin, Serkan

    2016-04-01

    When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. 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

  2. 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)

  3. Internal event analysis of Laguna Verde Unit 1 Nuclear Power Plant. System Analysis

    International Nuclear Information System (INIS)

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

    1993-01-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis of 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-TR-004 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-004 Volume 5: Appendices A, B and C. This volume presents the results of the system analysis for the Laguna Verde Unit 1 Nuclear Power Plant. The system analysis involved the development of logical models for all the systems included in the accident sequence event tree headings, and for all the support systems required to operate the front line systems. For the Internal Event analysis for Laguna Verde, 16 front line systems and 5 support systems were included. Detailed fault trees were developed for most of the important systems. Simplified fault trees focusing on major faults were constructed for those systems that can be adequately represent,ed using this kind of modeling. For those systems where fault tree models were not constructed, actual data were used to represent the dominant failures of the systems. The main failures included in the fault trees are hardware failures, test and maintenance unavailabilities, common cause failures, and human errors. The SETS and TEMAC codes were used to perform the qualitative and quantitative fault tree analyses. (Author)

  4. An assessment of the risk significance of human errors in selected PSAs and operating events

    International Nuclear Information System (INIS)

    Palla, R.L. Jr.; El-Bassioni, A.

    1991-01-01

    Sensitivity studies based on Probabilistic Safety Assessments (PSAs) for a pressurized water reactor and a boiling water reactor are described. In each case human errors modeled in the PSAs were categorized according to such factors as error type, location, timing, and plant personnel involved. Sensitivity studies were then conducted by varying the error rates in each category and evaluating the corresponding change in total core damage frequency and accident sequence frequency. Insights obtained are discussed and reasons for differences in risk sensitivity between plants are explored. A separate investigation into the role of human error in risk-important operating events is also described. This investigation involved the analysis of data from the USNRC Accident Sequence Precursor program to determine the effect of operator-initiated events on accident precursor trends, and to determine whether improved training can be correlated to current trends. The findings of this study are also presented. 5 refs., 15 figs., 1 tab

  5. 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

  6. Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes

    International Nuclear Information System (INIS)

    Huerta B, A.; Lopez M, R.

    1995-01-01

    This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and.analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

  7. Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes

    International Nuclear Information System (INIS)

    Huerta B, A.; Lopez M, R.

    1995-01-01

    This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. Monitoring Severe Accidents Using AI Techniques

    International Nuclear Information System (INIS)

    No, Young Gyu; Kim, Ju Hyun; Na, Man Gyun; Ahn, Kwang Il

    2011-01-01

    It is very difficult for nuclear power plant operators to monitor and identify the major severe accident scenarios following an initiating event by staring at temporal trends of important parameters. The objective of this study is to develop and verify the monitoring for severe accidents using artificial intelligence (AI) techniques such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH) and fuzzy neural network (FNN). The SVC and PNN are used for event classification among the severe accidents. Also, GMDH and FNN are used to monitor for severe accidents. The inputs to AI techniques are initial time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. In this study, 3 types of initiating events such as the hot-leg LOCA, the cold-leg LOCA and SGTR are considered and it is verified how well the proposed scenario identification algorithm using the GMDH and FNN models identifies the timings when the reactor core will be uncovered, when CET will exceed 1200 .deg. F and when the reactor vessel will fail. In cases that an initiating event develops into a severe accident, the proposed algorithm showed accurate classification of initiating events. Also, it well predicted timings for important occurrences during severe accident progression scenarios, which is very helpful for operators to perform severe accident management

  14. Insights from Severe Accident Analyses for Verification of VVER SAMG

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

  15. 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

  16. Uncertainties and severe-accident management

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1991-01-01

    Severe-accident management can be defined as the use of existing and or alternative resources, systems, and actions to prevent or mitigate a core-melt accident. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies

  17. 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.

  18. SEVERE ACCIDENT ISSUES RAISED BY THE FUKUSHIMA ACCIDENT AND IMPROVEMENTS SUGGESTED

    OpenAIRE

    SONG, JIN HO; KIM, TAE WOON

    2014-01-01

    This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accide...

  19. 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.)

  20. 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

  1. 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/ .

  2. 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.

  3. The Fukushima major accident. Seismic, nuclear and medical considerations; L'accident majeur de Fukushima. Considerations sismiques, nucleaires et medicales

    Energy Technology Data Exchange (ETDEWEB)

    Carpentier, Alain; Friedel, Jacques; Brezin, Edouard; Baulieu, Etienne-Emile; Courtillot, Vincent; Dercourt, Jean; Jaupart, Claude; Le Pichon, Xavier; Poirier, Jean-Paul; Salencon, Jean; Tapponnier, Paul; Dautray, Robert; Taquet, Philippe; Blanchet, Rene; Le Mouel, Jean-Louis; Chapron, Jean-Yves; Fanon, Joelle [Academie des sciences, 23, quai de Conti, 75006 Paris (France); BARD, Pierre-Yves [Observatoire des sciences de l' Univers de l' universite de Grenoble (France); Bernard, Pascal; Montagner, Jean-Paul; Armijo, Rolando; Shapiro, Nikolai; Tait, Steve [Institut de physique du globe de Paris (France); Cara, Michel [ecole et Observatoire des sciences de la Terre de l' universite de Strasbourg (France); Madariaga, Raul [ecole normale superieure, 45, rue d' Ulm / 29, rue d' Ulm, F-75230 Paris cedex 05 (France); Pecker, Alain [Academie des technologies, Grand Palais des Champs Elysees - Porte C - Avenue Franklin D. Roosevelt - 75008 Paris (France); Schindele, Francois [CEA-DAM, Arpajon (France); Douglas, John [BRGM, 3 avenue Claude-Guillemin - BP 36009, 45060 Orleans Cedex 2 (France)

    2011-07-01

    The first part of this voluminous report addresses mega-earthquakes and mega-tsunamis: scientific data, case of France (West Indies and metropolitan France), and socioeconomic aspects (governance, regulation, para-seismic protection). The second part deals with the nuclear accident at Fukushima: event sequence, situation of the nuclear industry in France after Fukushima, fuel cycle and future opportunities. The third part addresses health and environmental consequences. Each part is completed by a large number of documents in which some specific aspects are more precisely reported, commented and discussed

  4. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

    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)

    1993-07-01

    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 for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  5. 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

  6. Safety and risk questions following the nuclear incidents and accidents in Japan. Summary final report; Sicherheits- und Risikofragen im Nachgang zu den nuklearen Stoer- und Unfaellen in Japan. Zusammenfassender Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Mildenberger, Oliver

    2015-03-15

    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).

  7. 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.

  8. 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

  9. Sisifo-gas a computerised system to support severe accident training and management

    International Nuclear Information System (INIS)

    Castro, A.; Buedo, J.L.; Borondo, L.; Lopez, N.

    2001-01-01

    Nuclear Power Plants (NPP) will have to be prepared to face the management of severe accidents, through the development of Severe Accident Guides and sophisticated systems of calculation, as a supporting to the decision-making. SISIFO-GAS is a flexible computerized tool, both for the supporting to accident management and for education and training in severe accident. It is an interactive system, a visual and an easily handle one, and needs no specific knowledge in MAAP code to make complicate simulations in conditions of severe accident. The system is configured and adjusted to work in a BWR/6 technology plant with Mark III Containment, as it is Cofrentes NPP. But it is easily portable to every other kind of reactor, having the level 2 PSA (probabilistic safety analysis) of the plant to be able to establish the categories of the source term and the most important sequences in the progression of the accident. The graphic interface allows following in a very intuitive and formative way the evolution and the most relevant events in the accident, in the both system's way of work, training and management. (authors)

  10. 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

  11. 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

  12. Analysis of the Steam Generator Tubes Rupture Initiating Event

    International Nuclear Information System (INIS)

    Trillo, A.; Minguez, E.; Munoz, R.; Melendez, E.; Sanchez-Perea, M.; Izquierd, J.M.

    1998-01-01

    In PSA studies, Event Tree-Fault Tree techniques are used to analyse to consequences associated with the evolution of an initiating event. The Event Tree is built in the sequence identification stage, following the expected behaviour of the plant in a qualitative way. Computer simulation of the sequences is performed mainly to determine the allowed time for operator actions, and do not play a central role in ET validation. The simulation of the sequence evolution can instead be performed by using standard tools, helping the analyst obtain a more realistic ET. Long existing methods and tools can be used to automatism the construction of the event tree associated to a given initiator. These methods automatically construct the ET by simulating the plant behaviour following the initiator, allowing some of the systems to fail during the sequence evolution. Then, the sequences with and without the failure are followed. The outcome of all this is a Dynamic Event Tree. The work described here is the application of one such method to the particular case of the SGTR initiating event. The DYLAM scheduler, designed at the Ispra (Italy) JRC of the European Communities, is used to automatically drive the simulation of all the sequences constituting the Event Tree. Similarly to the static Event Tree, each time a system is demanded, two branches are open: one corresponding to the success and the other to the failure of the system. Both branches are followed by the plant simulator until a new system is demanded, and the process repeats. The plant simulation modelling allows the treatment of degraded sequences that enter into the severe accident domain as well as of success sequences in which long-term cooling is started. (Author)

  13. Investigation into the March 28, 1979 Three Mile Island accident by Office of Inspection and Enforcement (Investigative Report No. 50-320/79-10)

    International Nuclear Information System (INIS)

    1979-07-01

    On March 28, 1979, the Three Mile Island Unit 2 Nuclear Power Plant experienced the most severe accident in U.S. commercial nuclear power plant operating history. This report sets forth the facts concerning the events of the accident determined as a result of an investigation by the NRC Office of Inspection and Enforcement. The IE investigation is limited to two aspects of the accident: (1) Those related operational actions by the licensee during the period from before the initiating event until approximately 8:00 p.m., March 28, when primary coolant flow was re-established by starting a reactor coolant pump, and (2) Those steps taken by the licensee to control the release of radioactive material to the off-site environs, and to implement his emergency plan during the period from the initiation of the event to midnight, March 30. These investigation periods were selected because they include the licensee actions which most significantly affected the accident sequence and its results

  14. The Chernobyl accident consequences; Consequences de l'accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.)

  15. 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)

  16. Monitoring severe accidents using AI techniques

    Energy Technology Data Exchange (ETDEWEB)

    No, Young Gyu; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of); Lim, Dong Hyuk [Korea Institute of Nuclear Nonproliferation and Control, Daejon (Korea, Republic of)

    2012-05-15

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  17. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

    No, Young Gyu; Ahn, Kwang Il; Kim, Ju Hyun; Na, Man Gyun; Lim, Dong Hyuk

    2012-01-01

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  18. Probabilistic approach in treatment of deterministic analyses results of severe accidents

    International Nuclear Information System (INIS)

    Krajnc, B.; Mavko, B.

    1996-01-01

    Severe accidents sequences resulting in loss of the core geometric integrity have been found to have small probability of the occurrence. Because of their potential consequences to public health and safety, an evaluation of the core degradation progression and the resulting effects on the containment is necessary to determine the probability of a significant release of radioactive materials. This requires assessment of many interrelated phenomena including: steel and zircaloy oxidation, steam spikes, in-vessel debris cooling, potential vessel failure mechanisms, release of core material to the containment, containment pressurization from steam generation, or generation of non-condensable gases or hydrogen burn, and ultimately coolability of degraded core material. To asses the answer from the containment event trees in the sense of weather certain phenomenological event would happen or not the plant specific deterministic analyses should be performed. Due to the fact that there is a large uncertainty in the prediction of severe accidents phenomena in Level 2 analyses (containment event trees) the combination of probabilistic and deterministic approach should be used. In fact the result of the deterministic analyses of severe accidents are treated in probabilistic manner due to large uncertainty of results as a consequence of a lack of detailed knowledge. This paper discusses approach used in many IPEs, and which assures that the assigned probability for certain question in the event tree represent the probability that the event will or will not happen and that this probability also includes its uncertainty, which is mainly result of lack of knowledge. (author)

  19. 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.

  20. 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)

  1. 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.

  2. 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%.

  3. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  4. Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

    International Nuclear Information System (INIS)

    Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

    1982-01-01

    A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

  5. 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

  6. Treatment of whole-body radiation accident victims

    International Nuclear Information System (INIS)

    Drum, D.E.; Rappeport, J.M.

    1990-01-01

    This paper discusses how whole-body radiation exposure incidents present a number of unique challenges. The acute, nonstochastic effects of high doses of radiation over 25 rads (0.25 Gy) delivered to humans is generally manifest in rather categorical fashion; depending on the dose, either the patient is largely unharmed functionally or he is seriously injured. Radiation initiates microchemical changes within a nanosecond time frame; there exists no specific therapy to stop or reverse the sequence of events that follow. Thus, the range for effective therapeutic intervention is rather small, between 150 to 1500 rad (1.5 to 15 Gy) for humans. Nevertheless, it is likely that a large uncomplicated exposure to as much as 750 rad (7.5 Gy) might be survivable without dramatic measures such as bone marrow transplantation. Review of the available information about past accidents shows that the majority of radiation accidents are mixed injuries

  7. 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)

  8. 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)

  9. 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

  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. 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

  11. 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

  12. Environmental impacts of radiological consequences during the anticipated transients without scram (ATWS) events in nuclear power reactors

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    2011-01-01

    Anticipated transients without scram (ATWS), is one of the (worst case) accidents could happen if the system that provides a highly reliable means of shutting down the reactor (scram system )fails to work during a reactor event (anticipated transient).It has two general characteristics: (1) Initiation by a transient anticipated to occur one or more times in the life of reactor and ,(2) Assumed to proceed without scram.The types of events considered are those used for designing the plant .The evaluation of the radiological consequences during the assessment of the nuclear events,especially ATWS in nuclear power reactors, is very essential for environmental studies and public safety. In this paper, the root cases for nuclear events and dose calculation are presented. Scenario of accident sequences together with radiological impacts is illustrated for loss of coolant accident (LOCA) for a typical pressurized water reactor nuclear power plant. Recommendations for mitigating or preventing the release of radiation and high radioactive materials to environment are presented.

  13. 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

  14. Probabilistic risk assessment (PRA) update in light of the accident at Fukushima Daiichi Nuclear Power Station - 15461

    International Nuclear Information System (INIS)

    Maeda, K.; Abe, H.; Hirokawa, N.; Satou, C.

    2015-01-01

    We have performed internal and external event probabilistic risk assessments (PRA) for boiling water reactor power nuclear plants to identify the important accident sequence groups and to evaluate the effectiveness of the additional severe accident measures, regarding to the new regulatory requirements implemented after the accident at Fukushima Daiichi Nuclear Power Station in Japan in 2011. In addition, we will further update our PRA by extracting problems and improvements from the current PRA, by catching up the state-of-the-art knowledge, modern PRA methodologies in order to contribute voluntarily to safety improvement as well as to comply with regulations. In this document, prior to the extensive PRA updates, we would describe technical contents and qualitative results about PRA updates that have been performed preliminary so far, especially about the external event (seismic) PRA and how to model the additionally deployed severe accident measures (e.g. power supply car, fire engine) so that they can be function external hazards, such as component failure rate of equipment, human reliability 'out of control room', and mission time extension. (authors)

  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 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

  17. TITAN: a computer program for accident occurrence frequency analyses by component Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi [Department of Fuel Cycle Safety Research, 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); Kanai, Shigeru [Fuji Research Institute Corporation, Tokyo (Japan)

    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)

  18. 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)

  19. 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)

  20. 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

  1. 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.)

  2. 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)

  3. 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

  4. The Fukushima major accident. Seismic, nuclear and medical considerations

    International Nuclear Information System (INIS)

    Carpentier, Alain; Friedel, Jacques; Brezin, Edouard; Baulieu, Etienne-Emile; Courtillot, Vincent; Dercourt, Jean; Jaupart, Claude; Le Pichon, Xavier; Poirier, Jean-Paul; Salencon, Jean; Tapponnier, Paul; Dautray, Robert; Taquet, Philippe; Blanchet, Rene; Le Mouel, Jean-Louis; Chapron, Jean-Yves; Fanon, Joelle; BARD, Pierre-Yves; Bernard, Pascal; Montagner, Jean-Paul; Armijo, Rolando; Shapiro, Nikolai; Tait, Steve; Cara, Michel; Madariaga, Raul; Pecker, Alain; Schindele, Francois; Douglas, John

    2011-01-01

    The first part of this voluminous report addresses mega-earthquakes and mega-tsunamis: scientific data, case of France (West Indies and metropolitan France), and socioeconomic aspects (governance, regulation, para-seismic protection). The second part deals with the nuclear accident at Fukushima: event sequence, situation of the nuclear industry in France after Fukushima, fuel cycle and future opportunities. The third part addresses health and environmental consequences. Each part is completed by a large number of documents in which some specific aspects are more precisely reported, commented and discussed

  5. Identification of the operating crew's information needs for accident management

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. 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)

  7. 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

  8. 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)

  9. Transport and release of fission products during nuclear reactor accident

    International Nuclear Information System (INIS)

    Lee, K.W.; Kuhlman, M.R.; Gieseke, J.A.

    1984-01-01

    This study represents the identification and formulation of a systematic, mechanistic approach to estimating source terms and the implementation of this approach through calculations of fission products release to the environment for a large PWR reactor under a selected set of accident conditions. The development and improvement of calculational procedures is an evolutionary process and in the long term must be verified through experimental studies. It is anticipated that as additional information is obtained the accuracy of predictions can be improved and uncertainties reduced. Transport and deposition of radionuclides were found to be quite dependent on the accident sequences and the corresponding thremal hydraulic conditions. Reduced temperatures led to increased deposition of vapor species, and reduced flow rates to increased aerosol deposition. It is to be recognized that the estimates of release fractions are subject to uncertainties in the data and computer models employed in the calculations and are expected to have been influenced by assumptions regarding plant geometry, thermal hydraulics, deposition mechanisms, and sequence events. The effects of these assumptions will be investigated as this study continues. (Author)

  10. A comprehensive review of rollover accidents involving vehicles equipped with Electronic Stability Control (ESC) systems.

    Science.gov (United States)

    Padmanaban, Jeya; Shields, Leland E; Scheibe, Robert R; Eyges, Vitaly E

    2008-10-01

    This study investigated 478 police accident reports from 9 states to examine and characterize rollover crashes involving ESC-equipped vehicles. The focus was on the sequence of critical events leading to loss of control and rollover, and the interactions between the accident, driver, and environment. Results show that, while ESC is effective in reducing loss of control leading to certain rollover crashes, its effectiveness is diminished in others, particularly when the vehicle departs the roadway or when environmental factors such as slick road conditions or driver factors such as speeding, distraction, fatigue, impairment, or overcorrection are present.

  11. Analysis of two different types of hydrogen combustion during severe accidents in a typical pressurized water reactor

    International Nuclear Information System (INIS)

    Ko Yuchih; Lee Min

    2005-01-01

    Hydrogen combustion is an important phenomenon that may occur during severe accidents of Nuclear Power Plants (NPPs). Depending on the specific plant design, the initiating events, and mitigation actions executed, hydrogen combustion may have distinct characteristics and may damage the plant in various degrees. The worst scenario will be the catastrophic failure of containment. In this study two specific types of hydrogen combustion are analyzed to evaluate their impact on the containment integrity. In this paper, Station Blackout (SBO) and Loss of Coolant Accidents (LOCAs) sequences are analyzed using MAAP4 (Modular Accident Analysis Program) code. The former sequence is used to represent hydrogen combustion phenomenon under the condition that the reactor pressure vessel (RPV) breaches at high pressure and the latter sequence represents the phenomenon that RPV fails at low pressure. Two types of hydrogen combustion are observed in the simulation. The Type I hydrogen combustion represents global and instantaneous hydrogen combustion. Large pressure spike is created during the combustion and represents a threat to containment integrity. Type II hydrogen combustion is localized burn and burn continuously over a time period. There is hardly any impact of this type hydrogen burn on the containment pressurization rate. Both types of hydrogen combustion can occur in the severe accidents without any human intervention. From the accident mitigation point of view, operators should try to bring the containment into conditions that favor the Type II hydrogen combustion. (authors)

  12. 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

  13. 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

  14. 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.)

  15. 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.

  16. Industrial accidents triggered by lightning.

    Science.gov (United States)

    Renni, Elisabetta; Krausmann, Elisabeth; Cozzani, Valerio

    2010-12-15

    Natural disasters can cause major accidents in chemical facilities where they can lead to the release of hazardous materials which in turn can result in fires, explosions or toxic dispersion. Lightning strikes are the most frequent cause of major accidents triggered by natural events. In order to contribute towards the development of a quantitative approach for assessing lightning risk at industrial facilities, lightning-triggered accident case histories were retrieved from the major industrial accident databases and analysed to extract information on types of vulnerable equipment, failure dynamics and damage states, as well as on the final consequences of the event. The most vulnerable category of equipment is storage tanks. Lightning damage is incurred by immediate ignition, electrical and electronic systems failure or structural damage with subsequent release. Toxic releases and tank fires tend to be the most common scenarios associated with lightning strikes. Oil, diesel and gasoline are the substances most frequently released during lightning-triggered Natech accidents. Copyright © 2010 Elsevier B.V. All rights reserved.

  17. 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)

  18. 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

  19. 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...

  20. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  1. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  2. 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

  3. 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

  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. Estimation of cost per severe accident for improvement of accident protection and consequence mitigation strategies

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2013-01-01

    To assess the complex situations regarding the severe accidents such as what observed in Fukushima Accident, not only radiation protection aspects but also relevant aspects: health, environmental, economic and societal aspects; must be all included into the consequence assessment. In this study, the authors introduce the “cost per severe accident” as an index to analyze the consequences of severe accidents comprehensively. The cost per severe accident consists of various costs and consequences converted into monetary values. For the purpose of improvement of the accident protection and consequence mitigation strategies, the costs needed to introduce the protective actions, and health and psychological consequences are included in the present study. The evaluations of these costs and consequences were made based on the systematic consequence analysis using level 2 and 3 probabilistic safety assessment (PSA) codes. The accident sequences used in this analysis were taken from the results of level 2 seismic PSA of a virtual 1,100 MWe BWR-5. The doses to the public and the number of people affected were calculated using the level 3 PSA code OSCAAR of Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the outputs are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated based on the open documents on the Fukushima Accident regarding the cost of protective actions and compensations for psychological harms. Finally, optimized accident protection and consequence mitigation strategies are recommended taking into account the various aspects comprehensively using the cost per severe accident. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. (author)

  6. 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

  7. 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...

  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. 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

  10. 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)

  11. MAAP - modular program for analyses of severe accidents

    International Nuclear Information System (INIS)

    Henry, R.E.; Lutz, R.J.

    1990-01-01

    The MAAP computer code was developed by Westinghouse as a fast, user-friendly, integrated analytical tool for evaluations of the sequences and consequences of severe accidents. The code allows a fully integrated treatment of thermohydraulic behavior and of the fission products in the primary system, the containment, and the ancillary buildings. This ensures interactive inclusion of all thermohydraulic events and of fission product behavior. All important phenomena which may occur in a major accident are contained in the modular code. In addition, many of the important parameters affecting the multitude of different phenomena can be defined by the user. In this way, it is possible to study the accuracy of the predicted course and of the consequences of a series of major accident phenomena. The MAAP code was subjected to extensive benchmarking with respect to the results of the experimental and theoretical programs, the findings obtained in other safety analyses using computers and data from accidents and transients in plants actually in operation. With the expected connection of the validation and test programs, the computer code attains a quality standard meeting the most stringent requirements in safety analyses. The code will be enlarged further in order to expand the number of benchmarks and the resolution of individual comparisons, and to ensure that future MAAP models will be in better agreement with the experiments and experiences of industry. (orig.) [de

  12. Emergency feature. Great east Japan earthquake disaster Fukushima Daiichi accident

    International Nuclear Information System (INIS)

    Kawata, Tomio; Tsujikura, Yonezo; Kitamura, Toshiro

    2011-01-01

    The Tohoku Pacific Ocean earthquake occurred in March 11, 2011. The disastrous tsunami attacked Fukushima Daiichi nuclear power plants after automatically shutdown by the earthquake and all motor operated pumps became inoperable due to station black out. Despite the strenuous efforts of operators, if caused serious accident such as loss of cooling function, hydrogen explosion and release of large amount of radioactive materials into the environment, leading to nuclear power emergency that ordered resident to evacuate or remain indoors. This emergency feature consisted of four articles. The first was the interview with the president of JAIF (Japan Atomic Industrial Forum) on how to identify the cause of the accident completely, intensify safety assurance measures and promote discussions on a role of nuclear power in the nation's entire energy policy toward the reconstruction. Others were reactor states and events sequence after the accident with trend data of radiation in the reactor site, statement of president of AESJ (Atomic Energy Society of Japan) on nuclear crisis following Tohoku Pacific Ocean earthquake our response and my experience in evacuation life. (T. Tanaka)

  13. Radiological accident 'The Citadel' medical aspects

    International Nuclear Information System (INIS)

    Cardenas Herrera, Juan; Fernandez, Isis M.; Lopez, Gladys; Garcia, Omar; Lamadrid, Ana I.; Ramos, Enma O.; Villa, Rosario; Giron, Carmen M.; Escobar, Myrian; Zerpa, Miguel; Romero, Argenis H.; Medina, Julio; Laurenti, Zenia; Oliva, Maria T.; Sierra, Nitza; Lorenzo, Alexis

    2008-01-01

    The work exposes the medical actions carried out in the mitigation of the consequences of the accident and its main results. In a facility of storage of radioactive waste in Caracas, Venezuela, it was happened a radiological accident. This event caused radioactive contamination of the environment, as well as the irradiation and radioactive contamination of at least 10 people involved in the fact, in its majority children. Cuban institutions participated in response to the accident. Among the decisions adopted by the team of combined work Cuban-Venezuelan, we find the one of transferring affected people to Cuba, for their dosimetric and medical evaluation. Being designed a work strategy to develop the investigations to people affected by the radiological accident, in correspondence with the circumstances, magnitude and consequences of the accident. The obtained main results are: 100% presented affectations in its health, not associate directly to the accident, although the accident influenced in its psychological state. In 3 of studied people they were detected radioactive contamination with Cesium -137 with dose among 2.01 X 10-4 Sv up to 2.78 X 10-4 Sv. This accident demonstrated the necessity to have technical capacities to face these events and the importance of the international solidarity. (author)

  14. Evaluation of potential severe accidents during low power and shutdown operations at Surry: Unit 1, Volume 1

    International Nuclear Information System (INIS)

    Chu, T.L.; Pratt, W.T.

    1995-10-01

    This document contains a summarization of the results and insights from the Level 1 accident sequence analyses of internally initiated events, internally initiated fire and flood events, seismically initiated events, and the Level 2/3 risk analysis of internally initiated events (excluding fire and flood) for Surry, Unit 1. The analysis was confined to mid-loop operation, which can occur during three plant operational states (identified as POSs R6 and R10 during a refueling outage, and POS D6 during drained maintenance). The report summarizes the Level 1 information contained in Volumes 2--5 and the Level 2/3 information contained in Volume 6 of NUREG/CR-6144

  15. Evaluation of potential severe accidents during low power and shutdown operations at Surry: Unit 1, Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.L.; Pratt, W.T. [eds.; Musicki, Z. [Brookhaven National Lab., Upton, NY (United States)

    1995-10-01

    This document contains a summarization of the results and insights from the Level 1 accident sequence analyses of internally initiated events, internally initiated fire and flood events, seismically initiated events, and the Level 2/3 risk analysis of internally initiated events (excluding fire and flood) for Surry, Unit 1. The analysis was confined to mid-loop operation, which can occur during three plant operational states (identified as POSs R6 and R10 during a refueling outage, and POS D6 during drained maintenance). The report summarizes the Level 1 information contained in Volumes 2--5 and the Level 2/3 information contained in Volume 6 of NUREG/CR-6144.

  16. 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)

  17. Regulatory approach to enhanced human performance during accidents

    International Nuclear Information System (INIS)

    Palla, R.L. Jr.

    1990-01-01

    It has become increasingly clear in recent years that the risk associated with nuclear power is driven by human performance. Although human errors have contributed heavily to the two core-melt events that have occurred at power reactors, effective performance during an event can also prevent a degraded situation from progressing to a more serious accident, as in the loss-of-feedwater event at Davis-Besse. Sensitivity studies in which human error rates for various categories of errors in a probabilistic risk assessment (PRA) were varied confirm the importance of human performance. Moreover, these studies suggest that actions taken during an accident are at least as important as errors that occur prior to an initiating event. A program that will lead to enhanced accident management capabilities in the nuclear industry is being developed by the US Nuclear Regulatory Commission (NRC) and industry and is a key element in NRC's integration plan for closure of severe-accident issues. The focus of the accident management (AM) program is on human performance during accidents, with emphasis on in-plant response. The AM program extends the defense-in-depth principle to plant operating staff. The goal is to take advantage of existing plant equipment and operator skills and creativity to find ways to terminate accidents that are beyond the design basis. The purpose of this paper is to describe the NRC's objectives and approach in AM as well as to discuss several human performance issues that are central to AM

  18. From learning from accidents to teaching about accident causation and prevention: Multidisciplinary education and safety literacy for all engineering students

    International Nuclear Information System (INIS)

    Saleh, Joseph H.; Pendley, Cynthia C.

    2012-01-01

    In this work, we argue that system accident literacy and safety competence should be an essential part of the intellectual toolkit of all engineering students. We discuss why such competence should be taught and nurtured in engineering students, and provide one example for how this can be done. We first define the class of adverse events of interest as system accidents, distinct from occupational accidents, through their (1) temporal depth of causality and (2) diversity of agency or groups and individuals who influence or contribute to the accident occurrence/prevention. We then address the question of why the interest in this class of events and their prevention, and we expand on the importance of system safety literacy and the contributions that engineering students can make in the long-term towards accident prevention. Finally, we offer one model for an introductory course on accident causation and system safety, discuss the course logistics, material and delivery, and our experience teaching this subject. The course starts with the anatomy of accidents and is grounded in various case studies; these help illustrate the multidisciplinary nature of the subject, and provide the students with the important concepts to describe the phenomenology of accidents (e.g., initiating events, accident precursor or lead indicator, and accident pathogen). More importantly, the case studies invite a deep reflection on the underlying failure mechanisms, their generalizability, and the various safety levers for accident prevention. The course then proceeds to an exposition of defense-in-depth, safety barriers and principles, essential elements for an education in accident prevention, and it concludes with a presentation of basic concepts and tools for uncertainty and risk analysis. Educators will recognize the difficulties in designing a new course on such a broad subject. It is hoped that this work will invite comments and contributions from the readers, and that the journal will

  19. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

    International Nuclear Information System (INIS)

    Escorcia O, D.; Salazar S, E.

    2016-09-01

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  20. 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)

  1. 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

  2. Catalogue of generic plant states leading to core melt in PWRs: includes appendix 1: detailed description of sequences leading to core melt

    International Nuclear Information System (INIS)

    1996-11-01

    The Task Group on thermal-hydraulic system behaviour was given a mandate from PWG 2 on Coolant System-Behaviour with the approval of CSNI to deal with the topic of Accident Management. A writing group was set up to identify generic plant states leading to core melt for pressurized water reactors (PWR) and find 'possible approaches to accident management measures' (AM-Measures) for dealing with them. From a matrix of 15 initiating events and 12 system failures (i.e. from 180 possibilities), 32 event sequences have been identified as leading to core melt. Each sequence has been divided into characteristic plant state intervals according to safety function challenges. For each of the 141 defined characteristic plant state intervals, the members of the Writing Group made proposals for AM-Measures

  3. 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

  4. 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.

  5. 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)

  6. 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)

  7. Statistical evaluation of design-error related nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ott, K.O.; Marchaterre, J.F.

    1981-01-01

    In this paper, general methodology for the statistical evaluation of design-error related accidents is proposed that can be applied to a variety of systems that evolves during the development of large-scale technologies. The evaluation aims at an estimate of the combined ''residual'' frequency of yet unknown types of accidents ''lurking'' in a certain technological system. A special categorization in incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of U.S. nuclear power reactor technology, considering serious accidents (category 2 events) that involved, in the accident progression, a particular design inadequacy. 9 refs

  8. 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.

  9. Probabilistic Accident Progression Analysis with application to a LMFBR design

    International Nuclear Information System (INIS)

    Jamali, K.M.

    1982-01-01

    A method for probabilistic analysis of accident sequences in nuclear power plant systems referred to as ''Probabilistic Accident Progression Analysis'' (PAPA) is described. Distinctive features of PAPA include: (1) definition and analysis of initiator-dependent accident sequences on the component level; (2) a new fault-tree simplification technique; (3) a new technique for assessment of the effect of uncertainties in the failure probabilities in the probabilistic ranking of accident sequences; (4) techniques for quantification of dependent failures of similar components, including an iterative technique for high-population components. The methodology is applied to the Shutdown Heat Removal System (SHRS) of the Clinch River Breeder Reactor Plant during its short-term (0 -2 . Major contributors to this probability are the initiators loss of main feedwater system, loss of offsite power, and normal shutdown

  10. 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...

  11. 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

  12. 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.

  13. An analysis of Three Mile Island: the accident that shouldn't have happened

    International Nuclear Information System (INIS)

    Rubinstein, E.; Mason, J.F.

    1979-01-01

    The sequence of events in the nuclear reactor accident at Three Mile Island on March 28, 1979, is reported. Three problems thought to trigger the reactor accident were a persistent leak of reactor coolant, a closing of two valves in the auxillary feedwater system, and an apparent resin blockage in the transfer line that forced water back into the condensate lines of the air pumps. Hindsight indicates that a large amount of the damage to the reactor core could have been prevented if operators had closed the electromatic relief valve to end the loss of coolant and not throttled down the high pressure injection pumps in the emergency core cooling system. Steps taken to reestablish control of the reactor core are described

  14. 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

  15. Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

    2006-07-01

    A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

  16. LMFBR accident delineation study: approach and preliminary results

    International Nuclear Information System (INIS)

    Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

    1979-01-01

    Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

  17. 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

  18. 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

  19. A review for identification of initiating events in event tree development process on nuclear power plants

    International Nuclear Information System (INIS)

    Riyadi, Eko H.

    2014-01-01

    Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events

  20. 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)

  1. Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

    International Nuclear Information System (INIS)

    Petkov, Gueorgui

    2010-01-01

    Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

  2. MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

    International Nuclear Information System (INIS)

    Choi, Wonjun; Jeon, Joongoo; Kim, Nam Kyung; Kim, Sung Joong

    2017-01-01

    The representative example of the severe accident studies using the severe accident code is investigation of effectiveness of developed severe accident management (SAM) strategy considering the positive and adverse effects. In Korea, some numerical studies were performed to investigate the SAM strategy using various severe accident codes. Seo et.al performed validation of RCS depressurization strategy and investigated the effect of severe accident management guidance (SAMG) entry condition under small break loss of coolant accident (SBLOCA) without safety injection (SI), station blackout (SBO), and total loss of feed water (TLOFW) scenarios. The SGTR accident with the sequential mitigation actions according to the flow chart of SAMG was simulated by the MELCOR 1.8.6 code. Three scenariospreventing the RPV failure were investigated in terms of fission product release, hydrogen risk, and the containment pressure. Major conclusions can be summarized as follows: (1) According to the flow chart of SAMG, RPV failure can be prevented depending on the method of RCS depressurization. (2) To reduce the release of fission product during the injecting into SGs, a temporary opening of SDS before the injecting into SGs was suggested. These modified sequences of mitigation actions can reduce the release of fission product and the adverse effect of SDS.

  3. International aspects of nuclear accidents

    International Nuclear Information System (INIS)

    Uematsu, K.

    1989-09-01

    The accident at Chernobyl revealed that there were shortcomings and gaps in the existing international mechanisms and brought home to governments the need for stronger measures to provide better protection against the risks of severe accidents. The main thrust of international co-operation with regard to nuclear safety issues is aimed at achieving a uniformly high level of safety in nuclear power plants through continuous exchanges of research findings and feedback from reactor operating experience. The second type of problem posed in the event of an accident resulting in radioactive contamination of several countries relates to the obligation to notify details of the circumstances and nature of the accident speedily so that the countries affected can take appropriate protective measures and, if necessary, organize mutual assistance. Giving the public accurate information is also an important aspect of managing an emergency situation arising from a severe accident. Finally, the confusion resulting from the unwarranted variety of protective measures implemented after the Chernobyl accident has highlighted the need for international harmonization of the principles and scientific criteria applicable to the protection of the public in the event of an accident and for a more consistent approach to emergency plans. The international conventions on third party liability in the nuclear energy sector (Paris/Brussels Conventions and the Vienna Convention) provide for compensation for damage caused by nuclear accidents in accordance with the rules and jurisdiction that they lay down. These provisions impose obligations on the operator responsible for an accident, and the State where the nuclear facility is located, towards the victims of damage caused in another country

  4. The United States Department of Energy (DOE) Computerized Accident/Incident Reporting System (CAIRS)

    International Nuclear Information System (INIS)

    Briscoe, G.J.

    1993-01-01

    The Department of Energy's (DOE) Computerized Accident/Incident Reporting System (CAIRS) is a comprehensive data base containing more than 50,000 investigation reports of injury/illness, property damage and vehicle accident cases representing safety data from 1975 to the present for more than 150 DOE contractor organizations. A special feature is that the text of each accident report is translated using a controlled dictionary and rigid sentence structure called Factor Relationship and Sequence of Events (FRASE) that enhances the ability to retrieve specific types of information and to perform detailed analyses. DOE summary and individual contractor reports are prepared quarterly and annually. In addition, ''Safety Performance Profile'' reports for individual organizations are prepared to provide advance information to appraisal teams, and special topical reports are prepared for areas of concern such as an increase in the number of security injuries or environmental releases. The data base is open to all DOE and Contractor registered users with no access restrictions other than that required by the Privacy Act

  5. Methods for external event screening quantification: Risk Methods Integration and Evaluation Program (RMIEP) methods development

    International Nuclear Information System (INIS)

    Ravindra, M.K.; Banon, H.

    1992-07-01

    In this report, the scoping quantification procedures for external events in probabilistic risk assessments of nuclear power plants are described. External event analysis in a PRA has three important goals; (1) the analysis should be complete in that all events are considered; (2) by following some selected screening criteria, the more significant events are identified for detailed analysis; (3) the selected events are analyzed in depth by taking into account the unique features of the events: hazard, fragility of structures and equipment, external-event initiated accident sequences, etc. Based on the above goals, external event analysis may be considered as a three-stage process: Stage I: Identification and Initial Screening of External Events; Stage II: Bounding Analysis; Stage III: Detailed Risk Analysis. In the present report, first, a review of published PRAs is given to focus on the significance and treatment of external events in full-scope PRAs. Except for seismic, flooding, fire, and extreme wind events, the contributions of other external events to plant risk have been found to be negligible. Second, scoping methods for external events not covered in detail in the NRC's PRA Procedures Guide are provided. For this purpose, bounding analyses for transportation accidents, extreme winds and tornadoes, aircraft impacts, turbine missiles, and chemical release are described

  6. 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...

  7. 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...

  8. Risk analysis of releases from accidents during mid-loop operation at Surry

    International Nuclear Information System (INIS)

    Jo, J.; Lin, C.C.; Nimnual, S.; Mubayi, V.; Neymotin, L.

    1992-11-01

    Studies and operating experience suggest that the risk of severe accidents during low power operation and/or shutdown (LP/S) conditions could be a significant fraction of the risk at full power operation. Two studies have begun at the Nuclear Regulatory Commission (NRC) to evaluate the severe accident progression from a risk perspective during these conditions: One at the Brookhaven National Laboratory for the Surry plant, a pressurized water reactor (PWR), and the other at the Sandia National Laboratories for the Grand Gulf plant, a boiling water reactor (BWR). Each of the studies consists of three linked, but distinct, components: a Level I probabilistic risk analysis (PRA) of the initiating events, systems analysis, and accident sequences leading to core damage; a Level 2/3 analysis of accident progression, fuel damage, releases, containment performance, source term and consequences-off-site and on-site; and a detailed Human Reliability Analysis (HRA) of actions relevant to plant conditions during LP/S operations. This paper summarizes the approach taken for the Level 2/3 analysis at Surry and provides preliminary results on the risk of releases and consequences for one plant operating state, mid-loop operation, during shutdown

  9. A database system for the management of severe accident risk information, SARD

    International Nuclear Information System (INIS)

    Ahn, K. I.; Kim, D. H.

    2003-01-01

    The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies

  10. A database system for the management of severe accident risk information, SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K. I.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies.

  11. Environmental impact analysis for the main accidental sequences of ignitor

    International Nuclear Information System (INIS)

    Carpignano, A.; Francabandiera, S.; Vella, R.; Zucchetti, M.

    1996-01-01

    A safety analysis study has been applied to the Ignitor machine using Probabilistic Safety Assessment. The main initiating events have been identified, and accident sequences have been studied by means of traditional methods such as Failure Mode and Effect Analysis (FMEA), Fault Trees (FT) and Event Trees (ET). The consequences of the radioactive environmental releases have been assessed in terms of Effective Dose Equivalent (EDEs) to the Most Exposed Individuals (MEI) of the chosen site, by means of a population dose code. Results point out the low enviromental impact of the machine. 13 refs., 1 fig., 3 tabs

  12. 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

  13. MELCOR 1.8.2 calculations of selected sequences for the ABWR

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1994-07-01

    This report summarizes the results from MELCOR calculations of severe accident sequences in the ABWR and presents comparisons with MAAP calculations for the same sequences. MELCOR was run for two low-pressure and three high-pressure sequences to identify the materials which enter containment and are available for release to the environment (source terms), to study the potential effects of core-concrete interaction, and to obtain event timings during each sequence; the source terms include fission products and other materials such as those generated by core-concrete interactions. Sensitivity studies were done on the impact of assuming limestone rather than basaltic concrete and on the effect of quenching core debris in the cavity compared to having hot, unquenched debris present

  14. 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.)

  15. Analysis on the nitrogen drilling accident of Well Qionglai 1 (II: Restoration of the accident process and lessons learned

    Directory of Open Access Journals (Sweden)

    Yingfeng Meng

    2015-12-01

    Full Text Available All the important events of the accident of nitrogen drilling of Well Qionglai 1 have been speculated and analyzed in the paper I. In this paper II, based on the investigating information, the well log data and some calculating and simulating results, according to the analysis method of the fault tree of safe engineering, the every possible compositions, their possibilities and time schedule of the events of the accident of Well Qionglai 1 have been analyzed, the implications of the logging data have been revealed, the process of the accident of Well Qionglai 1 has been restored. Some important understandings have been obtained: the objective causes of the accident is the rock burst and the induced events form rock burst, the subjective cause of the accident is that the blooie pipe could not bear the flow burden of the clasts from rock burst and was blocked by the clasts. The blocking of blooie pipe caused high pressure in wellhead, the high pressure made the blooie pipe burst, natural gas came out and flared fire. This paper also thinks that the rock burst in gas drilling in fractured tight sandstone gas zone is objective and not avoidable, but the accidents induced from rock burst can be avoidable by improving the performance of the blooie pipe, wellhead assemblies and drilling tool accessories aiming at the downhole rock burst.

  16. A defense in depth approach for nuclear power plant accident management

    Energy Technology Data Exchange (ETDEWEB)

    Chih-Yao Hsieh; Hwai-Pwu Chou [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu, TW (China)

    2015-07-01

    An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident

  17. 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

  18. The Chernobyl accidents: Causes and Consequences

    International Nuclear Information System (INIS)

    Chihab-Eddine, A.

    1988-01-01

    The objective of this communication is to discuss the causes and the consequences of the Chernobyl accident. To facilitate the understanding of the events that led to the accident, the author gave a simplified introduction to the important physics that goes on in a nuclear reactor and he presented a brief description and features of chernobyl reactor. The accident scenario and consequences have been presented. The common contribution factors that led to both Three Mile Island and Chernobyl accidents have been pointed out.(author)

  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. 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.

  1. Golfech plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Golfech plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  2. Tricastin plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Tricastin plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  3. Bugey plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Bugey plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  4. Fessenheim plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Fessenheim plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  5. Chinon plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Chinon B plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  6. Blayais plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Blayais plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  7. Civaux plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Civaux plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  8. Cattenom plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Cattenom plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  9. Gravelines plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Gravelines plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  10. 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

  11. Source terms derived from analyses of hypothetical accidents, 1950-1986

    International Nuclear Information System (INIS)

    Stratton, W.R.

    1987-01-01

    This paper reviews the history of reactor accident source term assumptions. After the Three Mile Island accident, a number of theoretical and experimental studies re-examined possible accident sequences and source terms. Some of these results are summarized in this paper

  12. 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...

  13. 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.

  14. ANS severe accident program overview & planning document

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.

    1995-09-01

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

  15. Internal Accident Report: fill it out!

    CERN Multimedia

    2012-01-01

    It is important to report all accidents, near-misses and dangerous situations so that they can be avoided in the future.   Reporting these events allows the relevant services to take appropriate action and implement corrective and preventive measures. It should be noted that the routing of the internal accident report was recently changed to make sure that the people who need to know are informed. Without information, corrective action is not possible. Without corrective action, there is a risk that the events will recur. As soon as you experience or see something amiss, fill out an internal accident report! If you have any questions the HSE Unit will be happy to answer them. Contact us at safety-general@cern.ch. The HSE Unit

  16. French policy for managing the post-accident phase of a nuclear accident.

    Science.gov (United States)

    Gallay, F; Godet, J L; Niel, J C

    2015-06-01

    In 2005, at the request of the French Government, the Nuclear Safety Authority (ASN) established a Steering Committee for the Management of the Post-Accident Phase of a Nuclear Accident or a Radiological Emergency, with the objective of establishing a policy framework. Under the supervision of ASN, this Committee, involving several tens of experts from different backgrounds (e.g. relevant ministerial offices, expert agencies, local information commissions around nuclear installations, non-governmental organisations, elected officials, licensees, and international experts), developed a number of recommendations over a 7-year period. First published in November 2012, these recommendations cover the immediate post-emergency situation, and the transition and longer-term periods of the post-accident phase in the case of medium-scale nuclear accidents causing short-term radioactive release (less than 24 h) that might occur at French nuclear facilities. They also apply to actions to be undertaken in the event of accidents during the transportation of radioactive materials. These recommendations are an important first step in preparation for the management of a post-accident situation in France in the case of a nuclear accident. © The Chartered Institution of Building Services Engineers 2014.

  17. Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

    International Nuclear Information System (INIS)

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

    2015-01-01

    This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

  18. Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

  19. Risk-based ranking of dominant contributors to maritime pollution events

    International Nuclear Information System (INIS)

    Wheeler, T.A.

    1993-01-01

    This report describes a conceptual approach for identifying dominant contributors to risk from maritime shipping of hazardous materials. Maritime transportation accidents are relatively common occurrences compared to more frequently analyzed contributors to public risk. Yet research on maritime safety and pollution incidents has not been guided by a systematic, risk-based approach. Maritime shipping accidents can be analyzed using event trees to group the accidents into 'bins,' or groups, of similar characteristics such as type of cargo, location of accident (e.g., harbor, inland waterway), type of accident (e.g., fire, collision, grounding), and size of release. The importance of specific types of events to each accident bin can be quantified. Then the overall importance of accident events to risk can be estimated by weighting the events' individual bin importance measures by the risk associated with each accident bin. 4 refs., 3 figs., 6 tabs

  20. Upon the reconstruction of accidents triggered by tire explosion. Analytical model and case study

    Science.gov (United States)

    Gaiginschi, L.; Agape, I.; Talif, S.

    2017-10-01

    Accident Reconstruction is important in the general context of increasing road traffic safety. In the casuistry of traffic accidents, those caused by tire explosions are critical under the severity of consequences, because they are usually happening at high speeds. Consequently, the knowledge of the running speed of the vehicle involved at the time of the tire explosion is essential to elucidate the circumstances of the accident. The paper presents an analytical model for the kinematics of a vehicle which, after the explosion of one of its tires, begins to skid, overturns and rolls. The model consists of two concurent approaches built as applications of the momentum conservation and energy conservation principles, and allows determination of the initial speed of the vehicle involved, by running backwards the sequences of the road event. The authors also aimed to both validate the two distinct analytical approaches by calibrating the calculation algorithms on a case study

  1. Analysis of core damage frequency: Surry, Unit 1 internal events

    International Nuclear Information System (INIS)

    Bertucio, R.C.; Julius, J.A.; Cramond, W.R.

    1990-04-01

    This document contains the accident sequence analysis of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed and described here is an extensive of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments form numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.05-E-5 per year, with a 95% upper bound of 1.34E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency. 49 refs., 52 figs., 70 tabs

  2. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  3. 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)

  4. 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

  5. Assessment of the event at Rovno NPP owing to the unexpected opening and obstruction of the safety valve lock mechanism of the pressure compensator

    International Nuclear Information System (INIS)

    Alonso, C.

    1993-01-01

    The main objective of this analysis is to be able to identify the sequence of the accident and evaluate the real frequency observed in similar nuclear power plants, according to the experience registered in the data bases of the NPP accident information system (ISI-AES) as well as the quality assurance information system (ISKO). This work describes how the analysis of events in a VVER-440 reactor NPP was performed

  6. Addressing severe accidents in the CANDU 9 design

    International Nuclear Information System (INIS)

    Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

    1998-01-01

    CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

  7. ANS severe accident program overview ampersand planning document

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1995-09-01

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10 -6 /y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents

  8. Saint-Alban plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Saint-Alban plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  9. 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)

  10. 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)

  11. 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.

  12. 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.

  13. A review for identification of initiating events in event tree development process on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Riyadi, Eko H., E-mail: e.riyadi@bapeten.go.id [Center for Regulatory Assessment of Nuclear Installation and Materials, Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada 8 Jakarta 10120 (Indonesia)

    2014-09-30

    Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.

  14. Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

    International Nuclear Information System (INIS)

    2007-12-01

    , initiating events which may challenge fuel safety can, in general, be grouped into three basic categories: power excursion accident, power-cooling-mismatch accident and decrease of reactor coolant inventory. This publication has been aided by two important trends. First, the methods of accident analysis have been developed significantly in recent years for a better understanding of physical phenomena, computing capabilities and the integration of research results into code development and application. Second, extensive studies have been carried out to investigate the transient behaviour for postulated initiating events sequences in order to establish that the subsequent fuel conditions do not exceed allowable limits. More detailed information on available methods for analysis of fuel behaviour under accident conditions and provides practical guidance for use of the methods is provided in this publication. The publication is directed at analysts coordinating, performing or reviewing the analysis of fuel behaviour under accident conditions, both on the designer and utility as well as on the regulatory side

  15. Severe Accident Research Network (SARNET). Level 2 PSA work package: comparison of partners methods for uncertainties assessment

    International Nuclear Information System (INIS)

    Chaumont, B.; Haesendonck, M.; Vidal, S.; Eyink, J.; Loeffler, H.; Radu, G.; Kopustinskas, V.; Ming, A.; Guntay, S.; Gustavsson, V.; Ivanov, I.; Dienstbier, J.; Bareith, A.; Hollo, E.; Lajtha, G.

    2007-01-01

    The PSA2 work package (PSA2 WP) is a part of the Joined Programme Activity of the European Severe Accident Network (SARNET) related to level 2 PSA methodologies. The general objectives of this work package is to provide a comparison of the different methodologies used or under development for level 2 PSA application by the partners involved in the work package and to promote their harmonization. The PSA2 WP is organized into three main topics: methodologies in general, methodologies for uncertainties assessment, and dynamic reliability methods. The different tasks initially defined for these three topics are shortly described and the partners involved identified. Attention is then paid on the methodologies used so far by the different partners to assess the uncertainties in their level 2 PSA. A review of partners approaches to assess - as far as possible - the different sources of possible uncertainties is done for the different following topics: - uncertainties propagated from the level 1 PSA, - uncertainties (in sense of approximation) due to the binning of the level 1 sequences in Plant Damage, - uncertainties related to the structure of the Accident Progression Event Tree, - uncertainties related to the probabilities of stochastic events (system failure or recovery, human actions, some physical phenomena such as ignition of hydrogen combustion or triggering of steam explosion), - uncertainties elated to the modelling of the different physical phenomena, - uncertainties related to the cut-off frequency used in the probabilistic quantification of the Accident Progression Event Tree; - uncertainties related to the binning of level 2 sequences in Release Categories (variables not considered, values of eventual continuous variables). First conclusions of the comparison are given in terms of improvement needs and then of perspectives of the work for the following period of work. (authors)

  16. Mean importance measures for groups of events in fault trees

    International Nuclear Information System (INIS)

    Haskin, F.E.; Huang, Min

    1994-01-01

    The method of moments is applied to precisely determine the mean values of three importance measures: risk reduction, partial derivative, and variance reduction. Variance reduction calculations, in particular, are significantly improved by eliminating the imprecision associated with Monte Carlo estimates. The three importance measures are extended to permit analyses of the relative importance of groups of basic and initiating events. The partial derivative importance measure is extended by assessing the contribution of a group of events to the gradient of the top event frequency. The group importance measures are quantified for the overall fuel damage equation and for 14 dominant accident sequences from an independent probabilistic safety assessment of the K Production Reactor. This application demonstrates both the utility and the versatility of the group importance measures

  17. Mean importance measures for groups of events in fault trees

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E.; Huang, Min [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Sasser, M.K.; Stack, D.W. [Los Alamos National Lab., NM (United States)

    1993-10-12

    The method of moments is applied to precisely determine the mean values of three importance measures: risk reduction, partial derivative, and variance reduction. Variance reduction calculations, in particular, are significantly improved by eliminating the imprecision associated with Monte Carlo estimates. The three importance measures are extended to permit analyses of the relative importance of groups of basic and initiating events. The partial derivative importance measure is extended by assessing the contribution of a group of events to the gradient of the top event frequency. The group importance measures are quantified for the overall fuel damage equation and for 14 dominant accident sequences from an independent probabilistic safety assessment of the K Production Reactor. This application demonstrates both the utility and the versatility of the group importance measures.

  18. Mean importance measures for groups of events in fault trees

    International Nuclear Information System (INIS)

    Haskin, F.E.; Huang, Min

    1993-01-01

    The method of moments is applied to precisely determine the mean values of three importance measures: risk reduction, partial derivative, and variance reduction. Variance reduction calculations, in particular, are significantly improved by eliminating the imprecision associated with Monte Carlo estimates. The three importance measures are extended to permit analyses of the relative importance of groups of basic and initiating events. The partial derivative importance measure is extended by assessing the contribution of a group of events to the gradient of the top event frequency. The group importance measures are quantified for the overall fuel damage equation and for 14 dominant accident sequences from an independent probabilistic safety assessment of the K Production Reactor. This application demonstrates both the utility and the versatility of the group importance measures

  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. 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

  2. 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)

  3. Identification of NPP accidents using support vector classification

    Energy Technology Data Exchange (ETDEWEB)

    Back, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    In case of the accidents that happens in a nuclear power plants (NPPs), it is very important to identify its accidents for the operator. Therefore, in order to effectively manage the accidents, the initial short time trends of major parameters have to be observed and NPP accidents have to accurately be identified to provide its information to operators and technicians. In this regard, the objective of this study is to identify the accidents when the accidents happen in NPPs. In this study, we applied the support vector classification (SVC) model to classify the initiating events of critical accidents such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Input variables were used as the initial integral value of the signal measured in the reactor coolant system (RCS), steam generator, and containment vessel after reactor trip. The proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. In this study, the proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. We used an initial integral value of the simulated sensor signals to identify the NPP accidents. The training data was used to train the SVC model. And, the trained model was confirmed using the test data. As a result, it was known that it can accurately classify five events.

  4. External flooding event analysis in a PWR-W with MAAP5

    International Nuclear Information System (INIS)

    Fernandez-Cosials, Mikel Kevin; Jimenez, Gonzalo; Barreira, Pilar; Queral, Cesar

    2015-01-01

    Highlights: • External flooding preceded by a SCRAM is simulated with MAAP5.01. • Sensitivities include AFW-TDP, SLOCA and operator preventive actions. • SLOCA flow is the dominant factor in the sequences. • Vessel failure is avoidable with operator preventive actions. - Abstract: The Fukushima accident has drawn attention even more to the importance of external events and loss of energy supply on safety analysis. Since 2011, several Station Blackout (SBO) analyses have been done for all type of reactors. The most post-Fukushima studies analyze a pure and straight SBO transient, but the Fukushima accident was more complex than a standard SBO. At Fukushima accident, the SBO was a consequence of an external flooding from the tsunami and occurred 40 min after an emergency shutdown (SCRAM) caused by the earthquake. The first objective of this paper is to assume the consequences of an external flooding accident in a PWR site caused by a river flood, a dam break or a tsunami, where all the plant is damaged, not only the diesel generators. The second objective is to analyze possible actions to be performed in the time between the earthquake event (that causes a SCRAM) and the external flooding arrival, which could be applicable to accidents such as dam failures or river flooding in order to avoid more severe consequences, delay the core damage and improve the accident management. The results reveal how the actuation of the different systems and equipments affect the core damage time and how some actions could delay the core damage time enough to increase the possibility of AC power recovery

  5. Analysis of internal events for the Unit 1 of the Laguna Verde Nuclear Power Station. Appendixes; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Apendices

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

    This volume contains the appendices for the accident sequences analysis for those internally initiated events for Laguna Verde Unit 1, Nuclear Power Plant. The appendix A presents the comments raised by the Sandia National Laboratories technical staff as a result of the review of the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant. This review was performed during a joint Sandia/CNSNS multi-day meeting by the end 1992. Also included is a brief evaluation on the applicability of these comments to the present study. The appendix B presents the fault tree models printed for each of the systems included and.analyzed in the Internal Event Analysis for LVNPP. The appendice C presents the outputs of the TEMAC code, used for the cuantification of the dominant accident sequences as well as for the final core damage evaluation. (Author)

  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. Simulation of severe accident using March-3 computer code

    International Nuclear Information System (INIS)

    Fernandes, A.; Nakata, H.

    1991-01-01

    The severe accident sensitivity analysis utilizing the March-3 approximate modelization options has been performed. The reference results against which the present results have been compared were obtained from the best published results for the most representative accident sequences: TMLU, S sub(2)DC sub(r) and S sub(2)DCF sub(r) for the Zion-1 reactor. The results of the present sensitivity analysis revealed the presence of very crude modelizations, in the March-3 program, to represent the critical phenomenologies involved in the severe accident sequences considered, even though large uncertainties must still be taken into account due primarily to the scarcity of the integral benchmark data. (author)

  8. 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

  9. 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

  10. Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

    International Nuclear Information System (INIS)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee

    2016-01-01

    Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment

  11. Note on the stock market's reaction to the accident at Three Mile Island

    International Nuclear Information System (INIS)

    Spudeck, R.E.; Moyer, C.R.

    1989-01-01

    This note provides new information regarding the market reaction toward electric utility stocks that resulted both from the accident at Three Mile Island, and the events predating and postdating the accident. The results suggest that some of the market reaction heretofore ascribed to the accident resulted instead from regulatory activity occurring before the accident. We also provide results suggesting that regulatory activity by the Pennsylvania Public Utilities Commission in the wake of the accident served to offset a majority of the increased systematic risk resulting from the accident. Our results imply that previously reported lingering effects of the accident at Three Mile Island may be regulatory effects from events predating the accident

  12. Accident Analysis and Barrier Function (AEB) Method. Manual for Incident Analysis

    International Nuclear Information System (INIS)

    Svenson, Ola

    2000-02-01

    The Accident Analysis and Barrier Function (AEB) Method models an accident or incident as a series of interactions between human and technical systems. In the sequence of human and technical errors leading to an accident there is, in principle, a possibility to arrest the development between each two successive errors. This can be done by a barrier function which, for example, can stop an operator from making an error. A barrier function can be performed by one or several barrier function systems. To illustrate, a mechanical system, a computer system or another operator can all perform a given barrier function to stop an operator from making an error. The barrier function analysis consists of analysis of suggested improvements, the effectiveness of the improvements, the costs of implementation, probability of implementation, the cost of maintaining the barrier function, the probability that maintenance will be kept up to standards and the generalizability of the suggested improvement. The AEB method is similar to the US method called HPES, but differs from that method in different ways. To exemplify, the AEB method has more emphasis on technical errors than HPES. In contrast to HPES that describes a series of events, the AEB method models only errors. This gives a more focused analysis making it well suited for checking other HPES-type accident analyses. However, the AEB method is a generic and stand-alone method that has been applied in other fields than nuclear power, such as, in traffic accident analyses

  13. Accident Analysis and Barrier Function (AEB) Method. Manual for Incident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology

    2000-02-01

    The Accident Analysis and Barrier Function (AEB) Method models an accident or incident as a series of interactions between human and technical systems. In the sequence of human and technical errors leading to an accident there is, in principle, a possibility to arrest the development between each two successive errors. This can be done by a barrier function which, for example, can stop an operator from making an error. A barrier function can be performed by one or several barrier function systems. To illustrate, a mechanical system, a computer system or another operator can all perform a given barrier function to stop an operator from making an error. The barrier function analysis consists of analysis of suggested improvements, the effectiveness of the improvements, the costs of implementation, probability of implementation, the cost of maintaining the barrier function, the probability that maintenance will be kept up to standards and the generalizability of the suggested improvement. The AEB method is similar to the US method called HPES, but differs from that method in different ways. To exemplify, the AEB method has more emphasis on technical errors than HPES. In contrast to HPES that describes a series of events, the AEB method models only errors. This gives a more focused analysis making it well suited for checking other HPES-type accident analyses. However, the AEB method is a generic and stand-alone method that has been applied in other fields than nuclear power, such as, in traffic accident analyses.

  14. 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.

  15. A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

    International Nuclear Information System (INIS)

    Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

    2003-04-01

    As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations

  16. Analysis of unmitigated large break loss of coolant accidents using MELCOR code

    Science.gov (United States)

    Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

    2017-11-01

    In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

  17. 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.)

  18. Dampierre-en-Burly plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Dampierre-en-Burly plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  19. Belleville-sur-Loire plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Belleville-sur-Loire plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  20. Nogent-sur-Seine plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Nogent-sur-Seine plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  1. Analysis of core damage frequency, Surry, Unit 1 internal events appendices

    International Nuclear Information System (INIS)

    Bertucio, R.C.; Julius, J.A.; Cramond, W.R.

    1990-04-01

    This document contains the appendices for the accident sequence analyses of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed is an extensive reanalysis of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.0E-5 per year, with a 95% upper bound of 1.3E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency

  2. Cerebrovascular Accidents Associated with Sorafenib in Hepatocellular Carcinoma

    OpenAIRE

    Saif, Muhammad W.; Isufi, Iris; Peccerillo, Jennifer; Syrigos, Kostas N.

    2011-01-01

    Sorafenib is an oral angiogenetic multikinase inhibitor approved in the treatment of renal and hepatocellular carcinoma. Bleeding and venous thrombotic events have been described with angiogenetic agents but cerebrovascular accidents are rarely reported. We report two cases of patients with hepatocellular carcinoma who developed a cerebrovascular accident while on sorafenib. Neither patient had any risk factors for the cerebrovascular events apart from gender and age in the second patient. La...

  3. 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)

  4. Analysis of accidents at the LPR (Radiochemical Processes Laboratory)

    International Nuclear Information System (INIS)

    Kaufmann, F.; Boutet, L.I.

    1987-01-01

    Accidents are defined as not planned events that may result in the emission of significative quantities of radioactive materials to the environment. The pilot plant has been specifically designed to prevent this type of accidents but there still exists the possibility that one or more accidents can be produced during the plant life. In a first phase, the emission of radionuclides to the environment were evaluated for 13 credible accidents. In a second phase, by means of the calculation program SEDA, specially adapted to this purpose, the critical doses of critical group were calculated for each accident. Due to the small capacity of the pilot plant and the long cooling period of treated fuel, it is concluded that the radiological consequences for the external environment are of very small magnitude. In this way, without need of developing complex fault- or event-trees, it is shown that any of the accidents falls into the non acceptable zone of Farmer diagram. (Author)

  5. The management of severe accidents in modern pressure tube reactors

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Blahnik, C.; Snell, V.G.; Duffey, R.B.

    2007-01-01

    Advanced new reactor designs resist severe accidents through a balance between prevention and mitigation. This balance is achieved by designing to ensure that such accidents are very rare; and by limiting core damage progression and releases from the plant in the event of such rare accidents. These design objectives are supported by a suitable combination of probabilistic safety analysis, engineering judgment and experimental and analytical study. This paper describes the approach used for the Advanced CANDU Reactor TM -1000 (ACR-1000) design, which includes provisions to both prevent and mitigate severe accidents. The paper describes the use of PSA as a 'design assist' tool; the analysis of core damage progression pathways; the definition of the core damage states; the capability of the mitigating systems to stop and control severe accident events; and the severe accident management opportunities for consequence reduction. (author)

  6. 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.)

  7. Tchernobyl: a severe accident and its image

    International Nuclear Information System (INIS)

    Strazzulla, J.

    1996-01-01

    This paper gives a strong criticism about the false informations that were disseminated by the mass media immediately after the Tchernobyl accident. This accident is taken as an example to illustrate a common attitude in journalistic comments of geopolitical events. (J.S.). 1 photo

  8. 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)

  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. Analysis of system and of course of events

    International Nuclear Information System (INIS)

    Hoertner, H.; Kersting, E.J.; Puetter, B.M.

    1986-01-01

    The analysis of the system and of the course of events is used to determine the frequency of core melt-out accidents and to describe the safety-related boundary conditions of appropriate accidents. The lecture is concerned with the effect of system changes in the reference plant and the effect of triggering events not assessed in detail or not sufficiently assessed in detail in phase A of the German Risk Study on the frequency of core melt-out accidents, the minimum requirements for system functions for controlling triggering events, i.e. to prevent core melt-out accidents, the reliability data important for reliability investigations and frequency assessments. (orig./DG) [de

  11. The role of internal and external control for mitigating or preventing LMR accidents

    International Nuclear Information System (INIS)

    Waltar, A.E.; Padilla, A.; Seeman, S.E.

    1987-01-01

    For the safety assessment of LMFBRs, much effort has been devoted to the analyses of LOF accident sequences with an emphasis on initiating-phase (IP) energetics (LOF-d-TOP event). Important knowledge and experiences on the IP energetics have been accumulated through reactor studies and in-pile experiment analyses, typically for the CABRI experiments. The present paper summarises the current understanding of key phenomenology relevant to the IP energetics based on the CABRI experiment analyses and the validation study for the PAPAS-2S, SAS3D and SAS4A codes. (author)

  12. 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

  13. An analysis of the Three Mile Island accident

    International Nuclear Information System (INIS)

    Brooks, G.L.; Siddal, E.

    1980-09-01

    Starting with a systematic analysis of the chain of events that took place during the Three Mile Island accident, the authors assess the significance of the four distinct phases of the accident. Inferences that can be drawn with respect to the safety of CANDU reactors are discussed. A rational reaction to the accident is suggested, and several factors are shown not to have played an important part, contrary to public impressions. The authors point out that over-reaction to the accident could detract from public safety. The Canadian response to the accident is discussed. (auth)

  14. Accident analysis for nuclear power plants

    International Nuclear Information System (INIS)

    2002-01-01

    Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

  15. Impact of rainstorm and runoff modeling on predicted consequences of atmospheric releases from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Brown, W.D.; Wayland, J.R.

    1980-05-01

    A general temperate latitude cyclonic rainstorm model is presented which describes the effects of washout and runoff on consequences of atmospheric releases of radioactive material from potential nuclear reactor accidents. The model treats the temporal and spatial variability of precipitation processes. Predicted air and ground concentrations of radioactive material and resultant health consequences for the new model are compared to those of the original WASH-1400 model under invariant meteorological conditions and for realistic weather events using observed meteorological sequences. For a specific accident under a particular set of meteorological conditions, the new model can give significantly different results from those predicted by the WASH-1400 model, but the aggregate consequences produced for a large number of meteorological conditions are similar

  16. Perspectives on the economic risks of LWR accidents

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Burke, R.P.

    1986-01-01

    Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant

  17. Computerized accident management support system: development for severe accident management

    International Nuclear Information System (INIS)

    Garcia, V.; Saiz, J.; Gomez, C.

    1998-01-01

    The activities involved in the international Halden Reactor Project (HRP), sponsored by the OECD, include the development of a Computerized Accident Management Support System (CAMS). The system was initially designed for its operation under normal conditions, operational transients and non severe accidents. Its purpose is to detect the plant status, analyzing the future evolution of the sequence (initially using the APROS simulation code) and the possible recovery and mitigation actions in case of an accident occurs. In order to widen the scope of CAMS to severe accident management issues, the integration of the MAAP code in the system has been proposed, as the contribution of the Spanish Electrical Sector to the project (with the coordination of DTN). To include this new capacity in CAMS is necessary to modify the system structure, including two new modules (Diagnosis and Adjustment). These modules are being developed currently for Pressurized Water Reactors and Boiling Water REactors, by the engineering of UNION FENOSA and IBERDROLA companies (respectively). This motion presents the characteristics of the new structure of the CAMS, as well as the general characteristics of the modules, developed by these companies in the framework of the Halden Reactor Project. (Author)

  18. Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment.

  19. 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

  20. The significance of domino effect in chemical accidents

    OpenAIRE

    Hemmatian, Behrouz; Abdolhamidzadeh, B; Darbra Roman, Rosa Maria; Casal Fàbrega, Joaquim

    2014-01-01

    A historical survey was performed on 330 accidents involving domino effect, occurred in process/storage plants and in the transportation of hazardous materials; only accidents occurred after 1st-January-1961 have been considered. The main features – geographical location, type of accident, materials involved, origin and causes, consequences, domino sequences – were analyzed, with special consideration to the situation in the developing countries and compared to those from other previous surve...

  1. Saint-Laurent-des-Eaux plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Saint-Laurent-des-Eaux plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  2. Evaluation of major polluting accidents in China-Results and perspectives

    International Nuclear Information System (INIS)

    Hou Yu; Zhang Tianzhu

    2009-01-01

    Lessons learnt from accidents are essential sources for updating state-of-the-art requirements in pollution accident prevention. To improve this input in the People's Republic of China in a systematic way, a database for collecting and evaluating major pollution accidents is being established. This is being done in co-operation with Chinese Society for Environment Sciences and other national Institutions. At the time of writing over 80 major events from 2002-2006 have been collected. In this paper, a summary evaluation on the major polluting events in China from 2002 to 2006 is presented and some basic lessons drawn shown. There is no a systematic pollution accident notification system currently in China. The results from root cause analysis underline the importance of emergency measures, maintenance, human factor issues and the role of safety organization. Chronic pollution, especially water pollution and air pollution should be paid the same attention as the sudden pollution. It is important to keep in mind that collecting information from major accidents represents a small percentage of the actual number of events taking place.

  3. Analysis of local subassembly accident in KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Jeong, Kwan Seong; Hahn, Do Hee

    2000-10-01

    Subassembly Accidents (S-A) in the Liquid Metal Reactor (LMR) may cause extensive clad and fuel melting and are thus regarded as a potential whole core accident initiator. The possibility of S-A occurrence must be very low frequency by the design features, and reactor must have specific instrumentation to interrupt the S-A sequences by causing a reactor shutdown. The evaluation of the relevant initiators, the event sequences which follow them, and their detection are the essence of the safety issue. Particularly, the phenomena of flow blockage caused by foreign materials and/or the debris from the failed fuel pin have been researched world-widely. The foreign strategies for dealing with the S-A and the associated safety issues with experimental and theoretical R and D results are reviewed. This report aims at obtaining information to reasonably evaluate the thermal-hydraulic effect of S-A for a wire-wrapped LMR fuel pin bundle. The mechanism of blockage formation and growth within a pin bundle and at the subassembly entrance is reviewed in the phenomenological aspect. Knowledge about the recent LMR subassembly design and operation procedure to prevent flow blockage will be reflected for KALIMER design later. The blockage analysis method including computer codes and related analytical models are reviewed. Especially SABRE4 code is discussed in detail. Preliminary analyses of flow blockage within a 271-pin driver subassembly have been performed using the SABRE4 computer code. As a result no sodium boiling occurred for the central 24-subchannel blockage as well as 6-subchannel blockage.

  4. Safety and Health Standard 110: Incident/accident reporting and investigation

    Energy Technology Data Exchange (ETDEWEB)

    Sones, K. [West Kootenay Power, BC (Canada)

    1999-10-01

    Incident/accident reporting requirements in effect at West Kootenay Power are discussed. Details provided include definitions of low risk, high risk, and critical events, the incidents to be reported, the nature of the reports, the timelines, the investigation to be undertaken for each type of incident/accident, counselling services available to employees involved in serious incidents, and the procedures to be followed in accidents involving serious injury to non-employees. The emphasis is on the `critical five` high risk events and the procedures relating to them.

  5. 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

  6. Integrated analyzing method for the progress event based on subjects and predicates in events

    International Nuclear Information System (INIS)

    Minowa, Hirotsugu; Munesawa, Yoshiomi

    2014-01-01

    It is expected to make use of the knowledge that was extracted by analyzing the mistakes of the past to prevent recurrence of accidents. Currently main analytic style is an analytic style that experts decipher deeply the accident cases, but cross-analysis has come to an end with extracting the common factors in the accident cases. We propose an integrated analyzing method for progress events to analyze among accidents in this study. Our method realized the integration of many accident cases by the integration connecting the common keyword called as 'Subject' or 'Predicate' that are extracted from each progress event in accident cases or near-miss cases. Our method can analyze and visualize the partial risk identification and the frequency to cause accidents and the risk assessment from the data integrated accident cases. The result of applying our method to PEC-SAFER accident cases identified 8 hazardous factors which can be caused from tank again, and visualized the high frequent factors that the first factor was damage of tank 26% and the second factor was the corrosion 21%, and visualized the high risks that the first risk was the damage 3.3 x 10 -2 [risk rank / year] and the second risk was the destroy 2.5 x 10 -2 [risk rank / year]. (author)

  7. ALWR severe accident issue resolution in support of updated emergency planning

    International Nuclear Information System (INIS)

    Additon, Stephen L.; Leaver, David E.; Sorrell, Steven W.; Theofanous, Theo G.

    2004-01-01

    . The severe accident risk characteristics of the ALWRs reflect an emphasis on accident prevention, which is quantified in the URD as a maximum permissible core damage frequency of less than one occurrence in 100,000 reactor years. For severe accident sequences of a frequency lower than this criterion, the URD safety policy requires provisions to arrest, mitigate, and contain the accident and, accordingly, opportunities to terminate a core melt sequence are provided whenever practical at every stage of core degradation. This includes design provisions to maximize the chances of success for reflooding the reactor by depressurizing the primary system, provisions to ensure retention of core debris in the reactor vessel by cooling the outside of the reactor vessel, and provisions for a more favorable geometry for core debris cooling in the reactor cavity in order to slow and then terminate a core-concrete interaction. For all risk-significant branches of the containment event tree, it must be demonstrated that early containment failure is avoided. This paper addresses the severe accident issue resolution tasks which were undertaken by the U.S. ALWR Program and ARSAP to ensure that the capability of passive ALWRs to arrest, mitigate and contain severe accidents would be sufficient to justify a significant change in the appropriate emergency planning requirements. The next section summarizes all of the issue resolution activities that will culminate in the issuance by the U.S. Nuclear Regulatory Commission (NRC) of a Final Safety Evaluation Report for the passive ALWR URD, scheduled for January 1994. The following section addresses more recent activities undertaken by ARSAP to enhance the issue resolution basis and to provide additional confirmatory evidence supporting the URD criteria. Included are the ongoing activities to establish a technical case, if possible, for in-vessel retention for the passive PWR and for the accommodation of ex-vessel steam explosions in the

  8. Accident at the Three Mile Island Nuclear Powerplant. Part 1. Oversight hearings before a task force of the Subcommittee on Energy and the Environment of the Committee on Interior and Insular Affairs, House of Representatives, Ninety-Sixth Congress

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Committee on Interior and Insular Affairs conducted an informal review of the accident beginning on March 28, 1979 at the Three Mile Island Nuclear Power Plant. Officials of the Nuclear Regulatory Commission, plant operating personnel employed by General Public Utilities, and representatives of the reactor manufacturer, Babcock and Wilcox Company, related their activities during the accident and their analyses of the sequence of events

  9. WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    Zhegang Ma

    2013-09-01

    The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

  10. Use of casual tree method for investigation of incidents and accidents involving radioactive materials

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Senne Junior, Murillo; Marques, Raissa Oliveira

    2013-01-01

    There are many methodologies used for investigation of accidents to facilitate the search of the factors that cause these events in different areas of industry. These can be called proactive methods, if they are used before the occurrence of the events, or reactive methods that are applied after the occurrence of the incident or accident, and are used as a basis of information to prevent further events. One of these methods is the Causal Tree Method (CTM). The basic idea of this technique is that incidents and accidents result from variations in usual processes. These variations can be related to the individual, the task, the material or the environment. The tree starts with the end event (incident or accident) and works backwards. The facts relating to the end event are used in the construction of the causal tree. The end event is the starting point and only the facts that contributed to the incident or accident should be selected. The analyst has to identify and list the variations and then display them in the analytic tree, showing causal relations. The objective of this paper is to test the application of the CTM method in investigation of incidents and accidents involving radioactive materials, in order to evaluate its efficiency on finding the typical factors causing these events. (author)

  11. Severe Accidents: French Regulatory Practice for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Colin, M.

    1997-01-01

    In the framework of a continuous and iterative process, the French Safety Authority asks the utility EDF to implement equipment and procedure modifications on the operating reactors, in order to cope with the most likely Severe Accident sequences. As a result of Probabilistic Safety Assessments published in 1990, important equipment and procedure modifications are being implemented on the French PWRs to improve the safety in shutdown states. The implementation of another set of modifications against some reactivity accident sequences is also in progress. More recently, the Safety Authority expressed specific Severe Accident requirements in terms of instrumentation, equipment qualification, high pressure core melt accidents and hydrogen risk prevention. In that respect, EDF was asked to implement hydrogen recombiners on its reactors. On the other hand, the French Safety authority is involved with its German counterpart in the assessment process of the European Pressurized Water Reactor Project. In consistency with the common recommendations of the Safety Authorities involved, Severe Accident provisions for this reactor are being taken into account at the design stage

  12. On-site emergency intervention plan for nuclear accident situation at INR-Pitesti TRIGA reactor

    International Nuclear Information System (INIS)

    Oprea, I.; Margenu, S.; Preda, M.

    2001-01-01

    A nuclear incident is defined as a series of events leading to release of radioactive materials into the environment of sufficient concentration to make necessary protective actions. The decision to initiate a protective action is a complex process. The benefits of taking the action is weighed against the involved risk and constraints. In addition the decision will be made under difficult emergency conditions, probably with little detailed information available. Therefore, considerable planing is necessary to reduce to manageable levels the types of decisions leading to effective responses to protect the public in the event of a nuclear incident. The sequence of events for developing emergency plans and responding to nuclear incidents will vary according to individual circumstances, because the international recommendations and site-specific emergency plans cannot provide detailed guidance for all accident scenarios and variations in local conditions. Flexibility must be maintained in emergency response to reflect the actual circumstances encountered (e.g. source term characteristics, the large number of possible weather conditions and environmental situation such as time of the day, season of the year, land use and soil types, population distribution and economic structures, uncertainties in the availability of technical and administrative support and the behaviour of the population). This further complicates the decision-making process, especially under accident conditions where there are time pressures and psychological stress. Therefore one the most important problems in the case of a nuclear emergency is quantifying all these very different types of off-site consequences. Last years, and in particular since the Chernobyl accident, there has been a considerable increase in the resources allocated to development of computerised systems which allow for predicting the radiological impact of accidents and to provide information in a manageable and effective form to

  13. Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

    2004-11-01

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

  14. Accident diagnosis system based on real-time decision tree expert system

    Science.gov (United States)

    Nicolau, Andressa dos S.; Augusto, João P. da S. C.; Schirru, Roberto

    2017-06-01

    Safety is one of the most studied topics when referring to power stations. For that reason, sensors and alarms develop an important role in environmental and human protection. When abnormal event happens, it triggers a chain of alarms that must be, somehow, checked by the control room operators. In this case, diagnosis support system can help operators to accurately identify the possible root-cause of the problem in short time. In this article, we present a computational model of a generic diagnose support system based on artificial intelligence, that was applied on the dataset of two real power stations: Angra1 Nuclear Power Plant and Santo Antônio Hydroelectric Plant. The proposed system processes all the information logged in the sequence of events before a shutdown signal using the expert's knowledge inputted into an expert system indicating the chain of events, from the shutdown signal to its root-cause. The results of both applications showed that the support system is a potential tool to help the control room operators identify abnormal events, as accidents and consequently increase the safety.

  15. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  16. 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

  17. Assessment of CRBR core disruptive accident energetics

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Bell, C.R.

    1984-03-01

    The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly

  18. 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.

  19. A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

    2003-04-01

    As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations.

  20. Return on experience on nuclear accidents

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-09-01

    After a presentation of the International Nuclear and radiological Events Scale (INES scale), of its levels and criteria, this article proposes brief recalls of some nuclear accidents which occurred in nuclear reactors: Chalk River in Canada (1952), Windscale in England (1957), the universal Canadian reactor (NRU in 1958), the SL1 reactor of the Idaho National Laboratory in the USA (1961), the Swiss Lucens reactor (1969), Saint-Laurent des Eaux in France (1969 and 1980). More detailed descriptions are then given for the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima accident in 2011. The main causes of these accidents are identified: loss of control of chain reaction, cooling defect on a stopped reactor, cooling defect on an operated reactor. Some lessons are drawn from these facts, and some characteristics of the EPR are outlined with respect with problems encountered in these accidents

  1. Computer Based Road Accident Reconstruction Experiences

    Directory of Open Access Journals (Sweden)

    Milan Batista

    2005-03-01

    Full Text Available Since road accident analyses and reconstructions are increasinglybased on specific computer software for simulationof vehicle d1iving dynamics and collision dynamics, and forsimulation of a set of trial runs from which the model that bestdescribes a real event can be selected, the paper presents anoverview of some computer software and methods available toaccident reconstruction experts. Besides being time-saving,when properly used such computer software can provide moreauthentic and more trustworthy accident reconstruction, thereforepractical experiences while using computer software toolsfor road accident reconstruction obtained in the TransportSafety Laboratory at the Faculty for Maritime Studies andTransport of the University of Ljubljana are presented and discussed.This paper addresses also software technology for extractingmaximum information from the accident photo-documentationto support accident reconstruction based on the simulationsoftware, as well as the field work of reconstruction expertsor police on the road accident scene defined by this technology.

  2. Statistical evaluation of design-error related accidents

    International Nuclear Information System (INIS)

    Ott, K.O.; Marchaterre, J.F.

    1980-01-01

    In a recently published paper (Campbell and Ott, 1979), a general methodology was proposed for the statistical evaluation of design-error related accidents. The evaluation aims at an estimate of the combined residual frequency of yet unknown types of accidents lurking in a certain technological system. Here, the original methodology is extended, as to apply to a variety of systems that evolves during the development of large-scale technologies. A special categorization of incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of the nuclear power reactor technology, considering serious accidents that involve in the accident-progression a particular design inadequacy

  3. Airborne concentrations of radioactive materials in severe accidents

    International Nuclear Information System (INIS)

    Ross, D.F. Jr.; Denning, R.S.

    1989-01-01

    Radioactive materials would be released to the containment building of a commercial nuclear reactor during each of the stages of a severe accident. Results of analyses of two accident sequences are used to illustrate the magnitudes of these sources of radioactive materials, the resulting airborne mass concentrations, the characteristics of the airborne aerosols, the potential for vapor forms of radioactive materials, the effectiveness of engineered safety features in reducing airborne concentrations, and the release of radioactive materials to the environment. Ability to predict transport and deposition of radioactive materials is important to assessing the performance of containment safety features in severe accidents and in the development of accident management procedures to reduce the consequences of severe accidents

  4. Learning from nuclear accident experience

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1984-01-01

    Statistical procedures are developed to estimate accident occurrence rates from historical event records, to predict future rates and trends, and to estimate the accuracy of the rate estimates and predictions. Maximum likelihood estimation is applied to several learning models, and results are compared to earlier graphical and analytical estimates. The models are based on (1) the cumulative number of operating years, (2) the cumulative number of plants built, and (3) accidents (explicitly), with the accident rate distinctly different before and after an accident. The statistical accuracies of the parameters estimated are obtained in analytical form using the Fisher information matrix. Using data on core damage accidents in electricity producing plants, it is estimated that the probability for a plant to have a serious flaw has decreased from 0.1 to 0.01 during the developmental phase of the nuclear industry. At the same time the equivalent frequency of accidents has decreased from 0.04 per reactor year to 0.0004 per reactor year, partly due to the increasing population of plants. 10 references, 7 figures, 2 tables

  5. 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

  6. Accident management-defence in depth in Indian PHWRS

    International Nuclear Information System (INIS)

    Jagannad, V.B.L.; Reddy, V.V.; Hajela, Sameer; Bhatia, C.M.; Nair, Suma

    2015-01-01

    Defence in Depth (DiD) is the established safety principle for the design of Nuclear Power Plants (NPPs). Accident at Fukushima Dai-ichi had highlighted the importance of provisions at Level-4 and 5 of DiD. Post Fukushima accident, on-site measures have been strengthened for Indian Nuclear Power Plants. On procedural front, Accident Management Guidelines have been introduced to handle events more severe than design basis accidents. This paper elaborates enhancement of Defence in Depth provisions for Indian Nuclear Power Plants. (author)

  7. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

    Waldman, R.M.; Vertullo, A.C.

    1987-01-01

    The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

  8. Accidents in industrial radiography in Brazil from 2005 to 2010

    International Nuclear Information System (INIS)

    Lopes, Ricardo Tadeu

    2011-01-01

    Analysis of accidents occurring in industrial radiography in Brazil from 2005 until 2010 led to the study of the main characteristics of the events, their risks and dangers. This study outlines the main doubts on the subject, through a simplified analysis of the contents of high dose reports sent to CNEN by the companies that provide services for industrial radiography and from examining the growing number of radioactive sources for industrial radiography in Brazil, over this period. We classified the recorded events, as incidents, accidents, negligence, sabotage, and others, and studied their main consequences. We concluded that from 76 accidents that occurred during that period - 25 were real accidents, 13 minor accidents and 22 were inadvertent incidents. We found that the rate of growth in the number of sources is much greater than the rate of growth of accidents, with a ratio of 7.57 between them. The continuation of this study over some years, will allow the construction of a pyramid of accidents like the one developed by the Insurance Company of North America, specifically for industrial radiography to forecast the number of incidents and accidents that lead to serious or fatal injury. (author)

  9. Accidental sequences associated with the containment of the pressurized water nuclear installation - INAP

    International Nuclear Information System (INIS)

    Natacci, Faustina Beatriz; Correa, Francisco

    2002-01-01

    The analysis of accidental sequences associated with the Containment is one of the most important tasks during the development of the Probabilistic Safety Assessment (PSA) of nuclear plants mainly because of its importance on the mitigation of consequences of severe postulated accident initiating events. This paper presents a first approach of the Containment analysis of the INAP identifying failures and events that can compromise its performance, and outlining accidental sequences and Containment end states. The initial plant damage states, which are the input for this study, are based on the event trees developed in the PSA level 1 for the INAP. It should be emphasized that since this PSA is still in a preliminary stage it is subjected to further completion. Consequently, the Containment analysis shall also be revised in order to incorporate, in an extension as complete as possible, all initial plant damage states, the corresponding event trees, and the related Containment end states. Finally, it can be concluded that the evaluation of the qualitative analysis presented herein allows a concise and broad knowledge of the qualitative analysis presented herein allows a concise and broad knowledge of the development of accidental sequences related to the Containment of the INAP. (author)

  10. Severe accident tests and development of domestic severe accident system codes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  11. Severe accident tests and development of domestic severe accident system codes

    International Nuclear Information System (INIS)

    2013-01-01

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  12. Internal event analysis of Laguna Verde Unit 1 Nuclear Power Plant. System Analysis; Analisis de Eventos Internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Analisis de sistemas

    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)

    1993-07-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the {sup I}nternal Event Analysis of Laguna Verde Unit 1 Nuclear Power Plant{sup ,} CNSNS-TR-004, in five volumes. The reports are organized as follows: CNSNS-TR-004 Volume 1: Introduction and Methodology. CNSNS-TR-004 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-004 Volume 5: Appendices A, B and C. This volume presents the results of the system analysis for the Laguna Verde Unit 1 Nuclear Power Plant. The system analysis involved the development of logical models for all the systems included in the accident sequence event tree headings, and for all the support systems required to operate the front line systems. For the Internal Event analysis for Laguna Verde, 16 front line systems and 5 support systems were included. Detailed fault trees were developed for most of the important systems. Simplified fault trees focusing on major faults were constructed for those systems that can be adequately represent,ed using this kind of modeling. For those systems where fault tree models were not constructed, actual data were used to represent the dominant failures of the systems. The main failures included in the fault trees are hardware failures, test and maintenance unavailabilities, common cause failures, and human errors. The SETS and TEMAC codes were used to perform the qualitative and quantitative fault tree analyses. (Author)

  13. Accident Diagnosis and Prognosis Aide (ADPA)

    International Nuclear Information System (INIS)

    Gunter, A.D.; Touchton, R.A.

    1987-01-01

    This presentation provides a demonstration of a prototypical expert system developed by Technology Applications, Inc. (TAI) under a contract with the Department of Energy as a part of their Small Business Innovation Research Program. The Accident Diagnosis and Prognosis Aide (ADPA) Demonstration Prototype is a working scale model of a real-time expert system which: Diagnoses an accident situation (as well as a number of underlying failures, events, and conditions deduced along the way). Calculates the change in the likelihood of core damage as a function of the events and failures diagnosed. Dynamically generates a recovery procedure tailored to the specific plant state at hand

  14. The Impact of Severe Nuclear Accidents on National Decision for Nuclear Decommissioning

    International Nuclear Information System (INIS)

    Suh, Young A; Hornibrook, Carol; Yim, Man Sung

    2016-01-01

    Many researchers have tried to identify the impact of severe nuclear accidents on a country's or international nuclear energy policy [2-3]. However, there is little research on the influence of nuclear accidents and historical events on a country's decision to permanently shutdown an NPP versus international nuclear decommissioning trends. To demonstrate the correlation between a nuclear severe accident and the impact on world nuclear decommissioning, this research reviewed case studies of individual historical events, such as the St. Lucens, TMI, Chernobyl, Fukushima accidents and the series of events leading up to the collapse of the Soviet Union. For validation of the results of these case studies, a statistical analysis was conducted using the R code. This will be useful in explaining how international and national decommissioning strategies are affected by shutdown reasons, i.e. world historical events. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently In conclusion, nuclear severe accidents and historical events have an impact on the number of international NPPs that shutdown permanently and cancelled NPP construction. This directly impacts international nuclear decommissioning policy and nuclear energy policy trends. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently

  15. Flamanville plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Flamanville plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 2 parts: one part dedicated to the first 2 reactors of the plant and the second part to the EPR that is being built. Each part is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  16. Blackout sequence modeling for Atucha-I with MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es

  17. Analysis of risk reduction methods for interfacing system LOCAs [loss-of-coolant accidents] at PWRs

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1988-01-01

    The Reactor Safety Study (WASH-1400) predicted that Interfacing System Loss-of-Coolant Accidents (ISL) events were significant contributors to risk even though they were calculated to be relatively low frequency events. However, there are substantial uncertainties involved in determining the probability and consequences of the ISL sequences. For example, the assumed valve failure modes, common cause contributions and the location of the break/leak are all uncertain and can significantly influence the predicted risk from ISL events. In order to provide more realistic estimates for the core damage frequencies (CDFs) and a reduction in the magnitude of the uncertainties, a reexamination of ISL scenarios at PWRs has been performed by Brookhaven National Laboratory. The objective of this study was to investigate the vulnerability of pressurized water reactor designs to ISLs and identify any improvements that could significantly reduce the frequency/risk of these events

  18. Emotional stress and traffic accidents: the impact of separation and divorce.

    Science.gov (United States)

    Lagarde, Emmanuel; Chastang, Jean-François; Gueguen, Alice; Coeuret-Pellicer, Mireille; Chiron, Mireille; Lafont, Sylviane

    2004-11-01

    Personal responses to stressful life events are suspected of increasing the risk of serious traffic accidents. We analyzed data from a French cohort study (the GAZEL cohort), including a retrospective driving behavior questionnaire, from 13,915 participants (10,542 men age 52-62 years and 3373 women age 47-62 years in 2001). Follow-up data covered 1993-2000. Hazard ratios for serious accidents (n = 713) were computed by Cox's proportional hazard regression with time-dependent covariates. Separate analyses were also performed to consider only at-fault accidents. Marital separation or divorce was associated with an increased risk of a serious accident (all serious accidents: hazard ratio 2.9, 95% confidence interval = 1.7-5.0; at-fault accidents: 4.4, 2.3-8.3). The impact of separation and divorce did not differ according to alcohol consumption levels. Other life events associated with increased risk of serious accident were a child leaving home (all accidents: 1.2, 0.97-1.6; at-fault accidents: 1.5, 1.1-2.1), an important purchase (all accidents: 1.4, 1.1-1.7; at-fault accidents: 1.6, 1.2-2.1), and hospitalization of the partner (all accidents: 1.4, 1.1-2.0). This study suggests that recent separation and divorce are associated with an increase in serious traffic accidents.

  19. Discrete dynamic event tree modeling and analysis of nuclear power plant crews for safety assessment

    International Nuclear Information System (INIS)

    Mercurio, D.

    2011-01-01

    Current Probabilistic Risk Assessment (PRA) and Human Reliability Analysis (HRA) methodologies model the evolution of accident sequences in Nuclear Power Plants (NPPs) mainly based on Logic Trees. The evolution of these sequences is a result of the interactions between the crew and plant; in current PRA methodologies, simplified models of these complex interactions are used. In this study, the Accident Dynamic Simulator (ADS), a modeling framework based on the Discrete Dynamic Event Tree (DDET), has been used for the simulation of crew-plant interactions during potential accident scenarios in NPPs. In addition, an operator/crew model has been developed to treat the response of the crew to the plant. The 'crew model' is made up of three operators whose behavior is guided by a set of rules-of-behavior (which represents the knowledge and training of the operators) coupled with written and mental procedures. In addition, an approach for addressing the crew timing variability in DDETs has been developed and implemented based on a set of HRA data from a simulator study. Finally, grouping techniques were developed and applied to the analysis of the scenarios generated by the crew-plant simulation. These techniques support the post-simulation analysis by grouping similar accident sequences, identifying the key contributing events, and quantifying the conditional probability of the groups. These techniques are used to characterize the context of the crew actions in order to obtain insights for HRA. The model has been applied for the analysis of a Small Loss Of Coolant Accident (SLOCA) event for a Pressurized Water Reactor (PWR). The simulation results support an improved characterization of the performance conditions or context of operator actions, which can be used in an HRA, in the analysis of the reliability of the actions. By providing information on the evolution of system indications, dynamic of cues, crew timing in performing procedure steps, situation

  20. 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)