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Sample records for advanced accident sequence

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  2. 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

  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. 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

  5. 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...

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  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. 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)

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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)

  2. 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

  3. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  4. 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.

  5. 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.

  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 Sequence Precursor Analysis for SGTR by Using Dynamic PSA Approach

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. 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

  9. The Integrated Approach to the Accident Evaluation for Advanced LWRs

    International Nuclear Information System (INIS)

    Oriolo, F.; Paci, S.

    1998-01-01

    The present paper discusses some relevant phenomena occurring in advanced LWRs during postulated accident scenarios. In particular, the operation of ESF is the starting point for analysis of those phenomena that cause the mutual influence between PS and containment in these plants. As a consequence, it is highlighted as accident analyses which treat PS and containment phenomena completely separated may be not adequate when applied to innovate reactors. Exemplified thermal-hydraulic analysis are presented for AP600 and SBWR, using the FUMO integrated model, for highlight accident evolution taking into account these interactions. The architecture of this integrated code is presented highlighting the importance of an integrated approach to the safety analysis in innovative reactors. (author)

  10. 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

  11. 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)

  12. Incorporation of advanced accident analysis methodology into safety analysis reports

    International Nuclear Information System (INIS)

    2003-05-01

    The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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.

  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. 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. 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

  4. 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.

  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. Consideration of severe accidents in design of advanced WWER reactors

    International Nuclear Information System (INIS)

    Fedorov, V.G.; Rogov, M.F.; Podshibyakin, A.K.; Fil, N.S.; Volkov, B.E.; Semishkin, V.P.

    1998-01-01

    Severe accident related requirements formulated in General Regulations for Nuclear Power Plant Safety (OPB-88), in Nuclear Safety Regulations for Nuclear Power Stations' Reactor Plants (PBYa RU AS-89) and in other NPP nuclear and radiation guides of the Russian Gosatomnadzor are analyzed. In accordance with these guides analyses of beyond design basis accidents should be performed in the reactor plant design. Categorization of beyond design basis accidents leading to severe accidents should be made on occurrence probability and severity of consequences. Engineered features and measures intended for severe accident management should be provided in reactor plant design. Requirements for severe accident analyses and for development of measures for severe accident management are determined. Design philosophy and proposed engineered measures for mitigation of severe accidents and decrease of radiation releases are demonstrated using examples of large, WWER-1000 (V-392), and medium size WWER-640 (V-407) reactor plant designs. Mitigation of severe accidents and decrease of radiation releases are supposed to be conducted on basis of consistent realization of the defense in depth concept relating to application of a system of barriers on the path of spreading of ionizing radiation and radioactive materials to the environment and a set of engineered measures protecting these barriers and retaining their effectiveness. Status of fulfilled by OKB Gidropress and other Russian organizations experimental and analytical investigations of severe accident phenomena supporting design decisions and severe accident management procedures is described. Status of the works on retention of core melt inside the WWER-640 reactor vessel is also characterized

  7. 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)

  8. Evaluating advancements in accident investigations using a novel framework

    NARCIS (Netherlands)

    Karanikas, N.; Soltani, P.; de Boer, R.J.; Roelen, A.

    2015-01-01

    Safety is monitored by various proactive and reactive methods, including the investigation of adverse accidents and incidents, which are collectively known as safety investigations. In this study we demonstrate how accident and incident investigation reports can be useful to identify implicit safety

  9. 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)

  10. 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

  11. Digital Recovery Sequencer - Advanced Concept Ejection Seats

    National Research Council Canada - National Science Library

    Ross, David A; Cotter, Lee; Culhane, David; Press, Matthew J

    2005-01-01

    .... Continued usage of the Analog Sequencer is undesirable due to limitations with respect to its installed life, electronic component obsolescence, flexibility to accommodate seat safety improvements...

  12. 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)

  13. 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

  14. Recent advances in severe accident technology - direct containment heating in advanced light water reactors

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1993-01-01

    The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

  15. 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)

  16. 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

  17. 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)

  18. 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

  19. 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

  20. A severe accident analysis for the system-integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Jae, Moosung

    2015-01-01

    The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

  1. Factors correlated with traffic accidents as a basis for evaluating Advanced Driver Assistance Systems.

    Science.gov (United States)

    Staubach, Maria

    2009-09-01

    This study aims to identify factors which influence and cause errors in traffic accidents and to use these as a basis for information to guide the application and design of driver assistance systems. A total of 474 accidents were examined in depth for this study by means of a psychological survey, data from accident reports, and technical reconstruction information. An error analysis was subsequently carried out, taking into account the driver, environment, and vehicle sub-systems. Results showed that all accidents were influenced by errors as a consequence of distraction and reduced activity. For crossroad accidents, there were further errors resulting from sight obstruction, masked stimuli, focus errors, and law infringements. Lane departure crashes were additionally caused by errors as a result of masked stimuli, law infringements, expectation errors as well as objective and action slips, while same direction accidents occurred additionally because of focus errors, expectation errors, and objective and action slips. Most accidents were influenced by multiple factors. There is a safety potential for Advanced Driver Assistance Systems (ADAS), which support the driver in information assimilation and help to avoid distraction and reduced activity. The design of the ADAS is dependent on the specific influencing factors of the accident type.

  2. 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.

  3. Advances in operational safety and severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. [VTT Automation (Finland)

    2002-02-01

    A project on reactor safety was carried out as a part of the NKS programme during 1999-2001. The objective of the project was to obtain a shared Nordic view of certain key safety issues related to the operating nuclear power plants in Finland and Sweden. The focus of the project was on selected central aspects of nuclear reactor safety that are of common interest for the Nordic nuclear authorities, utilities and research bodies. The project consisted of three sub-projects. One of them concentrated on the problems related to risk-informed deci- sion making, especially on the uncertainties and incompleteness of probabilistic safety assessments and their impact on the possibilities to use the PSA results in decision making. Another sub-project dealt with questions related to maintenance, such as human and organisational factors in maintenance and maintenance management. The focus of the third sub-project was on severe accidents. This sub-project concentrated on phenomenological studies of hydrogen combustion, formation of organic iodine, and core re-criticality due to molten core coolant interaction in the lower head of reactor vessel. Moreover, the current status of severe accident research and management was reviewed. (au)

  4. 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.

  5. Recent Advances In The Management Of Cerebro-Vascular Accident

    African Journals Online (AJOL)

    Benin Journal of Postgraduate Medicine. Journal Home · ABOUT THIS JOURNAL · Advanced Search · Current Issue · Archives · Journal Home > Vol 9, No 1 (2007) >. Log in or Register to get access to full text downloads.

  6. Advanced Approach to Consider Aleatory and Epistemic Uncertainties for Integral Accident Simulations

    International Nuclear Information System (INIS)

    Peschke, Joerg; Kloos, Martina

    2013-01-01

    The use of best-estimate codes together with realistic input data generally requires that all potentially important epistemic uncertainties which may affect the code prediction are considered in order to get an adequate quantification of the epistemic uncertainty of the prediction as an expression of the existing imprecise knowledge. To facilitate the performance of the required epistemic uncertainty analyses, methods and corresponding software tools are available like, for instance, the GRS-tool SUSA (Software for Uncertainty and Sensitivity Analysis). However, for risk-informed decision-making, the restriction on epistemic uncertainties alone is not enough. Transients and accident scenarios are also affected by aleatory uncertainties which are due to the unpredictable nature of phenomena. It is essential that aleatory uncertainties are taken into account as well, not only in a simplified and supposedly conservative way but as realistic as possible. The additional consideration of aleatory uncertainties, for instance, on the behavior of the technical system, the performance of plant operators, or on the behavior of the physical process provides a quantification of probabilistically significant accident sequences. Only if a safety analysis is able to account for both epistemic and aleatory uncertainties in a realistic manner, it can provide a well-founded risk-informed answer for decision-making. At GRS, an advanced probabilistic dynamics method was developed to address this problem and to provide a more realistic modeling and assessment of transients and accident scenarios. This method allows for an integral simulation of complex dynamic processes particularly taking into account interactions between the plant dynamics as simulated by a best-estimate code, the dynamics of operator actions and the influence of epistemic and aleatory uncertainties. In this paper, the GRS method MCDET (Monte Carlo Dynamic Event Tree) for probabilistic dynamics analysis is explained

  7. 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

  8. 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)

  9. 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

  10. 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

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

    International Nuclear Information System (INIS)

    Han, Seok Jung; Yang, Joon Eon

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  13. Considerations on monitoring needs of advanced, passive safety light water reactors for severe accident management

    International Nuclear Information System (INIS)

    Bava, G.; Zambardi, F.

    1992-01-01

    This paper deals with problems concerning information and related instrumentation needs for Accident Management (AM), with special emphasis on Severe Accidents (SA) in the new advanced, passive safety Light Water Reactors (PLWR), presently in a development stage. The passive safety conception adopted in the plants concerned goes parallel with a deeper consideration of SA, that reflects the need of increasing the plant resistance against conditions going beyond traditional ''design basis accidents''. Further, the role of Accident Management (AM) is still emphasized as last step of the defence in depth concept, in spite of the design efforts aimed to reduce human factor importance; as a consequence, the availability of pertinent information on actual plant conditions remains a necessary premise for performing preplanned actions. This information is essential to assess the evolution of the accident scenarios, to monitor the performances of the safety systems, to evaluate the ultimate challenge to the plant safety, and to implement the emergency operating procedures and the emergency plans. Based on these general purposes, the impact of the new conception on the monitoring structure is discussed, furthermore reference is made to the accident monitoring criteria applied in current plants to evaluate the requirements for possible solutions. (orig.)

  14. 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.

  15. 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)

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

    International Nuclear Information System (INIS)

    Shabon, M.H.

    2013-01-01

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

  17. 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

  18. 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)

  19. 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)

  20. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

  1. 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)

  2. 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

  3. The study of steam explosions in nuclear systems. Advanced Reactor Severe Accident Program

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Yuen, W.W.; Angelini, S.; Chen, X.

    1995-01-01

    This report presents an overview of the steam explosion issue in nuclear reactor safety and our approach to assessing it. Key physics, models, and computational tools are described, and illustrative results are presented for ex-vessel steam explosions in an open pool geometry. An extensive set of appendices facilitate access to previously reported work that is an integral part of this effort. These appendices include key developments in our approach, key advances in our understanding from physical and numerical experiments, and details of the most advanced computational results presented in this report. Of major significance are the following features: A consistent two-dimensional treatment for both premixing and propagation which in practical settings are ostensibly at least two-dimensional phenomena; experimental demonstration of voiding and microinteractions which represent key behaviors in premixing and propagation respectively; demonstration of the explosion venting phenomena in open pool geometries which, therefore, can be counted on as a very important mitigative feature; and introduction of the idea of penetration cutoff as a key mechanism prohibiting large-scale premixing in usual ex-vessel situations involving high pour velocities and subcooled pools. This report is intended as an overview and is to be followed by code manuals for PM-ALPHA and ESPROSE.m, respective verification reports, and application documents for reactor-specific applications. The applications will employ the Risk Oriented Accident Analysis Methodology (ROAAM) to address the safety importance of potential steam explosions phenomena in evaluated severe accidents for passive Advanced Light Water Reactors (ALWRs)

  4. The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

    International Nuclear Information System (INIS)

    Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

    2015-01-01

    Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

  5. Advanced surrogate model and sensitivity analysis methods for sodium fast reactor accident assessment

    International Nuclear Information System (INIS)

    Marrel, A.; Marie, N.; De Lozzo, M.

    2015-01-01

    Within the framework of the generation IV Sodium Fast Reactors, the safety in case of severe accidents is assessed. From this statement, CEA has developed a new physical tool to model the accident initiated by the Total Instantaneous Blockage (TIB) of a sub-assembly. This TIB simulator depends on many uncertain input parameters. This paper aims at proposing a global methodology combining several advanced statistical techniques in order to perform a global sensitivity analysis of this TIB simulator. The objective is to identify the most influential uncertain inputs for the various TIB outputs involved in the safety analysis. The proposed statistical methodology combining several advanced statistical techniques enables to take into account the constraints on the TIB simulator outputs (positivity constraints) and to deal simultaneously with various outputs. To do this, a space-filling design is used and the corresponding TIB model simulations are performed. Based on this learning sample, an efficient constrained Gaussian process metamodel is fitted on each TIB model outputs. Then, using the metamodels, classical sensitivity analyses are made for each TIB output. Multivariate global sensitivity analyses based on aggregated indices are also performed, providing additional valuable information. Main conclusions on the influence of each uncertain input are derived. - Highlights: • Physical-statistical tool for Sodium Fast Reactors TIB accident. • 27 uncertain parameters (core state, lack of physical knowledge) are highlighted. • Constrained Gaussian process efficiently predicts TIB outputs (safety criteria). • Multivariate sensitivity analyses reveal that three inputs are mainly influential. • The type of corium propagation (thermal or hydrodynamic) is the most influential

  6. 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)

  7. Design measures for prevention and mitigation of severe accidents at advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-06-01

    Over 8500 reactor-years of operating experience have been accumulated with the current nuclear energy systems. New generations of nuclear power plants are being developed, building upon this background of experience. During the last decade, requirements for equipment specifically intended to minimize releases of radioactive material to the environment in the event of a core melt accident have been introduced, and designs for new plants include measures for preventing and mitigating a range of severe accident scenarios. The IAEA Technical Committee Meeting on Impact of Severe Accidents on Plant Design and Layout of Advanced Water Cooled Reactors was jointly organized by the Department of Nuclear Energy and the Department of Nuclear Safety to review measures which are being incorporated into advanced water cooled reactor designs for preventing and mitigating severe accidents, the status of experimental and analytical investigations of severe accident phenomena and challenges which support design decisions and accident management procedures, and to understand the impact of explicitly addressing severe accidents on the cost of nuclear power plants. This publication is intended to provide an objective source of information on this topic. It includes 14 papers presented at the Technical Committee meeting held in Vienna between 21-25 October 1996. It also includes a Summary and Findings of the Working Groups. The papers were grouped in three sections. A separate abstract was prepared for each paper

  8. 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

  9. 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)

  10. Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

    2007-01-01

    A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

  11. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided

  12. Application of advanced statistical methods in assessment of the late phase of a nuclear accident

    International Nuclear Information System (INIS)

    Hofman, R.

    2008-01-01

    The paper presents a new methodology for improving of estimates of radiological situation on terrain in the late phase of a nuclear accident. Methods of Bayesian filtering are applied to the problem. The estimates are based on combination of modeled and measured data provided by responsible authorities. Exploiting information on uncertainty of both the data sources, we are able to produce improved estimate of the true situation on terrain. We also attempt to account for model error, which is unknown and plays crucial role in accuracy of the estimates. The main contribution of this paper is application of an approach based on advanced statistical methods, which allows for estimating of model error covariance structure upon measurements. Model error is estimated on basis of measured-minus-observed residuals evaluated upon measured and modeled values. The methodology is demonstrated on a sample scenario with simulated measurements. (authors)

  13. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios

  14. Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

    International Nuclear Information System (INIS)

    Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

    1995-01-01

    The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

  15. Application of advanced statistical methods in assessment of the late phase of a nuclear accident

    International Nuclear Information System (INIS)

    Hofman, R.

    2009-01-01

    The paper presents a new methodology for improving of estimates of radiological situation on terrain in the late phase of a nuclear accident. Methods of Bayesian filtering are applied to the problem. The estimates are based on combination of modeled and measured data provided by responsible authorities. Exploiting information on uncertainty of both the data sources, we are able to produce improved estimate of the true situation on terrain. We also attempt to account for model error, which is unknown and plays crucial role in accuracy of the estimates. The main contribution of this paper is application of an approach based on advanced statistical methods, which allows for estimating of model error covariance structure upon measurements. Model error is estimated on basis of measured-minus-observed residuals evaluated upon measured and modeled values. The methodology is demonstrated on a sample scenario with simulated measurements. (authors)

  16. 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.

  17. 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

  18. 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)

  19. 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.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. Safety Assessment of Advanced Imaging Sequences I: Measurements

    DEFF Research Database (Denmark)

    Jensen, Jørgen Arendt; Rasmussen, Morten Fischer; Pihl, Michael Johannes

    2016-01-01

    intensity measurement program. The approach can measure and store data for a full imaging sequence in 3.8 to 8.2 s per spatial position. Based on Ispta, MI, and probe surface temperature, the method gives the ability to determine whether a sequence is within US FDA limits, or alternatively indicate how......A method for rapid measurement of intensities (Ispta), mechanical index (MI), and probe surface temperature for any ultrasound scanning sequence is presented. It uses the scanner’s sampling capability to give an accurate measurement of the whole imaging sequence for all emissions to yield the true...... measurement system (Onda Corporation, Sunnyvale, CA, USA). Four different sequences have been measured: a fixed focus emission, a duplex sequence containing B-mode and flow emissions, a vector flow sequence with B-mode and flow emissions in 17 directions, and finally a synthetic aperture (SA) duplex flow...

  2. 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)

  3. 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)

  4. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht, H

    1986-01-01

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

  8. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.; Shin, S.T.

    1995-09-01

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of k eff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough

  9. Advancing analytical algorithms and pipelines for billions of microbial sequences.

    Science.gov (United States)

    Gonzalez, Antonio; Knight, Rob

    2012-02-01

    The vast number of microbial sequences resulting from sequencing efforts using new technologies require us to re-assess currently available analysis methodologies and tools. Here we describe trends in the development and distribution of software for analyzing microbial sequence data. We then focus on one widely used set of methods, dimensionality reduction techniques, which allow users to summarize and compare these vast datasets. We conclude by emphasizing the utility of formal software engineering methods for the development of computational biology tools, and the need for new algorithms for comparing microbial communities. Such large-scale comparisons will allow us to fulfill the dream of rapid integration and comparison of microbial sequence data sets, in a replicable analytical environment, in order to describe the microbial world we inhabit. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  12. Analysis of Sequence Diagram Layout in Advanced UML Modelling Tools

    Directory of Open Access Journals (Sweden)

    Ņikiforova Oksana

    2016-05-01

    Full Text Available System modelling using Unified Modelling Language (UML is the task that should be solved for software development. The more complex software becomes the higher requirements are stated to demonstrate the system to be developed, especially in its dynamic aspect, which in UML is offered by a sequence diagram. To solve this task, the main attention is devoted to the graphical presentation of the system, where diagram layout plays the central role in information perception. The UML sequence diagram due to its specific structure is selected for a deeper analysis on the elements’ layout. The authors research represents the abilities of modern UML modelling tools to offer automatic layout of the UML sequence diagram and analyse them according to criteria required for the diagram perception.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. Advances in safety countermeasures at the Tomari NPP of Hokkaido Electric Power on the basis of Fukushima Daiichi NPP accident. Fire protection and other advances

    International Nuclear Information System (INIS)

    Shibata, Taku; Dasai, Katsumi

    2014-01-01

    Fire protections for the nuclear power plants have been based on the fire laws and the conventional guide. After Fukushima Daiichi NPP accident, many safety countermeasures - also about Fire Protection - have been discussed in the Japanese authorities. This paper shows our present activities in the Tomari NPP about the fire protections from the view points of Fire Prevention, Fire Detection/Suppression Systems and Fire Protection, and other advances. (author)

  15. 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)

  16. Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

    International Nuclear Information System (INIS)

    Park, Jeong-Yong; KIM, Hyun-Gil; JUNG, Yang-Il; PARK, Dong-Jun; KOO, Yang-Hyun

    2013-01-01

    The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al 3 Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a

  17. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Parry-Romberg syndrome: findings in advanced magnetic resonance imaging sequences - case report

    Energy Technology Data Exchange (ETDEWEB)

    Paula, Rafael Alfenas de; Ribeiro, Bruno Niemeyer de Freitas, E-mail: alfenas85@gmail.com [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Hospital Universitario Clementino Fraga Filho; Bahia, Paulo Roberto Valle [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de radiologia; Ribeiro, Renato Niemeyer de Freitas [Hospital de Clinica de Jacarepagua, Rio de Janeiro, RJ (Brazil); Carvalho, Lais Balbi de [Universidade Presidente Antonio Carlos (Unipac), Juiz de Fora, MG (Brazil)

    2014-05-15

    Parry-Romberg syndrome is a rare disease characterized by progressive hemifacial atrophy associated with other systemic changes, including neurological symptoms. Currently, there are few studies exploring the utilization of advanced magnetic resonance sequences in the investigation of this disease. The authors report the case of a 45-year-old patient and describe the findings at structural magnetic resonance imaging and at advanced sequences, correlating them with pathophysiological data. (author)

  2. 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)

  3. 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)

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

    International Nuclear Information System (INIS)

    Niwa, Yuji; Takahashi, Hideaki.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  6. Advances of Single-Cell Sequencing Technique in Tumors

    Directory of Open Access Journals (Sweden)

    Ji-feng FENG

    2017-03-01

    Full Text Available With the completion of human genome project (HGP and the international HapMap project as well as rapid development of high-throughput biochip technology, whole genomic sequencing-targeted analysis of genomic structures has been primarily finished. Application of single cell for the analysis of the whole genomics is not only economical in material collection, but more importantly, the cell will be more purified, and the laboratory results will be more accurate and reliable. Therefore, exploration and analysis of hereditary information of single tumor cells has become the dream of all researchers in the field of basic research of tumors. At present, single-cell sequencing (SCS on malignancies has been widely used in the studies of pathogeneses of multiple malignancies, such as glioma, renal cancer and hematologic neoplasms, and in the studies of the metastatic mechanism of breast cancer by some researchers. This study mainly reviewed the SCS, the mechanisms and the methods of SCS in isolating tumor cells, and application of SCS technique in tumor-related basic research and clinical treatment.

  7. Safety Assessment of Advanced Imaging Sequences II: Simulations

    DEFF Research Database (Denmark)

    Jensen, Jørgen Arendt

    2016-01-01

    .6%, when using the impulse response of the probe estimated from an independent measurement. The accuracy is increased to between -22% to 24.5% for MI and between -33.2% to 27.0% for Ispta.3, when using the pressure response measured at a single point to scale the simulation. The spatial distribution of MI...... Mechanical Index (MI) and Ispta.3 as required by FDA. The method is performed on four different imaging schemes and compared to measurements conducted using the SARUS experimental scanner. The sequences include focused emissions with an F-number of 2 with 64 elements that generate highly non-linear fields....... The simulation time is between 0.67 ms to 2.8 ms per emission and imaging point, making it possible to simulate even complex emission sequences in less than 1 s for a single spatial position. The linear simulations yield a relative accuracy on MI between -12.1% to 52.3% and for Ispta.3 between -38.6% to 62...

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

    International Nuclear Information System (INIS)

    Muta, Hitoshi; Muramatsu, Ken

    2017-01-01

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

  9. 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

  10. Recent insights from severe accident research and implications for containment criteria for advanced LWRs

    International Nuclear Information System (INIS)

    Speis, T.P.; King, T.L.; Eltawila, F.

    1992-01-01

    The Severe Accident Research Program (SARP) was begun after the TMI-2 accident in March, 1979. The rule for dealing with the generation of large quantity of hydrogen in BWRs and Ice Condenser PWRs was promulgated by the Nuclear Regulatory Commission (NRC). The NRC issued severe Accident Policy Statement in 1985, and the revised SARP in 1989. In this paper, the current understanding of the more important phenomena and the associated mechanical and thermal loads to the containment is described, and the on-going works are summarized. The containment loadings in severe accidents are listed, and direct containment heating and the liner failure in BWR Mark I are added. A great deal of informations obtained on the early phase of melt progression are shown. The current understanding of the severe accident phenomena related to the containment and the on-going related research efforts are discussed more in detail. Fuel-coolant interaction including alpha-mode containment failure, direct containment heating, hydrogen deflagration and detonation, core-concrete interaction and debris coolability are described. (K.I.)

  11. Advanced evacuation model managed through fuzzy logic during an accident in LNG terminal

    Energy Technology Data Exchange (ETDEWEB)

    Stankovicj, Goran; Petelin, Stojan [Faculty for Maritime Studies and Transport, University of Ljubljana, Portorozh (Sierra Leone); others, and

    2014-07-01

    Evacuation of people located inside the enclosed area of an LNG terminal is a complex problem, especially considering that accidents involving LNG are potentially very hazardous. In order to create an evacuation model managed through fuzzy logic, extensive influence must be generated from safety analyses. A very important moment in the optimal functioning of an evacuation model is the creation of a database which incorporates all input indicators. The output result is the creation of a safety evacuation route which is active at the moment of the accident. (Author)

  12. Advance of Hazardous Operation Robot and its Application in Special Equipment Accident Rescue

    Science.gov (United States)

    Zeng, Qin-Da; Zhou, Wei; Zheng, Geng-Feng

    A survey of hazardous operation robot is given out in this article. Firstly, the latest researches such as nuclear industry robot, fire-fighting robot and explosive-handling robot are shown. Secondly, existing key technologies and their shortcomings are summarized, including moving mechanism, control system, perceptive technology and power technology. Thirdly, the trend of hazardous operation robot is predicted according to current situation. Finally, characteristics and hazards of special equipment accident, as well as feasibility of hazardous operation robot in the area of special equipment accident rescue are analyzed.

  13. 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)

  14. Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

    International Nuclear Information System (INIS)

    Hadid Subki, M.; )

    2014-01-01

    The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

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

    International Nuclear Information System (INIS)

    Naito, Masanori

    2011-01-01

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

  16. The influence of the technologically advanced evacuation models on the risk analyses during accidents in LNG terminal

    Energy Technology Data Exchange (ETDEWEB)

    Stankovicj, Goran; Petelin, Stojan [Faculty for Maritime Studies and Transport, University of Ljubljana, Portorozh (Sierra Leone); others, and

    2014-07-01

    The evacuation of people located in different safety zones of an LNG terminal is a complex problem considering that the accidents involving LNG are very hazardous and post the biggest threat to the safety of the people located near the LNG leakage. The safety risk criteria define the parameters which one LNG terminal should meet in terms of safety. Those criteria also contain an evacuation as an evasive action with the objective to mitigate the influence of the LNG accident on the people at risk. Till date, not a lot of attention has been paid to technologically advanced evacuations intended for LNG terminals. Creating the technologically advanced evacuation influences directly on the decrease of the probability of fatalities P{sub f,i}, thus influencing the calculation of the individual risk as well as the societal risk which results in the positioning of the F-N curve in the acceptable part of the ALARP zone. With this paper, we aim to present the difference between the safety analyses in cases when conservative data for P{sub f,i} is being used while calculating the risk, and in cases when real data for P{sub f,i} is been used. (Author)

  17. Recent advances in nanopore-based nucleic acid analysis and sequencing

    International Nuclear Information System (INIS)

    Shi, Jidong; Fang, Ying; Hou, Junfeng

    2016-01-01

    Nanopore-based sequencing platforms are transforming the field of genomic science. This review (containing 116 references) highlights some recent progress on nanopore-based nucleic acid analysis and sequencing. These studies are classified into three categories, biological, solid-state, and hybrid nanopores, according to their nanoporous materials. We begin with a brief description of the translocation-based detection mechanism of nanopores. Next, specific examples are given in nanopore-based nucleic acid analysis and sequencing, with an emphasis on identifying strategies that can improve the resolution of nanopores. This review concludes with a discussion of future research directions that will advance the practical applications of nanopore technology. (author)

  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. 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

  20. Joint CEC/OECD(NEA) workshop on recent advances in reactor accident consequence assessment

    International Nuclear Information System (INIS)

    Olast, M.; Sinnaeve, J.

    1988-01-01

    The workshop on probabilistic accident consequence assessment techniques and their applications aims at a review of the present knowledge of all the work in this field. This includes the atmospheric dispersion and deposition modelling, with comparison of the different approaches, the exposure pathways with emphasis on post-deposition processes, the health effects with emphasis on the consequences of the Hiroshima and Nagasaki data re-evaluation, the countermeasures and their economic consequences, the uncertainty analysis of the models and finally the applications of these models as aids to decision making

  1. A DNA sequence element that advances replication origin activation time in Saccharomyces cerevisiae.

    Science.gov (United States)

    Pohl, Thomas J; Kolor, Katherine; Fangman, Walton L; Brewer, Bonita J; Raghuraman, M K

    2013-11-06

    Eukaryotic origins of DNA replication undergo activation at various times in S-phase, allowing the genome to be duplicated in a temporally staggered fashion. In the budding yeast Saccharomyces cerevisiae, the activation times of individual origins are not intrinsic to those origins but are instead governed by surrounding sequences. Currently, there are two examples of DNA sequences that are known to advance origin activation time, centromeres and forkhead transcription factor binding sites. By combining deletion and linker scanning mutational analysis with two-dimensional gel electrophoresis to measure fork direction in the context of a two-origin plasmid, we have identified and characterized a 19- to 23-bp and a larger 584-bp DNA sequence that are capable of advancing origin activation time.

  2. Smart Sensing of the Aux. Feed-water Pump Performance in NPP Severe Accidents Using Advanced GMDH Method

    Energy Technology Data Exchange (ETDEWEB)

    No, Young Gyu; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In order to develop and verify the models, a number of data obtained by simulating station black out (SBO) scenario for the optimized power reactor 1000 (OPR1000) using MARS code were used. Most of monitoring systems for component have been suggested by using the directly measured data. However, it is very difficult to acquire data related to safety-critical component' status. Therefore, it is necessary to develop the new method that combines the data-based equipped with learning system and data miming technique. Many data-based modeling methods have been applied successfully to nuclear engineering area, such as signal validation, plant diagnostics and event identification. Also, the data miming is the process of analyzing data from different perspectives and summarizing it into useful information. In this study, the smart sensing technique was developed using advanced group method of data handing (GMDH) model. The original GMDH is an inductive self organizing algebraic model. The advanced GMDH model is equipped with a fuzzy concept. The proposed advanced GMDH model enhances the original GMDH model by reducing the effect of outliers and noise. The advanced GMDH uses different weightings according to their importance which is specified by the fuzzy membership grade. The developed model was verified using SBO accident simulation data for the OPR1000 nuclear power plant acquired with MARS code. Also, the advanced GMDH model was trained using the simulated development data and verified with simulated test data. The development and test data sets were independent. The simulation results show that the performance of the developed advanced GMDH model was very satisfactory, as shown in Table 1. Therefore, if the developed model can be optimized using diverse and specific data, it will be possible to predict the performance of Aux. feed water pump accurately.

  3. Protein Science by DNA Sequencing: How Advances in Molecular Biology Are Accelerating Biochemistry.

    Science.gov (United States)

    Higgins, Sean A; Savage, David F

    2018-01-09

    A fundamental goal of protein biochemistry is to determine the sequence-function relationship, but the vastness of sequence space makes comprehensive evaluation of this landscape difficult. However, advances in DNA synthesis and sequencing now allow researchers to assess the functional impact of every single mutation in many proteins, but challenges remain in library construction and the development of general assays applicable to a diverse range of protein functions. This Perspective briefly outlines the technical innovations in DNA manipulation that allow massively parallel protein biochemistry and then summarizes the methods currently available for library construction and the functional assays of protein variants. Areas in need of future innovation are highlighted with a particular focus on assay development and the use of computational analysis with machine learning to effectively traverse the sequence-function landscape. Finally, applications in the fundamentals of protein biochemistry, disease prediction, and protein engineering are presented.

  4. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

  5. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

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

    International Nuclear Information System (INIS)

    1986-11-01

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

  7. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    International Nuclear Information System (INIS)

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

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

    International Nuclear Information System (INIS)

    Bujor, A.

    1991-01-01

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

  9. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    Tong, L.L.; Huang, G.F.; Cao, X.W.

    2015-01-01

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  10. LWR severe accident simulation: Iodine behaviour in FPT2 experiment and advances on containment iodine chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Girault, N., E-mail: nathalie.girault@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Bosland, L. [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Dickinson, S. [National Nuclear Laboratory, Harwell, Oxon OX11 0QT (United Kingdom); Funke, F. [AREVA NP Gmbh, PO Box 1109, 91001 Erlangen (Germany); Guentay, S. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Herranz, L.E. [Centro des Investigaciones Energeticas, MedioAmbiantales y Tecnologicas, av. Complutense 2, 28040 Madrid (Spain); Powers, D. [Sandia National Laboratories, New Mexico, PO Box 5800, Albuquerque, NM 87185 (United States)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer Short term gaseous iodine fraction can be produced either in primary circuit or on containment condensing surfaces. Black-Right-Pointing-Pointer Gaseous radiolytic reactions convert volatile iodine into non-volatile iodine oxide particulates. Black-Right-Pointing-Pointer Alkaline and evaporating sump decrease the iodine volatility in containment. Black-Right-Pointing-Pointer Release of volatile iodine from containment surfaces explained the long term stationary residual gaseous iodine concentration. - Abstract: The Phebus Fission Product (FP) Program studies key phenomena of severe accidents in water-cooled nuclear reactors. In the framework of the Phebus program, five in-pile experiments have been performed that cover fuel rod degradation and behaviour of fission products released via the coolant circuit into the containment vessel. The focus of this paper is on iodine behaviour during the Phebus FPT2 test. FPT2 used a 33 GWd/t uranium dioxide fuel enriched to 4.5%, re-irradiated in situ for 7 days to a burn-up of 130 MWd/t. This test was performed to study the impact of steam-poor conditions and boric acid on the fission product chemistry. For the containment vessel, more specifically, the objective was to study iodine chemistry in an alkaline sump under evaporating conditions. The iodine results of the Phebus FPT2 test confirmed many of the essential features of iodine behaviour in the containment vessel provided by the first two Phebus tests, FPT0 and FPT1. These are the existence of an early gaseous iodine fraction, the persistence of low gaseous iodine concentrations and the importance of the sump in suppressing the iodine partitioning from sump to atmosphere. The main new insights provided by the Phebus FPT2 test were the iodine desorption from stainless steel walls deposits and the role of the evaporating sump in further iodine depletion in the containment atmosphere. The current paper presents an interpretation of

  11. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

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

  13. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  14. Next-generation sequencing for endocrine cancers: Recent advances and challenges.

    Science.gov (United States)

    Suresh, Padmanaban S; Venkatesh, Thejaswini; Tsutsumi, Rie; Shetty, Abhishek

    2017-05-01

    Contemporary molecular biology research tools have enriched numerous areas of biomedical research that address challenging diseases, including endocrine cancers (pituitary, thyroid, parathyroid, adrenal, testicular, ovarian, and neuroendocrine cancers). These tools have placed several intriguing clues before the scientific community. Endocrine cancers pose a major challenge in health care and research despite considerable attempts by researchers to understand their etiology. Microarray analyses have provided gene signatures from many cells, tissues, and organs that can differentiate healthy states from diseased ones, and even show patterns that correlate with stages of a disease. Microarray data can also elucidate the responses of endocrine tumors to therapeutic treatments. The rapid progress in next-generation sequencing methods has overcome many of the initial challenges of these technologies, and their advantages over microarray techniques have enabled them to emerge as valuable aids for clinical research applications (prognosis, identification of drug targets, etc.). A comprehensive review describing the recent advances in next-generation sequencing methods and their application in the evaluation of endocrine and endocrine-related cancers is lacking. The main purpose of this review is to illustrate the concepts that collectively constitute our current view of the possibilities offered by next-generation sequencing technological platforms, challenges to relevant applications, and perspectives on the future of clinical genetic testing of patients with endocrine tumors. We focus on recent discoveries in the use of next-generation sequencing methods for clinical diagnosis of endocrine tumors in patients and conclude with a discussion on persisting challenges and future objectives.

  15. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    International Nuclear Information System (INIS)

    Ruggles, A.E.; Cheng, L.Y.; Dimenna, R.A.; Griffith, P.; Wilson, G.E.

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis

  16. Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.; Xiang, J.Y.

    1995-01-01

    Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment

  17. Assessment of radiological impact due to a hypothetical core disruptive accident for PFBR using an advanced atmospheric dispersion system

    International Nuclear Information System (INIS)

    Srinivas, C.V.; Venkatesan, R.; Natarajan, A.

    2004-01-01

    Radiological impact due to air borne effluent dispersion from a hypothetical Core Disruptive Accident (CDA) scenario for Prototype Fast Breeder Reactor (PFBR) at Kalpakkam coastal site is estimated using an advanced system consisting of a 3-d meso-scale atmospheric model and a random walk particle dispersion model. A simulation of dispersion for CDA carried out for a typical summer day on 24th May 2003 predicted development of land-sea breeze circulation and Thermal Internal Boundary Layer (TIBL) at Kalpakkam site, which have been confirmed by observations. Analysis of dose distribution corresponding to predicted atmospheric conditions shows maximum dose from stack releases beyond the site boundary at about 4 km during TIBL fumigation and stable conditions respectively. A multi mode spatial concentration distribution has been noticed with diurnal meandering of wind under land sea breeze circulation. Over a meso-scale range of 25 km, turning of plume under sea breeze and maximum concentration along plume centerline at distances of 3 to 10 km have been noticed. The study has enabled to simulate the more complex meteorological situation that is actually present at the site. (author)

  18. Study on severe accident fuel dispersion behavior in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    Core flow blockage events have been determined to represent a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Hat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. The results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

  19. 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

  20. 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)

  1. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wirth, B. D. [Univ. of Tennessee, Knoxville, TN (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to

  2. Reactivity insertion accident analysis

    International Nuclear Information System (INIS)

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

    1990-04-01

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

  3. 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

  4. Advancing Eucalyptus genomics: identification and sequencing of lignin biosynthesis genes from deep-coverage BAC libraries

    Directory of Open Access Journals (Sweden)

    Kudrna David

    2011-03-01

    Full Text Available Abstract Background Eucalyptus species are among the most planted hardwoods in the world because of their rapid growth, adaptability and valuable wood properties. The development and integration of genomic resources into breeding practice will be increasingly important in the decades to come. Bacterial artificial chromosome (BAC libraries are key genomic tools that enable positional cloning of important traits, synteny evaluation, and the development of genome framework physical maps for genetic linkage and genome sequencing. Results We describe the construction and characterization of two deep-coverage BAC libraries EG_Ba and EG_Bb obtained from nuclear DNA fragments of E. grandis (clone BRASUZ1 digested with HindIII and BstYI, respectively. Genome coverages of 17 and 15 haploid genome equivalents were estimated for EG_Ba and EG_Bb, respectively. Both libraries contained large inserts, with average sizes ranging from 135 Kb (Eg_Bb to 157 Kb (Eg_Ba, very low extra-nuclear genome contamination providing a probability of finding a single copy gene ≥ 99.99%. Libraries were screened for the presence of several genes of interest via hybridizations to high-density BAC filters followed by PCR validation. Five selected BAC clones were sequenced and assembled using the Roche GS FLX technology providing the whole sequence of the E. grandis chloroplast genome, and complete genomic sequences of important lignin biosynthesis genes. Conclusions The two E. grandis BAC libraries described in this study represent an important milestone for the advancement of Eucalyptus genomics and forest tree research. These BAC resources have a highly redundant genome coverage (> 15×, contain large average inserts and have a very low percentage of clones with organellar DNA or empty vectors. These publicly available BAC libraries are thus suitable for a broad range of applications in genetic and genomic research in Eucalyptus and possibly in related species of Myrtaceae

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  7. Sequencing chemotherapy and radiotherapy in locoregional advanced breast cancer patients after mastectomy – a retrospective analysis

    International Nuclear Information System (INIS)

    Piroth, Marc D; Pinkawa, Michael; Gagel, Bernd; Stanzel, Sven; Asadpour, Branka; Eble, Michael J

    2008-01-01

    Combined chemo- and radiotherapy are established in breast cancer treatment. Chemotherapy is recommended prior to radiotherapy but decisive data on the optimal sequence are rare. This retrospective analysis aimed to assess the role of sequencing in patients after mastectomy because of advanced locoregional disease. A total of 212 eligible patients had a stage III breast cancer and had adjuvant chemotherapy and radiotherapy after mastectomy and axillary dissection between 1996 and 2004. According to concerted multi-modality treatment strategies 86 patients were treated sequentially (chemotherapy followed by radiotherapy) (SEQgroup), 70 patients had a sandwich treatment (SW-group) and 56 patients had simultaneous chemoradiation (SIM-group) during that time period. Radiotherapy comprised the thoracic wall and/or regional lymph nodes. The total dose was 45–50.4 Gray. As simultaneous chemoradiation CMF was given in 95.4% of patients while in sequential or sandwich application in 86% and 87.1% of patients an anthracycline-based chemotherapy was given. Concerning the parameters nodal involvement, lymphovascular invasion, extracapsular spread and extension of the irradiated region the three treatment groups were significantly imbalanced. The other parameters, e.g. age, pathological tumor stage, grading and receptor status were homogeneously distributed. Looking on those two groups with an equally effective chemotherapy (EC, FEC), the SEQ- and SW-group, the sole imbalance was the extension of LVI (57.1 vs. 25.6%, p < 0.0001). 5-year overall- and disease free survival were 53.2%/56%, 38.1%/32% and 64.2%/50%, for the sequential, sandwich and simultaneous regime, respectively, which differed significantly in the univariate analysis (p = 0.04 and p = 0.03, log-rank test). Also the 5-year locoregional or distant recurrence free survival showed no significant differences according to the sequence of chemo- and radiotherapy. In the multivariate analyses the sequence had no

  8. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

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

  9. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    International Nuclear Information System (INIS)

    2013-12-01

    Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

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

    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

  11. 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.

  12. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

    2012-05-15

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

  13. Next-Gen3: Sequencing, Modeling, and Advanced Biofuels - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Zengler, Karsten [Univ. of California, San Diego, CA (United States). Dept. of Pediatrics; Palsson, Bernhard [Univ. of California, San Diego, CA (United States). Dept. of Bioengineering; Lewis, Nathan [Univ. of California, San Diego, CA (United States). Dept. of Pediatrics

    2017-12-27

    Successful, scalable implementation of biofuels is dependent on the efficient and near complete utilization of diverse biomass sources. One approach is to utilize the large recalcitrant biomass fraction (or any organic waste stream) through the thermochemical conversion of organic compounds to syngas, a mixture of carbon monoxide (CO), carbon dioxide (CO2), and hydrogen (H2), which can subsequently be metabolized by acetogenic microorganisms to produce next-gen biofuels. The goal of this proposal was to advance the development of the acetogen Clostridium ljungdahlii as a chassis organism for next-gen biofuel production from cheap, renewable sources and to detail the interconnectivity of metabolism, energy conservation, and regulation of acetogens using next-gen sequencing and next-gen modeling. To achieve this goal we determined optimization of carbon and energy utilization through differential translational efficiency in C. ljungdahlii. Furthermore, we reconstructed a next-generation model of all major cellular processes, such as macromolecular synthesis and transcriptional regulation and deployed this model to predicting proteome allocation, overflow metabolism, and metal requirements in this model acetogen. In addition we explored the evolutionary significance of tRNA operon structure using the next-gen model and determined the optimal operon structure for bioproduction. Our study substantially enhanced the knowledgebaase for chemolithoautotrophs and their potential for advanced biofuel production. It provides next-generation modeling capability, offer innovative tools for genome-scale engineering, and provide novel methods to utilize next-generation models for the design of tunable systems that produce commodity chemicals from inexpensive sources.

  14. 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)

  15. Prokaryotic community profiling of local algae wastewaters using advanced 16S rRNA gene sequencing.

    Science.gov (United States)

    Limayem, Alya; Micciche, Andrew; Nayak, Bina; Mohapatra, Shyam

    2018-01-01

    Algae biomass-fed wastewaters are a promising source of lipid and bioenergy manufacture, revealing substantial end-product investment returns. However, wastewaters would contain lytic pathogens carrying drug resistance detrimental to algae yield and environmental safety. This study was conducted to simultaneously decipher through high-throughput advanced Illumina 16S ribosomal RNA (rRNA) gene sequencing, the cultivable and uncultivable bacterial community profile found in a single sample that was directly recovered from the local wastewater systems. Samples were collected from two previously documented sources including anaerobically digested (AD) municipal wastewater and swine wastewater with algae namely Chlorella spp. in addition to control samples, swine wastewater, and municipal wastewater without algae. Results indicated the presence of a significant level of Bacteria in all samples with an average of approximately 95.49% followed by Archaea 2.34%, in local wastewaters designed for algae cultivation. Taxonomic genus identification indicated the presence of Calothrix, Pseudomonas, and Clostridium as the most prevalent strains in both local municipal and swine wastewater samples containing algae with an average of 17.37, 12.19, and 7.84%, respectively. Interestingly, swine wastewater without algae displayed the lowest level of Pseudomonas strains algae indicates potential coexistence between these strains and algae microenvironment, suggesting further investigations. This finding was particularly relevant for the earlier documented adverse effects of some nosocomial Pseudomonas strains on algae growth and their multidrug resistance potential, requiring the development of targeted bioremediation with regard to the beneficial flora.

  16. Exome sequencing identifies early gastric carcinoma as an early stage of advanced gastric cancer.

    Directory of Open Access Journals (Sweden)

    Guhyun Kang

    Full Text Available Gastric carcinoma is one of the major causes of cancer-related mortality worldwide. Early detection and treatment leads to an excellent prognosis in patients with early gastric cancer (EGC, whereas the prognosis of patients with advanced gastric cancer (AGC remains poor. It is unclear whether EGCs and AGCs are distinct entities or whether EGCs are the beginning stages of AGCs. We performed whole exome sequencing of four samples from patients with EGC and compared the results with those from AGCs. In both EGCs and AGCs, a total of 268 genes were commonly mutated and independent mutations were additionally found in EGCs (516 genes and AGCs (3104 genes. A higher frequency of C>G transitions was observed in intestinal-type compared to diffuse-type carcinomas (P = 0.010. The DYRK3, GPR116, MCM10, PCDH17, PCDHB1, RDH5 and UNC5C genes are recurrently mutated in EGCs and may be involved in early carcinogenesis.

  17. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 1: Sections 1-9

    International Nuclear Information System (INIS)

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

    1995-04-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  18. 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.

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  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. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  3. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations. Volume 1: Sections 1-9

    International Nuclear Information System (INIS)

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

    1996-12-01

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

  4. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  5. The history and advances of reversible terminators used in new generations of sequencing technology.

    Science.gov (United States)

    Chen, Fei; Dong, Mengxing; Ge, Meng; Zhu, Lingxiang; Ren, Lufeng; Liu, Guocheng; Mu, Rong

    2013-02-01

    DNA sequencing using reversible terminators, as one sequencing by synthesis strategy, has garnered a great deal of interest due to its popular application in the second-generation high-throughput DNA sequencing technology. In this review, we provided its history of development, classification, and working mechanism of this technology. We also outlined the screening strategies for DNA polymerases to accommodate the reversible terminators as substrates during polymerization; particularly, we introduced the "REAP" method developed by us. At the end of this review, we discussed current limitations of this approach and provided potential solutions to extend its application. Copyright © 2013. Production and hosting by Elsevier Ltd.

  6. 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.

  7. Benchmarking severe accident computer codes for heavy water reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

    2010-07-01

    Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

  8. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

  9. 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

  10. 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

  11. 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.

  12. Advanced colorectal adenoma related gene expression signature may predict prognostic for colorectal cancer patients with adenoma-carcinoma sequence.

    Science.gov (United States)

    Li, Bing; Shi, Xiao-Yu; Liao, Dai-Xiang; Cao, Bang-Rong; Luo, Cheng-Hua; Cheng, Shu-Jun

    2015-01-01

    There are still no absolute parameters predicting progression of adenoma into cancer. The present study aimed to characterize functional differences on the multistep carcinogenetic process from the adenoma-carcinoma sequence. All samples were collected and mRNA expression profiling was performed by using Agilent Microarray high-throughput gene-chip technology. Then, the characteristics of mRNA expression profiles of adenoma-carcinoma sequence were described with bioinformatics software, and we analyzed the relationship between gene expression profiles of adenoma-adenocarcinoma sequence and clinical prognosis of colorectal cancer. The mRNA expressions of adenoma-carcinoma sequence were significantly different between high-grade intraepithelial neoplasia group and adenocarcinoma group. The biological process of gene ontology function enrichment analysis on differentially expressed genes between high-grade intraepithelial neoplasia group and adenocarcinoma group showed that genes enriched in the extracellular structure organization, skeletal system development, biological adhesion and itself regulated growth regulation, with the P value after FDR correction of less than 0.05. In addition, IPR-related protein mainly focused on the insulin-like growth factor binding proteins. The variable trends of gene expression profiles for adenoma-carcinoma sequence were mainly concentrated in high-grade intraepithelial neoplasia and adenocarcinoma. The differentially expressed genes are significantly correlated between high-grade intraepithelial neoplasia group and adenocarcinoma group. Bioinformatics analysis is an effective way to study the gene expression profiles in the adenoma-carcinoma sequence, and may provide an effective tool to involve colorectal cancer research strategy into colorectal adenoma or advanced adenoma.

  13. Accident management for severe accidents

    International Nuclear Information System (INIS)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

  14. Bacterial Pathogens and Community Composition in Advanced Sewage Treatment Systems Revealed by Metagenomics Analysis Based on High-Throughput Sequencing

    Science.gov (United States)

    Lu, Xin; Zhang, Xu-Xiang; Wang, Zhu; Huang, Kailong; Wang, Yuan; Liang, Weigang; Tan, Yunfei; Liu, Bo; Tang, Junying

    2015-01-01

    This study used 454 pyrosequencing, Illumina high-throughput sequencing and metagenomic analysis to investigate bacterial pathogens and their potential virulence in a sewage treatment plant (STP) applying both conventional and advanced treatment processes. Pyrosequencing and Illumina sequencing consistently demonstrated that Arcobacter genus occupied over 43.42% of total abundance of potential pathogens in the STP. At species level, potential pathogens Arcobacter butzleri, Aeromonas hydrophila and Klebsiella pneumonia dominated in raw sewage, which was also confirmed by quantitative real time PCR. Illumina sequencing also revealed prevalence of various types of pathogenicity islands and virulence proteins in the STP. Most of the potential pathogens and virulence factors were eliminated in the STP, and the removal efficiency mainly depended on oxidation ditch. Compared with sand filtration, magnetic resin seemed to have higher removals in most of the potential pathogens and virulence factors. However, presence of the residual A. butzleri in the final effluent still deserves more concerns. The findings indicate that sewage acts as an important source of environmental pathogens, but STPs can effectively control their spread in the environment. Joint use of the high-throughput sequencing technologies is considered a reliable method for deep and comprehensive overview of environmental bacterial virulence. PMID:25938416

  15. Ceramic Coatings for Clad (The C3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sickafus, Kurt E. [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Miller, Larry [Univ. of Tennessee, Knoxville, TN (United States); Weber, Bill [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States); Patel, Maulik [Univ. of Tennessee, Knoxville, TN (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Wolfe, Doug [Pennsylvania State Univ., University Park, PA (United States); Fratoni, Max [Univ. of California, Berkeley, CA (United States); Raj, Rishi [Univ. of Colorado, Boulder, CO (United States); Plunkett, Kenneth [Univ. of Colorado, Boulder, CO (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Hollis, Kendall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Chris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Comstock, Robert [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Partezana, Jonna [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Whittle, Karl [Univ. of Sheffield (United Kingdom); Preuss, Michael [Univ. of Manchester (United Kingdom); Withers, Philip [Univ. of Manchester (United Kingdom); Wilkinson, Angus [Univ. of Oxford (United Kingdom); Donnelly, Stephen [Univ. of Huddersfield (United Kingdom); Riley, Daniel [Australian Nuclear Science and Technology Organisation, Syndney (Australia)

    2017-02-14

    The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as

  16. Accident and emergency management

    International Nuclear Information System (INIS)

    Andersen, V.; Moellenbach, K.; Heinonen, R.; Jakobsson, S.; Kukko, T.; Berg, Oe.; Larsen, J.S.; Westgaard, T.; Magnusson, B.; Andersson, H.; Holmstroem, C.; Brehmer, B.; Allard, R.

    1988-06-01

    There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

  17. Intrusion detection system using Online Sequence Extreme Learning Machine (OS-ELM) in advanced metering infrastructure of smart grid.

    Science.gov (United States)

    Li, Yuancheng; Qiu, Rixuan; Jing, Sitong

    2018-01-01

    Advanced Metering Infrastructure (AMI) realizes a two-way communication of electricity data through by interconnecting with a computer network as the core component of the smart grid. Meanwhile, it brings many new security threats and the traditional intrusion detection method can't satisfy the security requirements of AMI. In this paper, an intrusion detection system based on Online Sequence Extreme Learning Machine (OS-ELM) is established, which is used to detecting the attack in AMI and carrying out the comparative analysis with other algorithms. Simulation results show that, compared with other intrusion detection methods, intrusion detection method based on OS-ELM is more superior in detection speed and accuracy.

  18. Advanced Whole-Genome Sequencing and Analysis of Fetal Genomes from Amniotic Fluid.

    Science.gov (United States)

    Mao, Qing; Chin, Robert; Xie, Weiwei; Deng, Yuqing; Zhang, Wenwei; Xu, Huixin; Zhang, Rebecca Yu; Shi, Quan; Peters, Erin E; Gulbahce, Natali; Li, Zhenyu; Chen, Fang; Drmanac, Radoje; Peters, Brock A

    2018-04-01

    Amniocentesis is a common procedure, the primary purpose of which is to collect cells from the fetus to allow testing for abnormal chromosomes, altered chromosomal copy number, or a small number of genes that have small single- to multibase defects. Here we demonstrate the feasibility of generating an accurate whole-genome sequence of a fetus from either the cellular or cell-free DNA (cfDNA) of an amniotic sample. cfDNA and DNA isolated from the cell pellet of 31 amniocenteses were sequenced to approximately 50× genome coverage by use of the Complete Genomics nanoarray platform. In a subset of the samples, long fragment read libraries were generated from DNA isolated from cells and sequenced to approximately 100× genome coverage. Concordance of variant calls between the 2 DNA sources and with parental libraries was >96%. Two fetal genomes were found to harbor potentially detrimental variants in chromodomain helicase DNA binding protein 8 ( CHD8 ) and LDL receptor-related protein 1 ( LRP1 ), variations of which have been associated with autism spectrum disorder and keratosis pilaris atrophicans, respectively. We also discovered drug sensitivities and carrier information of fetuses for a variety of diseases. We were able to elucidate the complete genome sequence of 31 fetuses from amniotic fluid and demonstrate that the cfDNA or DNA from the cell pellet can be analyzed with little difference in quality. We believe that current technologies could analyze this material in a highly accurate and complete manner and that analyses like these should be considered for addition to current amniocentesis procedures. © 2018 American Association for Clinical Chemistry.

  19. Investigations of the reflood-phase after a loss-of-coolant-accident of an advanced pressurized water reactor (APWR)

    International Nuclear Information System (INIS)

    Schumann, S.; Oldekop, W.

    1983-01-01

    Differences between a high converting advanced pressurized-water reactor (APWR) and a conventional PWR, which are relevant to the reflood-phase after LOCA are presented. The used code and its verification by PWR-reflood experiments is explained. Comparative calculations for APWR and PWR with several conservative assumptions for example cold-leg-injection only, yield nearly the same maximum midplane-temperatures for the average-channel. For the APWR, however, the upper half of the rod shows higher temperatures. Quenchfront and core-water-level increase more slowly. The differences in the reflood-thermohydraulics are analysed in detail. A conservative hot-channel calculation shows maximum temperatures of about 920 0 C. Finally the influence of conservative assumptions is described and the necessity of experiments pointed out. (orig.)

  20. Impact of Therapy Sequence with Alkylating Agents and MGMT Status in Patients with Advanced Neuroendocrine Tumors.

    Science.gov (United States)

    Krug, Sebastian; Boch, Michael; Rexin, Peter; Gress, Thomas M; Michl, Patrick; Rinke, Anja

    2017-05-01

    Alkylating chemotherapeutics with either a streptozotocin-(STZ) or temozolomide-(TEM) backbone are routinely used in patients with progressive and unresectable pancreatic neuroendocrine tumors (PNET). In addition, dacarbazine (DTIC) was described as an alternative alkylating therapy option for PNETs. The optimal treatment sequence with alkylating compounds and a potential use of O6-methylguanine-DNA methyltransferase (MGMT) level as predictive biomarker have not yet been sufficiently elucidated. The aim of our study was the evaluation of therapy sequence with either STZ-based treatment followed by DTIC (group A) or the inverse schedule with upfront DTIC (group B) and to correlate MGMT status with clinicopathological characteristics and response to therapy. We retrospectively analyzed 28 patients with neuroendocrine tumors (NET) who were treated with STZ-based therapy and DTIC. Additionally, in a second group MGMT immunohistochemistry was performed from primary and metastatic tumor sites. For statistical evaluation Kaplan-Meier analysis, Cox regression methods and Fisher's exact test were used. There was no difference of objective response and disease control between either STZ-based therapy followed by DTIC treatment (group A) after progression or the reverse sequence (group B). Median time to progression (TTP) was estimated to be 21 months in both arms. First-line STZ-based chemotherapy was not superior to first-line DTIC treatment (16 vs. 13 months; p=0.8). MGMT status did not correlate with clinicopathological characteristics or response to therapy with these alkylating agents. Upfront chemotherapy with either STZ-based treatment or DTIC monotherapy showed similar efficacy and median TTP rates. In this study, MGMT protein expression assessed by immunohistochemistry did not play an important role as a predictive marker for alkylating agents. Copyright© 2017, International Institute of Anticancer Research (Dr. George J. Delinasios), All rights reserved.

  1. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  2. Antibiotic abatement in different advanced oxidation processes coupled with a biological sequencing batch biofilm reactor

    International Nuclear Information System (INIS)

    Esplugas, M.; Gonzalez, O.; Benito, J.; Sans, C.

    2009-01-01

    During the last decade, the lack of fresh water is becoming a major concern. Recently, the present of recalcitrant products such as pharmaceuticals has caused a special interest due to their undefined environmental impact. Among these antibiotics are one of the numerous recalcitrant pollutants present in surface waters that might not be completely removed in the biological stage of sewage treatment plants because of their antibacterial nature. Advanced Oxidation Processes (AOPs) have proved to be highly efficient for the degradation of most organic pollutants in wastewaters. (Author)

  3. An advanced educational program for nuclear professionals with social scientific literacy. A collaborative initiative by UC Berkeley and Univ. of Tokyo on the Fukushima accident

    International Nuclear Information System (INIS)

    Juraku, Kohta; Nagasaki, Shinya; Ahn, Joonhong; Carson, Cathryn; Jensen, Mikael

    2011-01-01

    The authors have collaborated for over three years in developing an advanced educational program to cultivate leading engineers who can productively interact with other stakeholders. The program is organized under a partnership between the Nuclear Engineering Department of University of California, Berkeley (UCBNE) and the Global COE Program 'Nuclear Education and Research Initiative' (GoNERI) of the University of Tokyo, and is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology), Japan. We conducted two 'summer schools' in 2009 and 2010 as trial cases of the educational program. This year, in response to the Fukushima Daiichi nuclear accident, we decided to make our third summer school a venue for preliminary, yet multi-dimensional learning from that event. This school was held in Berkeley, CA, in the first week of August, with 12 lecturers and 18 students from various fields and countries. In this paper, we will explain the concept, aim, and design of our program; do a preliminary assessment of its effectiveness; introduce a couple of intriguing discussions held by participants; and discuss the program's implications for the post-Fukushima nuclear context. (author)

  4. TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

    International Nuclear Information System (INIS)

    Lime, J.F.; Boyack, B.E.

    1996-01-01

    An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

  5. Comparing treatment outcomes of different chemotherapy sequences during intensity modulated radiotherapy for advanced N-stage nasopharyngeal carcinoma patients

    International Nuclear Information System (INIS)

    Sun, Xueming; Zeng, Lei; Chen, Chunyan; Huang, Ying; Han, Fei; Xiao, Weiwei; Liu, Shuai; Lu, Taixiang

    2013-01-01

    N-stage is related to distant metastasis of nasopharyngeal carcinoma (NPC) patients. We performed this study to compare the efficacy of different chemotherapy sequences in advanced N-stage (N2 and N3) NPC patients treated with intensity modulated radiotherapy (IMRT). From 2001 to 2008, 198 advanced N-stage NPC patients were retrospectively analyzed. Thirty-three patients received IMRT alone. Concurrent chemoradiotherapy (CCRT) was delivered to 72 patients, neoadjuvant chemotherapy (NACT) + CCRT to 82 patients and CCRT + adjuvant chemotherapy (AC) to 11 patients. The 5-year overall survival rate, recurrence-free survival rate, distant metastasis-free survival rate and progress-free survival rate were 47.7% and 73.1%(p<0.001), 74.5% and 91.3% (p = 0.004), 49.2% and 68.5% (p = 0.018), 37.5% and 63.8% (p<0.001) in IMRT alone and chemoradiotherapy group. Subgroup analyses indicated that there were no significant differences among the survival curves of CCRT, NACT + CCRT and CCRT + AC groups. The survival benefit mainly came from CCRT. However, there was only an improvement attendency in distant metastasis-free survival rate of CCRT group (p = 0.107) when compared with RT alone group, and NACT + CCRT could significantly improve distant metastasis-free survival (p = 0.017). For advanced N-stage NPC patients, NACT + CCRT might be a reasonable treatment strategy

  6. Sequencing treatment of industrial wastewater with ultraviolet/H2O2 advanced oxidation and moving bed bioreactor

    Directory of Open Access Journals (Sweden)

    Mohammad Mehdi Mehrabani Ardekani

    2015-01-01

    Full Text Available Aims: The main purpose of this study was to determine the efficiency of a sequencing treatment including ultraviolet (UV/H 2 O 2 oxidation followed by a moving bed bioreactor (MBBR. Materials and Methods: Effect of solution pH, reaction time, and H 2 O 2 concentration were investigated for an industrial wastewater sample. The effluent of the advanced oxidation processes unit was introduced to the MBBR operated for three hydraulic retention times of 4, 8, and 12 h. Results: The optimum condition for industrial wastewater treatment via advanced oxidation was solution pH: 7, H 2 O 2 dose: 1000 mg/L and 90 min reaction time. These conditions led to 74.68% chemical oxygen demand (COD removal and 66.15% biochemical oxygen demand (BOD 5 removal from presedimentation step effluent that initially had COD and BOD 5 contents of 4,400 and 1,950 mg/L, respectively. Conclusion: Combination of UV/H 2 O 2 advanced oxidation with MBBR could result in effluents that meet water quality standards for discharge to receiving waters.

  7. Advanced handbook for accident analyses of German nuclear power plants; Weiterentwicklung eines Handbuches fuer Stoerfallanalysen deutscher Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Kerner, Alexander; Broecker, Annette; Hartung, Juergen; Mayer, Gerhard; Pallas Moner, Guim

    2014-09-15

    The advanced handbook of safety analyses (HSA) comprises a comprehensive electronic collection of knowledge for the compilation and conduction of safety analyses in the area of reactor, plant and containment behaviour as well as results of existing safety analyses (performed by GRS in the past) with characteristic specifications and further background information. In addition, know-how from the analysis software development and validation process is presented and relevant rules and regulations with regard to safety demonstration are provided. The HSA comprehensively covers the topic thermo-hydraulic safety analyses (except natural hazards, man-made hazards and malicious acts) for German pressurized and boiling water reactors for power and non-power operational states. In principle, the structure of the HSA-content represents the analytical approach utilized by safety analyses and applying the knowledge from safety analyses to technical support services. On the basis of a multilevel preparation of information to the topics ''compilation of safety analyses'', ''compilation of data bases'', ''assessment of safety analyses'', ''performed safety analyses'', ''rules and regulation'' and ''ATHLET-validation'' the HSA addresses users with different background, allowing them to enter the HSA at different levels. Moreover, the HSA serves as a reference book, which is designed future-oriented, freely configurable related to the content, completely integrated into the GRS internal portal and prepared to be used by a growing user group.

  8. Revisiting the mitogenomic phylogeny of Salmoninae: new insights thanks to recent sequencing advances

    Directory of Open Access Journals (Sweden)

    Jose L. Horreo

    2017-09-01

    Full Text Available The phylogeny of the Salmonidae family, the only living one of the Order Salmoniformes, remains still unclear because of several reasons. Such reasons include insufficient taxon sampling and/or DNA information. The use of complete mitochondrial genomes (mitogenomics could provide some light on it, but despite the high number of mitogenomes of species belonging to this family published during last years, an integrative work containing all this information has not been done. In this work, the phylogeny of 46 Salmonidae species was inferred from their mitogenomic sequences. Results include a Bayesian molecular-dated phylogenetic tree with very high statistical support showing Coregoninae and Salmoninae as sister subfamilies, as well as several new phylogenetic relationships among species and genus of the family. All these findings contribute to improve our understanding of the Salmonidae systematics and could have consequences on related evolutionary studies, as well as highlight the importance of revisiting phylogenies with integrative studies.

  9. Training simulator for advanced gas-cooled reactor (AGR) shutdown sequence equipment

    International Nuclear Information System (INIS)

    Shankland, J.P.; Nixon, G.L.

    1978-01-01

    Successful shutdown of nuclear plant is of prime importance for both safety and economic reasons and large sums of money are spent on equipment to make shutdowns fully automatic, thus removing the possibility of operator errors. While this aim can largely be realized, one must consider the possibility of automatic equipment or plant failures when operators are required to take manual action, and off-line training facilities should be available to operating staff to minimize the risk of incorrect actions being taken. This paper presents the practice adopted at Hunterston 'B' Nuclear Power Station to solve this problem and concerns the computer-based training simulator for the Reactor Shutdown Sequence Equipment (RSSE) which was commissioned in January 1977. The plant associated with shutdown is briefly described and the reasoning which shows the need for a simulator is outlined. The paper also gives details of the comprehensive facilities available on the simulator and goes on to describe the form that shutdown training takes and the experience gained at this time. (author)

  10. Proceedings of the first part of a joint OECD(NEA)/CEC workshop on recent advances in reactor accident consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-15

    The first part of the Joint Workshop, organised by the NEA, is focused on the progress achieved in the work of CSNI's GRECA (Group of Experts on Accident Consequences). The program is composed of the following papers. Session 1: characteristics of the Chernobyl release and fallout that affect transport and behaviour of radioactive substances in the environment; Chernobyl accident and hot particles in the fallout; radionuclides associated with colloids and particles in the Chernobyl fallout; source term in the Chernobyl accident; long range transport of radionuclides; parameters in consequence calculations for an urban area. Session 2: review of evaluations concerning radionuclide transfer to foodstuffs via plants in view of the data available after the Chernobyl accident; GRECA review of Chernobyl data on transfer to animal products; Chernobyl accident radiometric data (Cs-137 in fresh water fishes of north Italy lakes); distribution of Cs-137 in water sediment and fish in the Ijsselmeer (Netherlands); uptake in the human body resulting from the Chernobyl accident; radioactivity of people in the nordic countries following the Chernobyl accident; preparations for an international study to evaluate long-range transport models against the Chernobyl accident

  11. Intrusion detection system using Online Sequence Extreme Learning Machine (OS-ELM in advanced metering infrastructure of smart grid.

    Directory of Open Access Journals (Sweden)

    Yuancheng Li

    Full Text Available Advanced Metering Infrastructure (AMI realizes a two-way communication of electricity data through by interconnecting with a computer network as the core component of the smart grid. Meanwhile, it brings many new security threats and the traditional intrusion detection method can't satisfy the security requirements of AMI. In this paper, an intrusion detection system based on Online Sequence Extreme Learning Machine (OS-ELM is established, which is used to detecting the attack in AMI and carrying out the comparative analysis with other algorithms. Simulation results show that, compared with other intrusion detection methods, intrusion detection method based on OS-ELM is more superior in detection speed and accuracy.

  12. Road work zone accident studies : Advanced Research On Road Work Zone Safety Standard in Europe ARROWS Task 2.2 internal report. On behalf of the European Union, Directorate-General for Transport DG VII-E3, Transport RTD Programme.

    NARCIS (Netherlands)

    Gundy, C.M.

    1998-01-01

    The primary objective of this study, part of the ARROWS (Advanced Research on Road Work Zone Safety Standards in Europe) project, is to draw conclusions about the nature and extent of work zone traffic accidents. To that end, existing empirical studies concerning work zone traffic accidents have

  13. Normal accidents

    International Nuclear Information System (INIS)

    Perrow, C.

    1989-01-01

    The author has chosen numerous concrete examples to illustrate the hazardousness inherent in high-risk technologies. Starting with the TMI reactor accident in 1979, he shows that it is not only the nuclear energy sector that bears the risk of 'normal accidents', but also quite a number of other technologies and industrial sectors, or research fields. The author refers to the petrochemical industry, shipping, air traffic, large dams, mining activities, and genetic engineering, showing that due to the complexity of the systems and their manifold, rapidly interacting processes, accidents happen that cannot be thoroughly calculated, and hence are unavoidable. (orig./HP) [de

  14. Stress in accident and post-accident management at Chernobyl

    International Nuclear Information System (INIS)

    Girard, P.; Dubreuil, G.H.

    1996-01-01

    The effects of the Chernobyl nuclear accident on the psychology of the affected population have been much discussed. The psychological dimension has been advanced as a factor explaining the emergence, from 1990 onwards, of a post-accident crisis in the main CIS countries affected. This article presents the conclusions of a series of European studies, which focused on the consequences of the Chernobyl accident. These studies show that the psychological and social effects associated with the post-accident situation arise from the interdependency of a number of complex factors exerting a deleterious effect on the population. We shall first attempt to characterise the stress phenomena observed among the population affected by the accident. Secondly, we will be presenting an anlysis of the various factors that have contributed to the emerging psychological and social features of population reaction to the accident and in post-accident phases, while not neglecting the effects of the pre-accident situation on the target population. Thirdly, we shall devote some initial consideration to the conditions that might be conducive to better management of post-accident stress. In conclusion, we shall emphasise the need to restore confidence among the population generally. (Author)

  15. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  16. 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)

  17. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  18. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  19. De novo sequencing of circulating miRNAs identifies novel markers predicting clinical outcome of locally advanced breast cancer

    Directory of Open Access Journals (Sweden)

    Wu Xiwei

    2012-03-01

    Full Text Available Abstract Background MicroRNAs (miRNAs have been recently detected in the circulation of cancer patients, where they are associated with clinical parameters. Discovery profiling of circulating small RNAs has not been reported in breast cancer (BC, and was carried out in this study to identify blood-based small RNA markers of BC clinical outcome. Methods The pre-treatment sera of 42 stage II-III locally advanced and inflammatory BC patients who received neoadjuvant chemotherapy (NCT followed by surgical tumor resection were analyzed for marker identification by deep sequencing all circulating small RNAs. An independent validation cohort of 26 stage II-III BC patients was used to assess the power of identified miRNA markers. Results More than 800 miRNA species were detected in the circulation, and observed patterns showed association with histopathological profiles of BC. Groups of circulating miRNAs differentially associated with ER/PR/HER2 status and inflammatory BC were identified. The relative levels of selected miRNAs measured by PCR showed consistency with their abundance determined by deep sequencing. Two circulating miRNAs, miR-375 and miR-122, exhibited strong correlations with clinical outcomes, including NCT response and relapse with metastatic disease. In the validation cohort, higher levels of circulating miR-122 specifically predicted metastatic recurrence in stage II-III BC patients. Conclusions Our study indicates that certain miRNAs can serve as potential blood-based biomarkers for NCT response, and that miR-122 prevalence in the circulation predicts BC metastasis in early-stage patients. These results may allow optimized chemotherapy treatments and preventive anti-metastasis interventions in future clinical applications.

  20. 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

  1. 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

  2. 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)

  3. 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

  4. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

  5. Twenty-third water reactor safety information meeting: Volume 2, Human factors research; Advanced I and C hardware and software; Severe accident research; Probabilistic risk assessment topics; Individual plant examination: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 2, present topics in human factors research, advanced instrumentation and control hardware and software, severe accident research, probabilistic risk assessment, and individual plant examination. Individual papers have been cataloged separately.

  6. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Buchanan, J.R.; Lorenz, R.A.; Yamashita, T.

    1986-01-01

    On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

  7. 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

  8. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  9. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U3Si2 Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miao, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \\uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \\usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022~\\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\\\ \

  10. 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)

  11. Study of fission products (Cs, Ba, Mo, Ru) behaviour in irradiated and simulated nuclear fuels during severe accidents using X-ray absorption Spectroscopy, SIMS and EPMA

    International Nuclear Information System (INIS)

    Geiger, Ernesto

    2016-01-01

    The identification of Fission Products (FP) release mechanism from irradiated nuclear fuels during a severe accident is of main importance for the development of codes for the estimation of the source-term (nature and quantity of radionuclides released into the environment). among the many FP Ba, Cs, Mo and Ru present a particular interest, since they may interact with each other or other elements and thus affect their release. In the framework of this thesis, two work axes have been set up in order to identify, firstly, the chemical phases initially present before the accident and, secondly, their evolution during the accident itself. The experimental approach consisted in reproducing nuclear severe accidents conditions at laboratory scale using both irradiated fuels and model materials (natural UO_2 doped with 12 FP). The advantage of these latter is the possibility of using characterization methods such as X-ray absorption Spectroscopy which are not available for irradiated fuels. Three irradiated fuel samples have been studied, representative to an initial state (before the accident), to an intermediate stage (1773 K) and to an advanced stage (2873 K) of a nuclear severe accident. Regarding to model materials, many accident sequences have been carried out, from 573 to 1973 K. Experimental results have allowed to establish a new release mechanism, considering both reducing and oxidizing conditions during an accident. These results have also demonstrated the importance of model materials as a complement to irradiated nuclear fuels in the study of nuclear severe accidents. (author) [fr

  12. Prevention of pedestrian accidents.

    OpenAIRE

    Kendrick, D

    1993-01-01

    Child pedestrian accidents are the most common road traffic accident resulting in injury. Much of the existing work on road traffic accidents is based on analysing clusters of accidents despite evidence that child pedestrian accidents tend to be more dispersed than this. This paper analyses pedestrian accidents in 573 children aged 0-11 years by a locally derived deprivation score for the years 1988-90. The analysis shows a significantly higher accident rate in deprived areas and a dose respo...

  13. 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.

  14. Advanced colorectal adenoma related gene expression signature may predict prognostic for colorectal cancer patients with adenoma-carcinoma sequence

    OpenAIRE

    Li, Bing; Shi, Xiao-Yu; Liao, Dai-Xiang; Cao, Bang-Rong; Luo, Cheng-Hua; Cheng, Shu-Jun

    2015-01-01

    Background: There are still no absolute parameters predicting progression of adenoma into cancer. The present study aimed to characterize functional differences on the multistep carcinogenetic process from the adenoma-carcinoma sequence. Methods: All samples were collected and mRNA expression profiling was performed by using Agilent Microarray high-throughput gene-chip technology. Then, the characteristics of mRNA expression profiles of adenoma-carcinoma sequence were described with bioinform...

  15. Radiation accidents

    International Nuclear Information System (INIS)

    Saenger, E.L.

    1986-01-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity

  16. Chernobyl accident

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1995-01-01

    The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

  17. 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

  18. Reactor accidents and the environment

    International Nuclear Information System (INIS)

    Beattie, J.R.; Griffiths, R.F.; Kaiser, G.D.; Kinchin, G.H.

    1978-01-01

    This is a condensed version of a paper, entitled 'The Environmental Impact of Radioactive Releases from Accidents in Nuclear Power Reactors', by the authors, presented to the Nuclear Energy Panel of the International Atomic Energy Agency/United Nations Environmental Programme. Headings include - Effects of ionising radiation on man; number of deaths expected from leukaemia and other cancers; risk estimates for incidence of benign nodules and thyroid cancer; maximum permissible levels and emergency levels of radiation and radioactivity; ICRP recommended dose limits for members of the general public; atmospheric dispersion and modelling; ICRP emergency reference levels for 1 131 , Cs 137 , Ru 106 and Sr 90 ; environmental consequences of accidental releases from nuclear power reactors; environmental impact of accidents to Magnox gas-cooled reactors; environmental impact of accidents to advanced gas-cooled reactors; environmental impact of accidents to fast reactors; and nature of risks. consequences are examined in terms of early and late biological effects on man, and contamination of land areas. Serious accidents are of low probability of occurrence, and the risk of accidents to nuclear power reactors is estimated to be very small. 43 references. (U.K.)

  19. ATHLET validation using accident management experiments

    Energy Technology Data Exchange (ETDEWEB)

    Teschendorff, V.; Glaeser, H.; Steinhoff, F. [Gasellschaft fuer Anlagen - und Reaktorsicherheit (GSR) mbH, Garching (Germany)

    1995-09-01

    The computer code ATHLET is being developed as an advanced best-estimate code for the simulation of leaks and transients in PWRs and BWRs including beyond design basis accidents. The code has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialisation by a steady-state calculation, full-range drift-flux model, and dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The systematic validation of ATHLET is based on a well balanced set of integral and separate effect tests derived from the CSNI proposal emphasising, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities. PKL-III test B 2.1 simulates a cool-down procedure during an emergency power case with three steam generators isolated. Natural circulation under these conditions was investigated in detail in a pressure range of 4 to 2 MPa. The transient was calculated over 22000 s with complicated boundary conditions including manual control actions. The calculations demonstrations the capability to model the following processes successfully: (1) variation of the natural circulation caused by steam generator isolation, (2) vapour formation in the U-tubes of the isolated steam generators, (3) break-down of circulation in the loop containing the isolated steam generator following controlled cool-down of the secondary side, (4) accumulation of vapour in the pressure vessel dome. One conclusion with respect to the suitability of experiments simulating AM procedures for code validation purposes is that complete documentation of control actions during the experiment must be available. Special attention should be given to the documentation of operator actions in the course of the experiment.

  20. 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

  1. Chernobyl accident

    International Nuclear Information System (INIS)

    Capra, D.; Facchini, U.; Gianelle, V.; Ravasini, G.; Bacci, P.

    1988-01-01

    The radioactive cloud released during the Chernobyl accident reached the Padana plain and Lombardy in the night of April 30th 1986; the cloud remained in the northern Italian skies for a few days and then disappeared either dispersed by winds and washed by rains. The evidence in atmosphere of radionuclides as Tellurium, Iodine, Cesium, was promptly observed. The intense rain, in first week of may, washed the radioactivity and fall-out contamined the land, soil, grass. The present work concerns the overall contamination of the Northern Italy territory and in particular the radioactive fall-out in the Lakes region. Samples of soil have been measured at the gamma spectroscope; a correlation is found between the radionuclides concentration in soil samples and the rain intensity, when appropriate deposition models are considered. A number of measurements has been done on the Como'lake ecosystem: sediments, plankton, fishes and the overall fall-out in the area has been investigated

  2. Self-reported accidents

    DEFF Research Database (Denmark)

    Møller, Katrine Meltofte; Andersen, Camilla Sloth

    2016-01-01

    The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

  3. Accident Assessment

    International Nuclear Information System (INIS)

    Tripputi, Ivo; Lund, Ingemar

    2002-01-01

    There is a general feeling that decommissioning is an activity involving limited risks, compared to NPP operation, and in particular risks involving the general public. This is technically confirmed by licensing analysis and evaluations, where, once the spent fuel has been removed from the plant, the radioactivity inventory available to be released to the environment is very limited. Decommissioning activities performed so far in the world have also confirmed the first assumptions and no specific issue has been identified, in this field, to justify a completely new approach. Commercial interests in international harmonization, which could drive an in-depth discussion about the bases of this approach, are weak at the moment. However, there are several reasons why a discussion in an international framework about the Safety Case for decommissioning (and, in particular, about Accident Assessment) may be considered necessary and important, and why it may show some specific and peculiar aspects. An effort for a comprehensive and systematic D and D accident safety assessment of the decommissioning process is justified. It is necessary also to explore in a holistic way the aspects of industrial safety, and develop tools for the decision-making process optimization. The expected results are the implementation of appropriate and optimized protective measures in any event and of adequate on/off-site emergency plans for optimal public and workers protection. The experience from other decommissioning projects and large-scale industrial activities is essential to balance provisions and an Operating Experience review process (specific for decommissioning) should help to focus on real issues

  4. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions

  5. NPP Krsko Severe Accident Management Guidelines Implementation

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

    2002-01-01

    Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

  6. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  7. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  8. Mutation Detection in Patients With Advanced Cancer by Universal Sequencing of Cancer-Related Genes in Tumor and Normal DNA vs Guideline-Based Germline Testing.

    Science.gov (United States)

    Mandelker, Diana; Zhang, Liying; Kemel, Yelena; Stadler, Zsofia K; Joseph, Vijai; Zehir, Ahmet; Pradhan, Nisha; Arnold, Angela; Walsh, Michael F; Li, Yirong; Balakrishnan, Anoop R; Syed, Aijazuddin; Prasad, Meera; Nafa, Khedoudja; Carlo, Maria I; Cadoo, Karen A; Sheehan, Meg; Fleischut, Megan H; Salo-Mullen, Erin; Trottier, Magan; Lipkin, Steven M; Lincoln, Anne; Mukherjee, Semanti; Ravichandran, Vignesh; Cambria, Roy; Galle, Jesse; Abida, Wassim; Arcila, Marcia E; Benayed, Ryma; Shah, Ronak; Yu, Kenneth; Bajorin, Dean F; Coleman, Jonathan A; Leach, Steven D; Lowery, Maeve A; Garcia-Aguilar, Julio; Kantoff, Philip W; Sawyers, Charles L; Dickler, Maura N; Saltz, Leonard; Motzer, Robert J; O'Reilly, Eileen M; Scher, Howard I; Baselga, Jose; Klimstra, David S; Solit, David B; Hyman, David M; Berger, Michael F; Ladanyi, Marc; Robson, Mark E; Offit, Kenneth

    2017-09-05

    Guidelines for cancer genetic testing based on family history may miss clinically actionable genetic changes with established implications for cancer screening or prevention. To determine the proportion and potential clinical implications of inherited variants detected using simultaneous sequencing of the tumor and normal tissue ("tumor-normal sequencing") compared with genetic test results based on current guidelines. From January 2014 until May 2016 at Memorial Sloan Kettering Cancer Center, 10 336 patients consented to tumor DNA sequencing. Since May 2015, 1040 of these patients with advanced cancer were referred by their oncologists for germline analysis of 76 cancer predisposition genes. Patients with clinically actionable inherited mutations whose genetic test results would not have been predicted by published decision rules were identified. Follow-up for potential clinical implications of mutation detection was through May 2017. Tumor and germline sequencing compared with the predicted yield of targeted germline sequencing based on clinical guidelines. Proportion of clinically actionable germline mutations detected by universal tumor-normal sequencing that would not have been detected by guideline-directed testing. Of 1040 patients, the median age was 58 years (interquartile range, 50.5-66 years), 65.3% were male, and 81.3% had stage IV disease at the time of genomic analysis, with prostate, renal, pancreatic, breast, and colon cancer as the most common diagnoses. Of the 1040 patients, 182 (17.5%; 95% CI, 15.3%-19.9%) had clinically actionable mutations conferring cancer susceptibility, including 149 with moderate- to high-penetrance mutations; 101 patients tested (9.7%; 95% CI, 8.1%-11.7%) would not have had these mutations detected using clinical guidelines, including 65 with moderate- to high-penetrance mutations. Frequency of inherited mutations was related to case mix, stage, and founder mutations. Germline findings led to discussion or initiation of

  9. 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

  10. Severe Accident Research Program plan update

    International Nuclear Information System (INIS)

    1992-12-01

    In August 1989, the staff published NUREG-1365, ''Revised Severe Accident Research Program Plan.'' Since 1989, significant progress has been made in severe accident research to warrant an update to NUREG-1365. The staff has prepared this SARP Plan Update to: (1) Identify those issues that have been closed or are near completion, (2) Describe the progress in our understanding of important severe accident phenomena, (3) Define the long-term research that is directed at improving our understanding of severe accident phenomena and developing improved methods for assessing core melt progression, direct containment heating, and fuel-coolant interactions, and (4) Reflect the growing emphasis in two additional areas--advanced light water reactors, and support for the assessment of criteria for containment performance during severe accidents. The report describes recent major accomplishments in understanding the underlying phenomena that can occur during a severe accident. These include Mark I liner failure, severe accident scaling methodology, source term issues, core-concrete interactions, hydrogen transport and combustion, TMI-2 Vessel Investigation Project, and direct containment heating. The report also describes the major planned activities under the SARP over the next several years. These activities will focus on two phenomenological issues (core melt progression, and fuel-coolant interactions and debris coolability) that have significant uncertainties that impact our understanding and ability to predict severe accident phenomena and their effect on containment performance SARP will also focus on severe accident code development, assessment and validation. As the staff completes the research on severe accident issues that relate to current generation reactors, continued research will focus on efforts to independently evaluate the capability of new advanced light water reactor designs to withstand severe accidents

  11. The cost of nuclear accidents

    International Nuclear Information System (INIS)

    2015-01-01

    Proposed by a technical section of the SFEN, and based on a meeting with representatives of different organisations (OECD-NEA, IRSN, EDF, and European Nuclear Energy Forum), this publication addresses the economic consequences of a severe accident (level 6 or 7) within an electricity producing nuclear power plant. Such an assessment essentially relies on three pillars: release of radio-elements outside the reactor, the scenario of induced consequences, and the method of economic quantification. After a recall and a comment of safety arrangements, and of the generally admitted probability of such an accident, this document notices that several actors are concerned by nuclear energy and are trying to assess accident costs. The issue of how to assess a cost (or costs) of a nuclear accident is discussed: there are in fact several types of costs and consequences. Thus, some costs can be rather precisely quantified when some others can be difficult to assess or with uncertainty. The relevance of some cost categories appears to be a matter of discussion and one must not forget that consequences can occur on a long term. The need for methodological advances is outlined and three categories of technical objectives are identified for the assessment (efficiency of safety measures to be put forward to mitigate the risk via a better accident management, compensation of victims and nuclear civil responsibility, and comparison of electricity production sectors and assessment of externalisation to guide public choices). It is outlined that the impact of accidents depend on several factors, that the most efficient mean to limit consequences of accidents is of course to limit radioactive emissions

  12. Accident information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

  13. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  14. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  15. Advanced fuels safety comparisons

    International Nuclear Information System (INIS)

    Grolmes, M.A.

    1977-01-01

    The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

  16. Severe accident phenomena

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

    1995-02-01

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  17. Clinical Applicability of Whole-Exome Sequencing Exemplified by a Study in Young Adults with the Advanced Cryptogenic Cholestatic Liver Diseases

    Directory of Open Access Journals (Sweden)

    Maria Kulecka

    2017-01-01

    Full Text Available Background. The proper use of new medical tests in clinical practice requires the establishment of their value and range of diagnostic usefulness. While whole-exome sequencing (WES has already entered the medical practice, recognizing its diagnostic usefulness in multifactorial diseases has not yet been achieved. Aims. The objective of this study was to establish usability of WES in determining genetic background of chronic cholestatic liver disease (CLD in young patients. Methods. WES was performed on six young patients (between 17 and 22 years old with advanced fibrosis or cirrhosis due to CLD and their immediate families. Sequencing was performed on an Ion Proton sequencer. Results. On average, 19,673 variants were identified, of which from 7 to 14 variants of an individual were nonsynonymous, homozygous, recessively inherited, and considered in silico as pathogenic. Although monogenic cause of CLD has not been determined, several heterozygous rare variants and polymorphisms were uncovered in genes previously known to be associated with CLD, including ATP8B1, ABCB11, RXRA, and ABCC4, indicative of multifactorial genetic background. Conclusions. WES is a potentially useful diagnostic tool in determining genetic background of multifactorial diseases, but its main limitation results from the lack of opportunities for direct linkage between the uncovered genetic variants and molecular mechanisms of disease.

  18. 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

  19. Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

    International Nuclear Information System (INIS)

    Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

    2010-01-01

    SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

  20. 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

  1. InFusion: Advancing Discovery of Fusion Genes and Chimeric Transcripts from Deep RNA-Sequencing Data.

    Directory of Open Access Journals (Sweden)

    Konstantin Okonechnikov

    Full Text Available Analysis of fusion transcripts has become increasingly important due to their link with cancer development. Since high-throughput sequencing approaches survey fusion events exhaustively, several computational methods for the detection of gene fusions from RNA-seq data have been developed. This kind of analysis, however, is complicated by native trans-splicing events, the splicing-induced complexity of the transcriptome and biases and artefacts introduced in experiments and data analysis. There are a number of tools available for the detection of fusions from RNA-seq data; however, certain differences in specificity and sensitivity between commonly used approaches have been found. The ability to detect gene fusions of different types, including isoform fusions and fusions involving non-coding regions, has not been thoroughly studied yet. Here, we propose a novel computational toolkit called InFusion for fusion gene detection from RNA-seq data. InFusion introduces several unique features, such as discovery of fusions involving intergenic regions, and detection of anti-sense transcription in chimeric RNAs based on strand-specificity. Our approach demonstrates superior detection accuracy on simulated data and several public RNA-seq datasets. This improved performance was also evident when evaluating data from RNA deep-sequencing of two well-established prostate cancer cell lines. InFusion identified 26 novel fusion events that were validated in vitro, including alternatively spliced gene fusion isoforms and chimeric transcripts that include intergenic regions. The toolkit is freely available to download from http:/bitbucket.org/kokonech/infusion.

  2. A locally funded Puerto Rican parrot (Amazona vittata genome sequencing project increases avian data and advances young researcher education

    Directory of Open Access Journals (Sweden)

    Oleksyk Taras K

    2012-09-01

    Full Text Available Abstract Background Amazona vittata is a critically endangered Puerto Rican endemic bird, the only surviving native parrot species in the United States territory, and the first parrot in the large Neotropical genus Amazona, to be studied on a genomic scale. Findings In a unique community-based funded project, DNA from an A. vittata female was sequenced using a HiSeq Illumina platform, resulting in a total of ~42.5 billion nucleotide bases. This provided approximately 26.89x average coverage depth at the completion of this funding phase. Filtering followed by assembly resulted in 259,423 contigs (N50 = 6,983 bp, longest = 75,003 bp, which was further scaffolded into 148,255 fragments (N50 = 19,470, longest = 206,462 bp. This provided ~76% coverage of the genome based on an estimated size of 1.58 Gb. The assembled scaffolds allowed basic genomic annotation and comparative analyses with other available avian whole-genome sequences. Conclusions The current data represents the first genomic information from and work carried out with a unique source of funding. This analysis further provides a means for directed training of young researchers in genetic and bioinformatics analyses and will facilitate progress towards a full assembly and annotation of the Puerto Rican parrot genome. It also adds extensive genomic data to a new branch of the avian tree, making it useful for comparative analyses with other avian species. Ultimately, the knowledge acquired from these data will contribute to an improved understanding of the overall population health of this species and aid in ongoing and future conservation efforts.

  3. 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)

  4. 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)

  5. A cost-effectiveness analysis of first-line induction and maintenance treatment sequences in patients with advanced nonsquamous non-small-cell lung cancer in France

    Directory of Open Access Journals (Sweden)

    Taipale K

    2017-08-01

    Full Text Available Kaisa Taipale,1 Katherine B Winfree,2 Mark Boye,2 Mickael Basson,3 Ghassan Sleilaty,4 James Eaton,5 Rachel Evans,5 Christos Chouaid6 1Global Patient Outcomes and Real World Evidence International, Oy Eli Lilly Finland AB, Helsinki, Finland; 2Global Patient Outcomes and Real World Evidence, Eli Lilly and Company, Indianapolis, IN, USA; 3Corporate Affairs, Lilly France, Neuilly-sur-Seine, 4Bio-Medicines Medical Affairs, Lilly France, Neuilly-sur-Seine, France; 5ICON Health Economics and Epidemiology, ICON Plc, Milton Park, UK; 6Thoracic Oncology, Service de Pneumologie, Centre Hospitalier Intercommunal Créteil, Créteil, France Background: Comparative effectiveness and cost-effectiveness data for induction–maintenance (I–M sequences for the treatment of patients with nonsquamous non-small-cell lung cancer (nsqNSCLC are limited because of a lack of direct evidence. This analysis aimed to compare the cost-effectiveness of I–M pemetrexed with those of other I–M regimens used for the treatment of patients with advanced nsqNSCLC in the French health-care setting. Materials and methods: A previously developed global partitioned survival model was adapted to the France-only setting by restricting treatment sequences to include 12 I–M regimens most relevant to France, and incorporating French costs and resource-use data. Following a systematic literature review, network meta-analyses were performed to obtain hazard ratios for progression-free survival (PFS and overall survival (OS relative to gemcitabine + cisplatin (induction sequences or best supportive care (BSC (maintenance sequences. Modeled health-care benefits were expressed as life-years (LYs and quality-adjusted LYs (QALYs (estimated using French EuroQol five-dimension questionnaire tariffs. The study was conducted from the payer perspective (National Health Insurance. Cost- and benefit-model inputs were discounted at an annual rate of 4%. Results: Base-case results showed pemetrexed

  6. 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)

  7. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-01-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation

  8. #3653 Sequencing of plasma cfDNA from patients with locally advanced bladder cancer for surveillance and therapeutic efficacy monitoring

    DEFF Research Database (Denmark)

    Birkenkamp-Demtröder, Karin; Christensen, Emil; Sharma, Shruti

    ), 340 (67-2838) moderate and 223 (29-2955) low-impact SNVs or insertion-deletions (InDels) per tumor. High-impact mutations in known cancer genes such as TP53 (60%), KDM6A (34%), ARID1A (32%), RB1 (28%), BRCA2 (26%), FGFR3 (22%) and ERCC2 (20%) were identified with no significant difference between......, 50 patients diagnosed with locally advanced muscle-invasive bladder cancer (MIBC) and scheduled for chemotherapy were prospectively recruited between 2013 and 2017. In total, 42 patients received four cycles of cisplatin-based NAC prior to cystectomy and 82% showed response (pathologic downstaging......). Eight patients received 2-6 cycles of cisplatin-based first-line chemotherapy due to diagnosis of T4b or lymph node metastasis prior to cystectomy (3CR, 2PR, 2PD, 1 ongoing). So far, 8/50 patients (16%) experienced disease relapse and three patients had metastatic progression. The mean follow-up time...

  9. Sequencing of complete mitochondrial genomes confirms synonymization of Hyalomma asiaticum asiaticum and kozlovi, and advances phylogenetic hypotheses for the Ixodidae.

    Science.gov (United States)

    Liu, Zhi-Qiang; Liu, Yan-Feng; Kuermanali, Nuer; Wang, Deng-Feng; Chen, Shi-Jun; Guo, Hui-Ling; Zhao, Li; Wang, Jun-Wei; Han, Tao; Wang, Yuan-Zhi; Wang, Jie; Shen, Chen-Feng; Zhang, Zhuang-Zhi; Chen, Chuang-Fu

    2018-01-01

    Phylogeny of hard ticks (Ixodidae) remains unresolved. Mitochondrial genomes (mitogenomes) are increasingly used to resolve phylogenetic controversies, but remain unavailable for the entire large Hyalomma genus. Hyalomma asiaticum is a parasitic tick distributed throughout the Asia. As a result of great morphological variability, two subspecies have been recognised historically; until a morphological data-based synonymization was proposed. However, this hypothesis was never tested using molecular data. Therefore, objectives of this study were to: 1. sequence the first Hyalomma mitogenome; 2. scrutinise the proposed synonymization using molecular data, i.e. complete mitogenomes of both subspecies: H. a. asiaticum and kozlovi; 3. conduct phylogenomic and comparative analyses of all available Ixodidae mitogenomes. Results corroborate the proposed synonymization: the two mitogenomes are almost identical (99.6%). Genomic features of both mitogenomes are standard for Metastriata; which includes the presence of two control regions and all three "Tick-Box" motifs. Gene order and strand distribution are perfectly conserved for the entire Metastriata group. Suspecting compositional biases, we conducted phylogenetic analyses (29 almost complete mitogenomes) using homogeneous and heterogeneous (CAT) models of substitution. The results were congruent, apart from the deep-level topology of prostriate ticks (Ixodes): the homogeneous model produced a monophyletic Ixodes, but the CAT model produced a paraphyletic Ixodes (and thereby Prostriata), divided into Australasian and non-Australasian clades. This topology implies that all metastriate ticks have evolved from the ancestor of the non-Australian branch of prostriate ticks. Metastriata was divided into three clades: 1. Amblyomminae and Rhipicephalinae (Rhipicephalus, Hyalomma, Dermacentor); 2. Haemaphysalinae and Bothriocrotoninae, plus Amblyomma sphenodonti; 3. Amblyomma elaphense, basal to all Metastriata. We conclude that

  10. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

  11. 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 Advanced Light Water Reactor (ALWR) Program in the U.S. is a cooperative, cost-sharing undertaking between the U.S. government, industry, and a number of international participants, with the objective of developing the next generation of nuclear power plants. The ALWR designs emphasize improvements in safety and operational reliability through simplification, improved safety margins, innovative passive safety systems, enhanced man-machine interfaces, and incorporation of the lessons learned from the operation of existing LWR plants. An important component of the improved safety characteristics of ALWRs is the consideration of severe accidents in the plant design. The U.S. Department of Energy (DOE) initiated the Advanced Reactor Severe Accident Program (ARSAP) to assist in the transfer of severe accident technology from the U.S. national laboratories to the industry to implement this approach. The basic design requirements for this new generation of nuclear power plants were developed, under the management of the Electric Power Research Institute (EPRI) by the utilities and documented in the Utility Requirements Document (URD). The URD safety policy is based on the traditional 'defense-in-depth' approach, which emphasizes prevention through safety systems which prevent accidents from progressing to core damage, and mitigation to ensure that accidents are mitigated and contained. In a major departure from previous practice, severe accidents, including postulated core melt events, are specifically included in the defense-in-depth design considerations for ALWRs. As a result of this approach, the emergency planning assumptions and criteria warrant a review and reevaluation for ALWR designs. ALWRs present a risk profile that is significantly different than that which served as the basis for the emergency planning requirements for operating plants. The determination of this profile necessarily requires the characterization of the severe accident response of ALWRs

  12. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix XI. Analysis of comments on the draft WASH-1400 report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

  13. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effect of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  14. An analysis of uncertainty and of dependence on season of year of ingestion population dose arising from design basis accidents in advanced thermal reactors

    International Nuclear Information System (INIS)

    Nair, S.; Ponting, A.C.

    1985-03-01

    The results of a detailed study of ingestion collective dose from five limiting PWR design basis releases are presented, the PWR being chosen as being typical of an advanced thermal reactor for which source terms are readily available. The ingestion collective dose was calculated for a range of wind direction/weather scenarios for releases from a typical U.K. rural and a U.K. semi-urban site and scenarios identified where the ingestion pathway was of potential significance. The dependence of the ingestion collective dose for these cases on the season of year when the release occurs was investigated. An analysis was carried out of the uncertainty in the ''worst case'' ingestion calculations arising from uncertainties in foodchain input parameters. An efficient but comprehensive set of dynamic foodchain computer models was produced and the literature surveyed to produce probability distribution functions (PDF's) for all relevant independent input data items. These were used to produce output PDF's for food contamination levels and for ingestion collective dose from the five releases. Finally, the study has highlighted several areas central to ingestion collective dose assessments where the available data are inadequate. This led to the formulation of a set of future research requirements which will need to be met both to obtain a better fundamental understanding of foodchain transfer and to reduce uncertainties in ingestion collective dose assessments. (author)

  15. 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...

  16. Nuclear accident dosimetry

    International Nuclear Information System (INIS)

    1982-01-01

    The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

  17. Nuclear accident dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-12-31

    The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

  18. Implementation of severe accident management measures - Summary and conclusions

    International Nuclear Information System (INIS)

    2002-01-01

    The objectives of the meeting were: 1) to exchange information on activities in the area of SAM implementation and on the rationale for such actions, 2) to monitor progress made, 3) to identify cases of agreement or disagreement, 4) to discuss future orientations of work, 5) to make recommendations to the CSNI. Session summaries prepared by the Chairpersons and discussed by the whole writing group are given in Annex. During the first session, 'SAM Programmes Implementation', papers from one regulator and several utilities and national research institutes were presented to outline the status of implementation of SAM programmes in countries like Switzerland, Russia, Spain, Finland, Belgium and Korea. Also, the contribution of SAM to the safety of Japanese plants (in terms of core damage frequency) was quantified in a paper. One paper gave an overview on the situation regarding SAM implementation in Europe. The second session, 'SAM Approach', provided background and bases for Severe Accident Management in countries like Sweden, Japan, Germany and Switzerland, as well as for hardware features in advanced light water reactor designs, such as the European Pressurised Reactor (EPR), regarding Severe Accident Management. The third session, 'SAM Mitigation Measures', was about hardware measures, in particular those oriented towards hydrogen mitigation where fundamentally different approaches have been taken in Scandinavian countries, France, Germany and Korea. Three papers addressed specific contributions from research to provide a broader basis for the assumptions made in certain computer codes used for the assessment of plant risk arising from beyond-design accident sequences. The fourth session, 'Implementation of SAM Measures on VVER-1000 Reactors', was about the status of work on Severe Accident Management implementation in VVER reactors of existing design and in a new plant currently under construction. The overall picture is that Severe Accident Management has been

  19. 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)

  20. Dose calculations for severe LWR accident scenarios

    International Nuclear Information System (INIS)

    Margulies, T.S.; Martin, J.A. Jr.

    1984-05-01

    This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor Accident Consequences) code. Whole body and thyroid doses are presented for a selected set of weather cases. For each weather case these calculations were performed for various times and distances including three different dose pathways - cloud (plume) shine, ground shine and inhalation. During an emergency this information can be useful since it is immediately available for projecting offsite radiological doses based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms). It can be used for emergency drill scenario development as well

  1. Qualification of the core model DYN3D coupled with the code ATHLET as an advanced tool for the accident analysis of VVER type reactors. Pt. 2. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Rohde, U.

    2002-10-01

    differences in the thermohydraulics were assumed in the difficult modelling of the vertical once-through steam generator with steam superheating. Sensitivity analyses which considered the influence of the nodalisation and the impact of the coolant mixing model were performed for the DYN3D-ATHLET solution of the OECD benchmark. The solution of the benchmarks essentially contributed to the qualification of the code complex DYN3D-ATHLET as an advanced tool for the accident analysis for both VVER type reactors and Western PWRs. (orig.) [de

  2. Supervisor's accident investigation handbook

    International Nuclear Information System (INIS)

    1980-02-01

    This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

  3. NIF: Impacts of chemical accidents and comparison of chemical/radiological accident approaches

    International Nuclear Information System (INIS)

    Lazaro, M.A.; Policastro, A.J.; Rhodes, M.

    1996-01-01

    The US Department of Energy (DOE) proposes to construct and operate the National Ignition Facility (NIF). The goals of the NIF are to (1) achieve fusion ignition in the laboratory for the first time by using inertial confinement fusion (ICF) technology based on an advanced-design neodymium glass solid-state laser, and (2) conduct high-energy-density experiments in support of national security and civilian applications. The primary focus of this paper is worker-public health and safety issues associated with postulated chemical accidents during the operation of NIF. The key findings from the accident analysis will be presented. Although NIF chemical accidents will be emphasized, the important differences between chemical and radiological accident analysis approaches and the metrics for reporting results will be highlighted. These differences are common EIS facility and transportation accident assessments

  4. Serum homocysteine levels in cerebrovascular accidents.

    Science.gov (United States)

    Zongte, Zolianthanga; Shaini, L; Debbarma, Asis; Singh, Th Bhimo; Devi, S Bilasini; Singh, W Gyaneshwar

    2008-04-01

    Hyperhomocysteinemia has been considered an independent risk factor in the development of stroke. The present study was undertaken to evaluate serum homocysteine levels in patients with cerebrovascular accidents among the Manipuri population and to compare with the normal cases. Ninety-three cerebrovascular accident cases admitted in the hospital were enrolled for the study and twenty-seven age and sex matched individuals free from cerebrovascular diseases were taken as control group. Serum homocysteine levels were estimated by ELISA method using Axis homocysteine EIA kit manufactured by Ranbaxy Diagnostic Ltd. India. The finding suggests that hyperhomocysteinemia is associated with cerebrovascular accident with male preponderance, which increases with advancing age. However, whether hyperhomocysteinemia is the cause or the result of cerebrovascular accidents needs further investigations.

  5. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  6. 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.

  7. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  8. ACCIDENT PHENOMENA OF RISK IMPORTANCE PROJECT - Continued RESEARCH CONCERNING SEVERE ACCIDENT PHENOMENA AND MANAGEMENT IN Sweden

    International Nuclear Information System (INIS)

    Rolandson, S.; Mueller, F.; Loevenhielm, G.

    1997-01-01

    Since 1988 all reactors in Sweden have mitigating measures, such as filtered vents, implemented. In parallel with the work of implementing these measures, a cooperation effort (RAMA projects) between the Swedish utilities and the Nuclear Power Inspectorate was performed to acquire sufficient knowledge about severe accident research work. The on-going project has the name Accident Phenomena of Risk Importance 3. In this paper, we will give background information about severe accident management in Sweden. In the Accident Phenomena of Risk Importance 3 project we will focus on the work concerning coolability of melted core in lower plenum which is the main focus of the In-vessel Coolability Task Group within the Accident Phenomena of Risk Importance 3 project. The Accident Phenomena of Risk Importance 3 project has joined on international consortium and the in-vessel cooling experiments are performed by Fauske and Associates, Inc. in Burr Ridge, Illinois, United States America, Sweden also intends to do one separate experiment with one instrument penetration we have in Swedish/Finnish BWR's. Other parts of the Accident Phenomena of Risk Importance 3 project, such as support to level 2 studies, the research at Royal Institute of Technology and participation in international programs, such as Cooperative Severe Accident Research Program, Advanced Containment Experiments and PHEBUS will be briefly described in the paper

  9. Planning for the Handling of Radiation Accidents

    International Nuclear Information System (INIS)

    1969-01-01

    The developing atomic energy programmes and the widespread use of radiation sources in medicine, agriculture, industry and research have had admirable safety records. Throughout the world the number of known accidents in which persons have been exposed to harmful am ounts of ionizing radiation is relatively small, and only a few deaths have occurred. Meticulous precautions are being taken to maintain this good record in all work with radiation sources and to keep the exposure of persons as low as practicable. In spite of all the precautions that are taken, accidents may occur and they may be accompanied by the injury or death of persons and damage to property. It is only prudent to take those steps that are practicable to prevent accidents and to plan in advance the emergency action that would limit the injuries and damage caused by those accidents that do occur. Emergency plans should be sufficiently broad to cover unforeseen or very improbable accidents as well as those that are considered credible. Some accidents may involve only the workers in an establishment, those working directly with the source and possibly their colleagues. Other accidents may have consequences, notably in the form of radioactive contamination of the environment, that affect the general public, possibly far from the site of the accident. The preparation of plans for dealing with radiation accidents is therefore obligatory both for the various authorities that are responsible for protecting the health and the food and water supplies of the public, and for the operator of an installation containing radiation sources.

  10. 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

  11. Occurrence and countermeasures of urban power grid accident

    Science.gov (United States)

    Wei, Wang; Tao, Zhang

    2018-03-01

    With the advance of technology, the development of network communication and the extensive use of power grids, people can get to know power grid accidents around the world through the network timely. Power grid accidents occur frequently. Large-scale power system blackout and casualty accidents caused by electric shock are also fairly commonplace. All of those accidents have seriously endangered the property and personal safety of the country and people, and the development of society and economy is severely affected by power grid accidents. Through the researches on several typical cases of power grid accidents at home and abroad in recent years and taking these accident cases as the research object, this paper will analyze the three major factors that cause power grid accidents at present. At the same time, combining with various factors and impacts caused by power grid accidents, the paper will put forward corresponding solutions and suggestions to prevent the occurrence of the accident and lower the impact of the accident.

  12. 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

  13. 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.)

  14. 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)

  15. 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

  16. 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.)

  17. 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.)

  18. 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)

  19. Nuclear accidents and epidemiology

    International Nuclear Information System (INIS)

    1987-01-01

    A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

  20. Accidents (FARS) (National)

    Data.gov (United States)

    Department of Transportation — Accident - (1975-current): This data file (NTAD) contains information about crash characteristics and environmental conditions at the time of the crash. There is one...

  1. An overview of selected severe accident research and applications

    International Nuclear Information System (INIS)

    Hammersley, R.J.; Henry, R.E.

    2004-01-01

    Severe accident research is being conducted world wide by industry organizations, utilities, and regulatory agencies. As this research is disseminated, it is being applied by utilities when they perform their Individual Plant Examinations (IPEs) and consider the preparation of Accident Management programs. The research is associated with phenomenological assessments of containment challenges and associated uncertainties, severe accident codes and analysis tools, systematic evaluation processes, and accident management planning. The continued advancement of this research and its applications will significantly contribute to the enhanced safety and operation of nuclear power plants. (author)

  2. A framework for the assessment of severe accident management strategies

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.

    1993-09-01

    Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed

  3. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    1984-03-01

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  4. Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

    International Nuclear Information System (INIS)

    Frid, W.; Sandervaag, O.

    1997-01-01

    The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

  5. A framework for the assessment of severe accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kastenberg, W.E. [ed.; Apostolakis, G.; Dhir, V.K. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering] [and others

    1993-09-01

    Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable of propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.

  6. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  7. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  8. Instrumentation availability for a pressurized water reactor with a large dry containment during severe accidents

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1991-03-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a pressurized water reactor with a large dry containment. Results from this evaluation include the following: (a) identification of plant conditions that would impact instrument performance and information needs during severe accidents, (b) definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences, and (c) assessment of the availability of plant instrumentation during severe accidents. 16 refs., 3 figs., 4 tabs

  9. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  10. 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.

  11. Comparative analysis of a hypothetical 0.1 $/SEC transient overpower accident in an irradiated LMFBR core using different computer models

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Fremont, R. de; Renard, A.

    1982-01-01

    The Report gives the results of comparative calculations performed by the Whole Core Accident Codes Group which is a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee for a hypothetical transient overpower accident in an irradiated LMFBR core. Different computer codes from members of the European Community and the United States were used. The calculations are based on a Benchmark problem, using commonly agreed input data for the most important phenomena, such as the fuel pin failure threshold, FCl parameters, etc. Beside this, results with alternative assumptions for theoretical modelling are presented with the scope to show in a parametric way the influence of more advanced modelling capabilities and/or better (so-called best estimate) input data for the most important phenomena on the accident sequences

  12. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    Clauss, D.B.

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs

  13. 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

  14. Accident analysis for PRC-II reactor

    International Nuclear Information System (INIS)

    Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng

    1997-12-01

    The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward

  15. 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.

  16. Bus accident severity and passenger injury: evidence from Denmark

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Kaplan, Sigal

    2014-01-01

    Purpose Bus safety is a concern not only in developing countries, but also in the U.S. and Europe. In Denmark, disentangling risk factors that are positively or negatively related to bus accident severity and injury occurrence to bus passengers can contribute to promote safety as an essential...... principle of sustainable transit and advance the vision “every accident is one too many”. Methods Bus accident data were retrieved from the national accident database for the period 2002–2011. A generalized ordered logit model allows analyzing bus accident severity and a logistic regression enables...... examining occurrence of injury to bus passengers. Results Bus accident severity is positively related to (i) the involvement of vulnerable road users, (ii) high speed limits, (iii) night hours, (iv) elderly drivers of the third party involved, and (v) bus drivers and other drivers crossing in yellow or red...

  17. 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. Volume 2 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  18. 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

  19. Chernobyl accident and Danmark

    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. Volume 1 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  20. Communication and industrial accidents

    NARCIS (Netherlands)

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the

  1. Chapter 6: Accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2018-04-01

    Th chapter 6 presents the accidents of: 1) Stimos (Italy - May, 1975); 2) San Salvador (El Salvador - February 5, 1989); 3) Soreq (Israel - June 21, 1990); 4) Nesvizh (Belarus - October 26, 1991); 5) Illinois (USA - February, 1965); 6)Maryland (EUA - December 11, 1991); 7)Hanoi (Vietnam -November 17, 1992); 8)Fleurus (Belgium - March 11, 2006) and final remarks on accidents.

  2. A physical tool for severe accident mitigation studies

    Energy Technology Data Exchange (ETDEWEB)

    Marie, N., E-mail: nathalie.marie@cea.fr [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Bachrata, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France); Barjot, F. [EDF R& D, SINETICS, F-93141 Clamart (France); Marrel, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Gossé, S. [CEA, DEN, DPC, F-91191 Gif Sur Yvette (France); Bertrand, F. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France)

    2016-12-01

    Highlights: • Physical tool for mitigation studies devoted to SFR safety. • Physical models to describe the material discharge from core. • Comparison to SIMMER III results. • Studies for ASTRID safety assessment and support to core design. - Abstract: Within the framework of the Generation IV Sodium-cooled Fast Reactors (SFR) R&D program of CEA, the core behavior in case of severe accidents is being assessed. Such transients are usually simulated with mechanistic codes (such as SIMMER-III). As a complement to this code, which gives reference accidental transient, a physico-statistical approach is currently followed; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running simulation of extended accident sequences coupling low-dimensional physical models to advanced statistical analysis techniques. In this context, this paper presents such a low-dimensional physical tool (models and simulation results) dedicated to molten core materials discharge. This 0D tool handles heat transfers from molten (possibly boiling) pools, fuel crust evolution, phase separation/mixing of fuel/steel pools, radial thermal erosion of mitigation tubes, discharge of core materials and associated axial thermal erosion of mitigation tubes. All modules are coupled with a global neutronic evolution model of the degraded core. This physical tool is used to study and to define mitigation features (function of tubes devoted to mitigation inside the core, impact of absorbers falling into the degraded core…) to avoid energetic core recriticality during a secondary phase of a potential severe accident. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties would enable sensitivity analysis studies. This physical tool is described before presenting its comparison against SIMMER-III code results, including a space-and energy-dependent neutron transport kinetic

  3. Group unified accident reporting database (GUARD)

    International Nuclear Information System (INIS)

    Koene, W.; Waterfall, K.W.

    1991-01-01

    Significant advances have been made in recent years in enhancing the standard of safety within Shell Companies, such that safety has now been raised to a status equal to other primary business objectives. It is widely accepted that accident prevention is part of good business practice, and that a safe operation is normally an efficient operation. Safety programmes are being widely implemented which involve all employees from top management right down to the workforce including the contract staff, and the benefits are being realized. The effectiveness of any safety programme, however, must be continuously monitored, and in this respect injury and accident statistics play an important role as a prime indicator of safety performance. Statistics form part of the safety management process indicating the success of the safety programmes being implemented, and highlighting areas of weakness. Statistical information relating to the number and frequency of accidents, significant as it is, tells us little about how the accidents occur, or about how to improve the intrinsic safety of the operations. More detailed information on accident causes and lessons derived from the investigation of non-injurious accidents and near-misses is required for this, and for the setting of appropriate remedial actions. This paper concentrates on the feedback from accidents which have already occurred. This feedback plays a vital role as an indicator of safety performance upon which to judge the effectiveness of safety programmes, and also to provide important information relating to the immediate and underlying causes of accidents. To meet these requirements, however, a system for recording analyzing and communicating safety data is essential

  4. Radiological accidents in medical practice

    International Nuclear Information System (INIS)

    Cardenas Herrera, Juan

    2012-01-01

    Different radiological accidents that may occur in medical practice are shown. The following topics are focused: accident statistics for medical exposure, accidental medical exposures, radiotherapy accidents and potential accidental scenarios [es

  5. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  6. [Accidents and injuries at work].

    Science.gov (United States)

    Standke, W

    2014-06-01

    In the case of an accident at work, the person concerned is insured by law according to the guidelines of the Sozialgesetzbuch VII as far as the injuries have been caused by this accident. The most important source of information on the incident in question is the accident report that has to be sent to the responsible institution for statutory accident insurance and prevention by the employer, if the accident of the injured person is fatal or leads to an incapacity to work for more than 3 days (= reportable accident). Data concerning accidents like these are sent to the Deutsche Gesetzliche Unfallversicherung (DGUV) as part of a random sample survey by the institutions for statutory accident insurance and prevention and are analyzed statistically. Thus the key issues of accidents can be established and used for effective prevention. Although the success of effective accident prevention is undisputed, there were still 919,025 occupational accidents in 2011, with clear gender-related differences. Most occupational accidents involve the upper and lower extremities. Accidents are analyzed comprehensively and the results are published and made available to all interested parties in an effort to improve public awareness of possible accidents. Apart from reportable accidents, data on the new occupational accident pensions are also gathered and analyzed statistically. Thus, additional information is gained on accidents with extremely serious consequences and partly permanent injuries for the accident victims.

  7. 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

  8. Vaporization of structural materials in severe accidents

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1982-01-01

    Vaporized structural materials form the bulk of aerosol particles that can transport fission products in severe LWR accidents. As part of the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, a model has been developed based on a mass transport coefficient to describe the transport of materials from the surface of a molten pool. In many accident scenarios, the coefficient can be calculated from existing correlations for mass transfer by natural convection. Data from SASCHA fuel melting tests (Karlsruhe, Germany) show that the partial pressures of many of the melt components (Fe, Cr, Co, Mn, Sn) required for the model can be calculated from the vapor pressures of the pure species and Raoult's law. These calculations indicate much lower aerosol concentrations than reported in previous studies

  9. Accident tolerant fuel analysis

    International Nuclear Information System (INIS)

    2014-01-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant

  10. Database on aircraft accidents

    International Nuclear Information System (INIS)

    Nishio, Masahide; Koriyama, Tamio

    2012-09-01

    The Reactor Safety Subcommittee in the Nuclear Safety and Preservation Committee published the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' as the standard method for evaluating probability of aircraft crash into nuclear reactor facilities in July 2002. In response to the report, Japan Nuclear Energy Safety Organization has been collecting open information on aircraft accidents of commercial airplanes, self-defense force (SDF) airplanes and US force airplanes every year since 2003, sorting out them and developing the database of aircraft accidents for latest 20 years to evaluate probability of aircraft crash into nuclear reactor facilities. This year, the database was revised by adding aircraft accidents in 2010 to the existing database and deleting aircraft accidents in 1991 from it, resulting in development of the revised 2011 database for latest 20 years from 1991 to 2010. Furthermore, the flight information on commercial aircrafts was also collected to develop the flight database for latest 20 years from 1991 to 2010 to evaluate probability of aircraft crash into reactor facilities. The method for developing the database of aircraft accidents to evaluate probability of aircraft crash into reactor facilities is based on the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' described above. The 2011 revised database for latest 20 years from 1991 to 2010 shows the followings. The trend of the 2011 database changes little as compared to the last year's one. (1) The data of commercial aircraft accidents is based on 'Aircraft accident investigation reports of Japan transport safety board' of Ministry of Land, Infrastructure, Transport and Tourism. 4 large fixed-wing aircraft accidents, 58 small fixed-wing aircraft accidents, 5 large bladed aircraft accidents and 114 small bladed aircraft accidents occurred. The relevant accidents for evaluating

  11. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery managment concevtrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that 'active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk', and goes further in considering and formulating the key issue: 'The most fruitful path to follow in reducing risk even further is through the planning of accident management.' (author)

  12. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  13. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery management concentrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk, and goes further in considering and formulating the key issue: The most fruitful path to follow in reducing risk even further is through the planning of accident management

  14. 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.

  15. Targeted sequencing of plant genomes

    Science.gov (United States)

    Mark D. Huynh

    2014-01-01

    Next-generation sequencing (NGS) has revolutionized the field of genetics by providing a means for fast and relatively affordable sequencing. With the advancement of NGS, wholegenome sequencing (WGS) has become more commonplace. However, sequencing an entire genome is still not cost effective or even beneficial in all cases. In studies that do not require a whole-...

  16. Social impact of accidents

    International Nuclear Information System (INIS)

    Kuroda, Isao

    1997-01-01

    There is the quite big difference between technological risk and social risk feeling. Various biases of social and sensational factors on accidents must be considered to recognize this difference. 'How safe is safe enough' is the perpetual thema concerning with not only technology but also sociology. The safety goal in aircraft design and how making effort to improve the present safety status in civil jet aircrafts is discussed as an example of social risk allowance. INSAG under IAEA started to discuss the safety culture after Chernobyl nuclear power plant accident on 1986. Safety culture and risk communication are the most important procedures to relieve the social impact for accidents. (author)

  17. Severe accident behavior

    International Nuclear Information System (INIS)

    Denning, R.S.

    1986-01-01

    The purpose of this paper is to provide an overview of severe accident behavior. The term source term is defined and a brief history of the regulatory use of source term is presented. The processes in severe accidents in light water reactors are described with particular emphasis on the relationships between accident thermal-hydraulics and chemistry. Those factors which have the greatest impact on predicted source terms are identified. Design differences between plants that affect source term estimation are also described. The principal unresolved issues are identified that are the focus of ongoing research and debate in the technical community

  18. Management of accident risks

    International Nuclear Information System (INIS)

    Compes, P.C.

    1987-01-01

    The example of the Chernobyl accident and the statistics of the occurrence of accidents make clear the threat to humanity, if one cannot guarantee successful accident prevention in the use and distribution of the projects aimed at. The science of safety, as it is known in the Wuppertal model, makes its contribution to this vital task for the human community. It makes it necessary to create the essential dates and concepts, the methods, principles and techniques based on them and the associated instrumentation. (DG) [de

  19. 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)

  20. Overview of severe accident research at KAERI

    International Nuclear Information System (INIS)

    Kim, H.D.; Kim, S.B.; Hong, S.W.; Kim, D.H.

    2000-01-01

    The severe accident research program at Korea Atomic Energy Research Institute, within the framework of governmental 10 year long-term nuclear R and D program, aims at the development of assessment techniques and accident management strategies for the prevention and mitigation of potential risk. The research program includes experimental efforts, development of phenomena specific models and development of an integrated computer code. The results of research program is intended to be utilized for the design of the advanced light water reactor and development of accident management strategies for the operating reactors. The main focused areas of recent investigation at KAERI are experiments on in-vessel core debris retention (SONATA-IV) and fuel coolant interaction (TROI) along with the development of models and integrated computer code (MIDAS). (author)

  1. Dynamic Sequence Assignment.

    Science.gov (United States)

    1983-12-01

    D-136 548 DYNAMIIC SEQUENCE ASSIGNMENT(U) ADVANCED INFORMATION AND 1/2 DECISION SYSTEMS MOUNTAIN YIELW CA C A 0 REILLY ET AL. UNCLSSIIED DEC 83 AI/DS...I ADVANCED INFORMATION & DECISION SYSTEMS Mountain View. CA 94040 84 u ,53 V,..’. Unclassified _____ SCURITY CLASSIFICATION OF THIS PAGE REPORT...reviews some important heuristic algorithms developed for fas- ter solution of the sequence assignment problem. 3.1. DINAMIC MOGRAMUNIG FORMULATION FOR

  2. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  3. 22 CFR 102.11 - Arranging for the payment of expenses attendant upon an accident.

    Science.gov (United States)

    2010-04-01

    ... FUNCTIONS CIVIL AVIATION United States Aircraft Accidents Abroad § 102.11 Arranging for the payment of... authorizes in advance the expenditure of such funds on a reimbursable basis. In the absence of such advance...

  4. 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)

  5. Synergy effect in accident simulation

    International Nuclear Information System (INIS)

    Alba, C.; Carlin, F.; Chenion, J.; Gaussens, G.; Le Meur, M.; Petitjean, M.

    1984-05-01

    Accidental breaking of PWR coolant canalization would entail water vaporization into confinement enclosure. Equipments would be simultaneously subjected to temperature and pressure increase, chemical spray, and radiation action of reactor core products. Some equipments have to work after accident in order to stop reactor running and blow out water calories. Usually, in France, accident simulation tests are carried out sequentialy: irradiation followed by thermodynamical and chemical tests. Equipments working is essentially due to those polymer materials behaviour. Is the polymers behaviour the same when they are either subjected to sequential test, or an accident (simultaneous action of irradiation and thermodynamical and chemical sequence). In order to answer to this question, nine polymer materials were subjected to simultaneous and sequential test in CESAR cell. Experiments were carried out in CESAR device with thermodynamical chocks and a temperature and pressure decrease profil in presence or without irradiation. So, the test is either simultaneous or sequential. Mechanical properties change are determined for the following polymeric materials. Two polyamide-imide varnishes used in motors and coils; one epoxydic resin, glass fiber charged (electrical insulating); polyphenylene sulfide, glass fiber charged, the Ryton R4 (electrical insulating); three elastomeric materials: Hypalon, fire proof by bromine or by alumina EPDM (cables jacket); VAMAC which is a polyethylene methyl polymethacrylate copolymer; then a silicon thermoset material glass fiber charged (electrical insulating). After test, usually, mechanical and electrical properties change of polymer materials show sequential experiment is more severe than simultaneous test however, Hypalon does not follow this law. For this polymer simultaneous test appears more severe than sequential experiment [fr

  6. Accident resistant transport container

    Science.gov (United States)

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  7. Accident resistant transport container

    International Nuclear Information System (INIS)

    Andersen, J.A.; Cole, J.K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident

  8. Big nuclear accidents

    International Nuclear Information System (INIS)

    Marshall, W.; Billingon, D.E.; Cameron, R.F.; Curl, S.J.

    1983-09-01

    Much of the debate on the safety of nuclear power focuses on the large number of fatalities that could, in theory, be caused by extremely unlikely but just imaginable reactor accidents. This, along with the nuclear industry's inappropriate use of vocabulary during public debate, has given the general public a distorted impression of the risks of nuclear power. The paper reviews the way in which the probability and consequences of big nuclear accidents have been presented in the past and makes recommendations for the future, including the presentation of the long-term consequences of such accidents in terms of 'loss of life expectancy', 'increased chance of fatal cancer' and 'equivalent pattern of compulsory cigarette smoking'. The paper presents mathematical arguments, which show the derivation and validity of the proposed methods of presenting the consequences of imaginable big nuclear accidents. (author)

  9. Accidents in perspective

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1989-01-01

    The nuclear industry perspective and the public perspective on big nuclear accidents and leukaemia near nuclear sites are discussed. The industry perspective is that big accidents are so unlikely as to be virtually impossible and that leukaemia is not specifically associated with nuclear installations. Clusters of cancer with statistical significance occur in major cities. The public perspective is coloured by a prejudice and myth: the fear of radiation. The big nuclear accident is seen therefore as much more unacceptable than any other big accident. Risks associated with Sizewell-B nuclear station and the liquid gas depot at Canvey Island are discussed. The facts and figures are presented as tables and graphs. Given conflicting interpretations of the leukaemia problem the public inclines towards the more pessimistic view. (author)

  10. Boating Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  11. Occupational Accidents And Preventive Measures

    CERN Document Server

    Fassnacht, V

    2006-01-01

    This report presents the 2005 statistics concerning occupational accidents involving members of the CERN personnel and contractors' personnel. It sets out the accident frequency and severity rates and provides a breakdown of accidents by cause and injury. It also contains a summary analysis of the most serious accidents and the associated recommendations.

  12. Accidents with sulfuric acid

    OpenAIRE

    Rajković Miloš B.

    2006-01-01

    Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eigh...

  13. 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

  14. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

  15. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

  16. 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

  17. Road Traffic Accident Analysis of Ajmer City Using Remote Sensing and GIS Technology

    Science.gov (United States)

    Bhalla, P.; Tripathi, S.; Palria, S.

    2014-12-01

    With advancement in technology, new and sophisticated models of vehicle are available and their numbers are increasing day by day. A traffic accident has multi-facet characteristics associated with it. In India 93% of crashes occur due to Human induced factor (wholly or partly). For proper traffic accident analysis use of GIS technology has become an inevitable tool. The traditional accident database is a summary spreadsheet format using codes and mileposts to denote location, type and severity of accidents. Geo-referenced accident database is location-referenced. It incorporates a GIS graphical interface with the accident information to allow for query searches on various accident attributes. Ajmer city, headquarter of Ajmer district, Rajasthan has been selected as the study area. According to Police records, 1531 accidents occur during 2009-2013. Maximum accident occurs in 2009 and the maximum death in 2013. Cars, jeeps, auto, pickup and tempo are mostly responsible for accidents and that the occurrence of accidents is mostly concentrated between 4PM to 10PM. GIS has proved to be a good tool for analyzing multifaceted nature of accidents. While road safety is a critical issue, yet it is handled in an adhoc manner. This Study is a demonstration of application of GIS for developing an efficient database on road accidents taking Ajmer City as a study. If such type of database is developed for other cities, a proper analysis of accidents can be undertaken and suitable management strategies for traffic regulation can be successfully proposed.

  18. Database on aircraft accidents

    International Nuclear Information System (INIS)

    Nishio, Masahide; Koriyama, Tamio

    2013-11-01

    The Reactor Safety Subcommittee in the Nuclear Safety and Preservation Committee published 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' as the standard method for evaluating probability of aircraft crash into nuclear reactor facilities in July 2002. In response to this issue, Japan Nuclear Energy Safety Organization has been collecting open information on aircraft accidents of commercial airplanes, self-defense force (SDF) airplanes and US force airplanes every year since 2003, sorting out them and developing the database of aircraft accidents for the latest 20 years to evaluate probability of aircraft crash into nuclear reactor facilities. In this report the database was revised by adding aircraft accidents in 2011 to the existing database and deleting aircraft accidents in 1991 from it, resulting in development of the revised 2012 database for the latest 20 years from 1992 to 2011. Furthermore, the flight information on commercial aircrafts was also collected to develop the flight database for the latest 20 years from 1992 to 2011 to evaluate probability of aircraft crash into reactor facilities. The method for developing the database of aircraft accidents to evaluate probability of aircraft crash into reactor facilities is based on the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' described above. The 2012 revised database for the latest 20 years from 1992 to 2011 shows the followings. The trend of the 2012 database changes little as compared to the last year's report. (1) The data of commercial aircraft accidents is based on 'Aircraft accident investigation reports of Japan transport safety board' of Ministry of Land, Infrastructure, Transport and Tourism. The number of commercial aircraft accidents is 4 for large fixed-wing aircraft, 58 for small fixed-wing aircraft, 5 for large bladed aircraft and 99 for small bladed aircraft. The relevant accidents

  19. 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.

  20. Injury protection and accident causation parameters for vulnerable road users based on German In-Depth Accident Study GIDAS.

    Science.gov (United States)

    Otte, Dietmar; Jänsch, Michael; Haasper, Carl

    2012-01-01

    Within a study of accident data from GIDAS (German In-Depth Accident Study), vulnerable road users are investigated regarding injury risk in traffic accidents. GIDAS is the largest in-depth accident study in Germany. Due to a well-defined sampling plan, representativeness with respect to the federal statistics is also guaranteed. A hierarchical system ACASS (Accident Causation Analysis with Seven Steps) was developed in GIDAS, describing the human causation factors in a chronological sequence. The accordingly classified causation factors - derived from the systematic of the analysis of human accident causes ("7 steps") - can be used to describe the influence of accident causes on the injury outcome. The bases of the study are accident documentations over ten years from 1999 to 2008 with 8204 vulnerable road users (VRU), of which 3 different groups were selected as pedestrians n=2041, motorcyclists n=2199 and bicyclists n=3964, and analyzed on collisions with cars and trucks as well as vulnerable road users alone. The paper will give a description of the injury pattern and injury mechanisms of accidents. The injury frequencies and severities are pointed out considering different types of VRU and protective measures of helmet and clothes of the human body. The impact points are demonstrated on the car, following to conclusion of protective measures on the vehicle. Existing standards of protection devices as well as interdisciplinary research, including accident and injury statistics, are described. With this paper, a summarization of the existing possibilities on protective measures for pedestrians, bicyclists and motorcyclists is given and discussed by comparison of all three groups of vulnerable road users. Also the relevance of special impact situations and accident causes mainly responsible for severe injuries are pointed out, given the new orientation of research for the avoidance and reduction of accident patterns. 2010 Elsevier Ltd. All rights reserved.

  1. The need to study of bounding accident in reprocessing plant

    International Nuclear Information System (INIS)

    Segawa, Satoshi; Fujita, Kunio

    2013-01-01

    There is a clear consensus that the severe accident corresponds to the core damage accident for power reactors. On the other hand, for FCFs, there is no clear consensus on what is the accident to assess the safety in the region of beyond design basis, or what is the accident which has very low probability but large consequence. The need to examine a bounding consequence of each type of accident is explained to advance the rationality of safety management and regulation and, as a result, to reinforce the safety of a reprocessing plant. The likelihood of occurrence of an accident causing a bounding consequence should correspond to that of a severe accident at a nuclear power plant. The bounding consequence will be derived using the deterministic method and sound engineering judgment supplemented by the probabilistic method. Once an agreement on such a concept is reached among regulators, operators and related experts it will help to provide a solid basis to ensure the safety of a reprocessing plant independent of that of a nuclear power plant. In this paper, we show a preliminary risk profile of RRP calculated by QSA (Quantitative Safety Assessment) which JNFL developed. The profile shows that bounding consequences of various accidents in a range of occurrence frequency corresponding to a severe accident at a nuclear power plant. And we find that the bounding consequence of high-level liquid waste boiling is the largest among all in this range. Therefore, the risk of this event is shown in this paper as an example. To build a common consensus about bounding accidents among concerned parties will encourage regulatory body to introduce such an idea for more effective regulation with scientific rationality. Additionally the study of bounding accidents can contribute to substantial development for accident management strategy as reprocessing operators. (authors)

  2. 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

  3. Method of assessing severe accident management strategies

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.; Jae, M.; Lim, H.; Milici, T.; Park, H.; Swider, J.; Xing, L.; Yu, D.

    1991-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems, and actions to prevent or mitigate a severe accident. A significant number of probabilistic safety assessments (PSAs) have been completed that yield the principal plant vulnerabilities. These vulnerabilities can be categorized as (1) dominant sequences with respect to core-melt frequency. (2) dominant sequences with respect to various risk measures. (3) dominant threats that challenge safety functions. (4) dominant threats with respect to failure of safety systems. 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 considerations include uncertainties in key phenomena, operator behavior, system availability and behavior, and available information. This paper presents a methodology for assessing severe accident management strategies given the key uncertainties delineated at two workshops held at the University of California, Los Angeles. Based on decision trees and influence diagrams, the methodology is currently being applied to two case studies: cavity flooding in a pressurized water reactor (PWR) to prevent vessel penetration or failure, and drywell flooding in a boiling water reactor to prevent vessel and/or containment failure

  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. 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

  6. 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

  7. 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

  8. 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

  9. 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).

  10. 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)

  11. Medical aid in the initial period of radiation accident

    International Nuclear Information System (INIS)

    Selidovkin, G.D.

    1995-01-01

    The main tasks of medical arrangements on the initial stage of rendering aid after radiation accident are the prime medical classification of the injured persons among the personnel of the plant and population, and realization of measures to avoid the increase of doses. The volume of medical aid depends on the type of accident, on the after-accident radiation situation, on the influence of hazardous factors, on the number of people involved in accident situation and the spectrum of sanitary losses, etc., which is to be predicted in advance and to be taken into consideration when rendering aid. The proper and sufficient aid on the initial stage will build the foundation of the ultimate efficiency of medical aid after radiation accident. 14 refs

  12. System 80+ design features for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

    1993-01-01

    ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

  13. Accident analysis in research reactors

    International Nuclear Information System (INIS)

    Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

    2007-01-01

    With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)

  14. Plutonium accident resistant container project

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1978-09-01

    The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC (Nuclear Regulatory Commission). This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO 2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonably small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards

  15. Plutonium accident resistant container project

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1978-05-01

    The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC. This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonalby small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards

  16. The DOE technology development programme on severe accident management

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Moore, R.A.; Theofanous, T.G.

    1998-01-01

    The US Department of Energy (DOE) is sponsoring a programme in technology development aimed at resolving the technical issues in severe accident management strategies for advanced and evolutionary light water reactors (LWRs). The key objective of this effort is to achieve a robust defense-in-depth at the interface between prevention and mitigation of severe accidents. The approach taken towards this goal is based on the Risk Oriented Accident Analysis Methodology (ROAAM). Applications of ROAAM to the severe accident management strategy for the US AP600 advanced LWR have been effective both in enhancing the design and in achieving acceptance of the conclusions and base technology developed in the course of the work. This paper presents an overview of that effort and its key technical elements

  17. School accidents in Austria.

    Science.gov (United States)

    Schalamon, Johannes; Eberl, Robert; Ainoedhofer, Herwig; Singer, Georg; Spitzer, Peter; Mayr, Johannes; Schober, Peter H; Hoellwarth, Michael E

    2007-09-01

    The aim of this study was to obtain information about the mechanisms and types of injuries in school in Austria. Children between 0 and 18 years of age presenting with injuries at the trauma outpatient in the Department of Pediatric Surgery in Graz and six participating hospitals in Austria were evaluated over a 2-year prospective survey. A total of 28,983 pediatric trauma cases were registered. Personal data, site of the accident, circumstances and mechanisms of accident and the related diagnosis were evaluated. At the Department of Pediatric Surgery in Graz 21,582 questionnaires were completed, out of which 2,148 children had school accidents (10%). The remaining 7,401 questionnaires from peripheral hospitals included 890 school accidents (12%). The male/female ratio was 3:2. In general, sport injuries were a predominant cause of severe trauma (42% severe injuries), compared with other activities in and outside of the school building (26% severe injuries). Injuries during ball-sports contributed to 44% of severe injuries. The upper extremity was most frequently injured (34%), followed by lower extremity (32%), head and neck area (26%) and injuries to thorax and abdomen (8%). Half of all school related injuries occur in children between 10 and 13 years of age. There are typical gender related mechanisms of accident: Boys get frequently injured during soccer, violence, and collisions in and outside of the school building and during craft work. Girls have the highest risk of injuries at ball sports other than soccer.

  18. Core loss during a severe accident (COLOSS)

    International Nuclear Information System (INIS)

    Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

    2005-01-01

    The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

  19. 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

  20. Severe accident analysis and management in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

    Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

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

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

  2. 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

  3. 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.

  4. 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.

  5. Radiation accidents and dosimetry

    International Nuclear Information System (INIS)

    Sagstuen, E.; Theisen, H.; Henriksten, T.

    1982-12-01

    On September 2nd 1982 one of the employees of the gamma-irradiation facility at Institute for Energy Technology, Kjeller, Norway entered the irradiation cell with a 65.7 kCi *sp60*Co- source in unshielded position. The victim received an unknown radiation dose and died after 13 days. Using electron spin resonance spectroscopy, the radiation dose in this accident was subsequently determined based on the production of longlived free radicals in nitroglycerol tablets borne by the operator during the accident. He used nitroglycerol for heart problems and free radical are easily formed and trapped in sugar which is the main component of the tablets. Calibration experiments were carried out and the dose given to the tablets during the accident was determined to 37.2 +- 0.5 Gy. The general use of free radicals for dose determinations is discussed. (Auth.)

  6. Big nuclear accidents

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    Much of the debate on the safety of nuclear power focuses on the large number of fatalities that could, in theory, be caused by extremely unlikely but imaginable reactor accidents. This, along with the nuclear industry's inappropriate use of vocabulary during public debate, has given the general public a distorted impression of the safety of nuclear power. The way in which the probability and consequences of big nuclear accidents have been presented in the past is reviewed and recommendations for the future are made including the presentation of the long-term consequences of such accidents in terms of 'reduction in life expectancy', 'increased chance of fatal cancer' and the equivalent pattern of compulsory cigarette smoking. (author)

  7. Care of radiation accidents

    International Nuclear Information System (INIS)

    Renz, K.

    1983-01-01

    The small probability of a serious radiation accident happening dispenses neither the plants where radiation exposure occurs nor the employers' liability insurance associations from their obligation to make provision for such cases. On the other hand, the efforts involved in such preventive measures must be kept within reasonable limits. As a result of these considerations a concept for taking care of radiation accidents was developed that is based on already existing institutions. The most attention was demanded by questions of organization, logistics, communication and information. The syndrome appearing after acute whole-body irradiation is known. This syndrome in its different stages and the relative therapeutic measures form the basis for the organization of the care of radiation accidents. (orig./MG) [de

  8. Advanced DVR with Zero-Sequence Voltage Component and Voltage Harmonic Elimination for Three-Phase Three-Wire Distribution Systems

    Directory of Open Access Journals (Sweden)

    Margo P

    2009-11-01

    Full Text Available Dynamic Voltage Restorer (DVR is a power electronics device to protect sensitive load when voltage sag occurs. Commonly, sensitive loads are electronic-based devices which generate harmonics. The magnitude and phase of compensated voltage in DVR depend on grounding system and type of fault. If the system is floating, the zero sequence components do not appear on the load side. Meanwhile, in a neutral grounded system, voltage sag is extremely affected by zero sequence components. A blocking transformer is commonly installed in series with DVR to reduce the effect of zero sequence components. This paper proposes a new DVR control scheme that is capable of eliminating the blocking transformer and reducing harmonic distortion. The system uses fuzzy polar controller to replace the conventional PI or FL controller that is commonly used. By taking into account the zero sequence components in the controller design, the effects of zero sequence components can be compensated. Simulated results show the effectiveness of the proposed DVR controller

  9. Review of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Connelly, J.W.; Storr, G.J.

    1989-01-01

    Two types of severe reactor accidents - loss of coolant or coolant flow and transient overpower (TOP) accidents - are described and compared. Accidents in research reactors are discussed. The 1961 SL1 accident in the US is used as an illustration as it incorporates the three features usually combined in a severe accident - a design flaw or flaws in the system, a circumvention of safety circuits or procedures, and gross operator error. The SL1 reactor, the reactivity accident and the following fuel-coolant interaction and steam explosion are reviewed. 3 figs

  10. Severe Accident Test Station Design Document

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  11. Severe Accident Test Station Design Document

    International Nuclear Information System (INIS)

    Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

    2015-01-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  12. Criticality accident alarm system

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

    The American National Standard ANSI/ANS-8.3-1986, Criticality Accident Alarm System provides guidance for the establishment and maintenance of an alarm system to initiate personnel evacuation in the event of inadvertent criticality. In addition to identifying the physical features of the components of the system, the characteristics of accidents of concern are carefully delineated. Unfortunately, this ANSI Standard has led to considerable confusion in interpretation, and there is evidence that the ''minimum accident of concern'' may not be appropriate. Furthermore, although intended as a guide, the provisions of the standard are being rigorously applied, sometimes with interpretations that are not consistent. Although the standard is clear in the use of absorbed dose in free air of 20 rad, at least one installation has interpreted the requirement to apply to dose in soft tissue. The standard is also clear in specifying the response to both neutrons and gamma rays. An assembly of uranyl fluoride enriched to 5% 235 U was operated to simulate a potential accident. The dose, delivered in a free run excursion 2 m from the surface of the vessel, was greater than 500 rad, without ever exceeding a rate of 20 rad/min, which is the set point for activating an alarm that meets the standard. The presence of an alarm system would not have prevented any of the five major accidents in chemical operations nor is it absolutely certain that the alarms were solely responsible for reducing personnel exposures following the accident. Nevertheless, criticality alarm systems are now the subject of great effort and expense. 13 refs

  13. Accident at Harrisburg

    International Nuclear Information System (INIS)

    1979-05-01

    The course of events during the accident on 28 March 1979 at Three Mile Island-2 Reactor at Harrisburg, Pennsylvania, is described in detail. The effects (in the environment and within the safety containment) are described. The following points are then discussed: the possibility of a comparable accident occurring in the nuclear power stations in the German Federal Republic; the possibility of any point having been overlooked in the design of nuclear power stations in the Federal Republic; whether previous risk analyses are still valid; and how near the Three Mile Island reactor was to a core meltdown. Some conclusions are drawn. (U.K.)

  14. Mortal radiological accident

    International Nuclear Information System (INIS)

    Gimenez, J.C.

    1987-01-01

    After defining the concept of 'Radiological accident', statistical data from Radiation Emergency Assistance Center of ORNL (United States of America) are given about the deaths caused by acute irradiation between 1944 and April 24, 1986 -ie, the day before Chernobyl nuclear accident- as well as on the number of deaths caused by the latter. Next the different clinical stages of the Acute Irradiation Syndrome (AIS) as well as its possible treatment are described, and finally the different physical, clinical and biological characteristics linked to the AIS and to its diagnosis and prognosis are discussed. (M.E.L.) [es

  15. The TMI-2 accident

    International Nuclear Information System (INIS)

    Loureiro, L.A.

    1986-01-01

    A critical study about the technical and man-related facts in order to establish what is considered the worst commercial nuclear power accident until 1986. Radiological consequences and stress to the public are considered in contrast to antinuclear groups. This descriptive and technical study has the purpose to document written and oral opinions obtained abroad and then explain to the public in an easy language terminology. Preliminary study describing safety related systems fails and the accident itself with minute to minute description, conduct to the consequences and then, to learned lessons

  16. Increase the Safety of Road Traffic Accidents by Applying Clustering

    Directory of Open Access Journals (Sweden)

    Kos Goran

    2013-12-01

    Full Text Available In terms of continual increase of number of traffic accidents and alarming trend of increasing number of traffic accidents with catastrophic consequences for human life and health, it is necessary to actively research and develop methods to combat these trends. One of the measures is the implementation of advanced information systems in existing traffic environment. Accidents clusters, as databases of traffic accidents, introduce a new dimension in traffic systems in the form of experience, providing information on current accidents and the ones that have previously occurred in a given period. This paper proposes a new approach to predictive management of traffic processes, based on the collection of data in real time and is based on accidents clusters. The modern traffic information services collects road traffic status data from a wide variety of traffic sensing systems using modern ICT technologies, creating the most accurate road traffic situation awareness achieved so far. Road traffic situation awareness enhanced by accident clusters' data can be visualized and distributed in various ways (including the forms of dynamic heat maps and on various information platforms, suiting the requirements of the end-users. Accent is placed on their significant features that are based on additional knowledge about existing traffic processes and distribution of important traffic information in order to prevent and reduce traffic accidents.

  17. A Study on the Operation Strategy for Combined Accident including TLOFW accident

    International Nuclear Information System (INIS)

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

    2014-01-01

    It is difficult for operators to recognize the necessity of a feed-and-bleed (F-B) operation when the loss of coolant accident and failure of secondary side occur. An F-B operation directly cools down the reactor coolant system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. The plant is not always necessary the F-B operation when the secondary side is failed. It is not necessary to initiate an F-B operation in the case of a medium or large break because these cases correspond to low RCS pressure sequences when the secondary side is failed. If the break size is too small to sufficiently decrease the RCS pressure, the F-B operation is necessary. Therefore, in the case of a combined accident including a secondary cooling system failure, the provision of clear information will play a critical role in the operators' decision to initiate an F-B operation. This study focuses on the how we establish the operation strategy for combined accident including the failure of secondary side in consideration of plant and operating conditions. Previous studies have usually focused on accidents involving a TLOFW accident. The plant conditions to make the operators confused seriously are usually the combined accident because the ORP only focuses on a single accident and FRP is less familiar with operators. The relationship between CET and PCT under various plant conditions is important to decide the limitation of initiating the F-B operation to prevent core damage

  18. Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Albrecht, H.; Malinauskas, A.P.

    1978-01-01

    A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

  19. [Drugs and occupational accident].

    Science.gov (United States)

    Bratzke, H; Albers, C

    1996-02-01

    In a case of a fatal occupational accident (construction worker, fall from roof, urine test positive for cocaine and THC, e.g. cannabis) the question arised to what extent those drug-related occupational accidents occur. In the literature only few cases, mainly dealing with cannabis influence, have been reported, however, a higher number is suspected. Cocaine and other stimulating drugs (amphetamine) are more often used to increase physical fitness. By direct or indirect interference with vigilance these compounds may provoke accidents. Due to the lack of a legal basis proving of the influence of drugs at the working place is still very limited, although highly sensitive chemical-toxicological assay procedures are available to detect even the chronic abuse (in hair). In the general conditions of accident insurances a compensation is excluded when alcohol is involved, but drugs are not mentioned. It is indeed difficult to establish a concentration limit for drugs like that existing for alcohol (1.1%). In each case the assay of the drug involved and exact knowledge of its specific effects is in an essential prerequisite to prove the causal relationship.

  20. Note nuclear accidents combat

    International Nuclear Information System (INIS)

    1989-01-01

    In this document the starting points are described which underlie the new framework for the nuclear-accident combat in the Netherlands. All the elaboration of this is indicated in main lines. The juridical consequences of the proposed structure are enlightened and the sequel activities are indicated. (H.W.). 6 figs.; 8 tabs

  1. Measures against nuclear accidents

    International Nuclear Information System (INIS)

    1992-01-01

    A select committee appointed by the Norwegian Ministry of Social Affairs put forward proposals concerning measures for the improvement of radiation protection preparedness in Norway. On the basis on an assessment of the potential radiation accident threat, the report examines the process of response, and identifies the organizational and management factors that influence that process

  2. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-01-01

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP) [de

  3. Lessons learned from accidents investigations

    Energy Technology Data Exchange (ETDEWEB)

    Zuniga-Bello, P. [Consejo Nacional de Ciencia y Tecnologia (CONACYT), Mexico City (Mexico); Croft, J. [National Radiological Protection Board (United Kingdom); Glenn, J

    1997-12-31

    Accidents from three main practices: medical applications, industrial radiography and industrial irradiators are used to illustrate some common causes of accidents and the main lessons to be learned. A brief description of some of these accidents is given. Lessons learned from the described accidents are approached by subjects covering: safety culture, quality assurance, human factors, good engineering practice, defence in depth, security of sources, safety assessment and monitoring and verification compliance. (author)

  4. 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

  5. Lessons learned from accident investigations

    International Nuclear Information System (INIS)

    Zuniga-Bello, P.; Croft, J.R.; Glenn, J.

    1998-01-01

    Accidents in three main practices - medical applications, industrial radiography and industrial irradiators - are used to illustrate some common causes of accidents and the main lessons to be learned from them. A brief description of some of these accidents is given. Lessons learned from the accidents described are approached bearing in mind: safety culture, quality assurance, human factors, good engineering practice, defence in depth, security of sources, safety assessment and monitoring and verification compliance. (author)

  6. ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

    International Nuclear Information System (INIS)

    Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

    2002-01-01

    The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

  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. Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

    Energy Technology Data Exchange (ETDEWEB)

    Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others

    1997-07-01

    In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

  9. EPRI research on accident management

    International Nuclear Information System (INIS)

    Oehlberg, R.N.; Chao, J.

    1991-01-01

    The paper discusses Nuclear Regulatory Commission (NRC) efforts regarding severe reactor accident management and the Nuclear Management and Resources Council (NUMAEX), activities. (EPRI) Electric Power Research Institute accident management program consists of the two products just mentioned plus one related to severe accident plant status information and the MAAP 4.0 computer code. These are briefly discussed

  10. Accident management on french PWRS

    International Nuclear Information System (INIS)

    Queniart, D.

    1990-06-01

    After a brief recall of French safety rationale, the reactor operation and severe accident management is given. The research and development aimed at developing accident management procedures and emergency organization in France for the case of a NPP accident are also given

  11. 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)

  12. 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

  13. Sequence-dependent toxicity profile in modified FAMTX (fluorouracil-adriamycin-methotrexate) chemotherapy with lenograstim support for advanced gastric cancer: a feasibility study

    NARCIS (Netherlands)

    Westermann, A. M.; Taal, B. G.; Swart, M.; Boot, H.; Craanen, M.; Gerritsen, W. R.

    2000-01-01

    For advanced irresectible gastric cancer, sequential high-dose methotrexate and 5-fluorouracil (both on day 1) combined with adriamycin on day 15 (FAMTX regimen), cycled every 28 days, is a fairly effective but toxic treatment, with a high incidence of neutropenic fever, dose reductions and dose

  14. 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.

  15. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-07-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The accident sequence chosen for the calculating was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D was chosen with a wind speed of 5 m/s, and as extreme weather, Pasquill F with a wind speed of 2 m/s. 23 tabs., 37 ills., 20 refs. (author)

  16. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-01-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Ringhals reactor No 3/4. The accident sequence chosen for the calcualtions was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. A decontamination factor of 500 is used to account for the scrubber effect. Meteorological data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D, was chosen with a wind speed of 10 m/s, and as extreme weather, Pasquill E, with a wind speed of 2 m/s. 19 refs. (author)

  17. IG and TR single chain fragment variable (scFv) sequence analysis: a new advanced functionality of IMGT/V-QUEST and IMGT/HighV-QUEST.

    Science.gov (United States)

    Giudicelli, Véronique; Duroux, Patrice; Kossida, Sofia; Lefranc, Marie-Paule

    2017-06-26

    IMGT®, the international ImMunoGeneTics information system® ( http://www.imgt.org ), was created in 1989 in Montpellier, France (CNRS and Montpellier University) to manage the huge and complex diversity of the antigen receptors, and is at the origin of immunoinformatics, a science at the interface between immunogenetics and bioinformatics. Immunoglobulins (IG) or antibodies and T cell receptors (TR) are managed and described in the IMGT® databases and tools at the level of receptor, chain and domain. The analysis of the IG and TR variable (V) domain rearranged nucleotide sequences is performed by IMGT/V-QUEST (online since 1997, 50 sequences per batch) and, for next generation sequencing (NGS), by IMGT/HighV-QUEST, the high throughput version of IMGT/V-QUEST (portal begun in 2010, 500,000 sequences per batch). In vitro combinatorial libraries of engineered antibody single chain Fragment variable (scFv) which mimic the in vivo natural diversity of the immune adaptive responses are extensively screened for the discovery of novel antigen binding specificities. However the analysis of NGS full length scFv (~850 bp) represents a challenge as they contain two V domains connected by a linker and there is no tool for the analysis of two V domains in a single chain. The functionality "Analyis of single chain Fragment variable (scFv)" has been implemented in IMGT/V-QUEST and, for NGS, in IMGT/HighV-QUEST for the analysis of the two V domains of IG and TR scFv. It proceeds in five steps: search for a first closest V-REGION, full characterization of the first V-(D)-J-REGION, then search for a second V-REGION and full characterization of the second V-(D)-J-REGION, and finally linker delimitation. For each sequence or NGS read, positions of the 5'V-DOMAIN, linker and 3'V-DOMAIN in the scFv are provided in the 'V-orientated' sense. Each V-DOMAIN is fully characterized (gene identification, sequence description, junction analysis, characterization of mutations and amino

  18. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

    2010-01-01

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  19. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  20. Cirrus Airframe Parachute System and Odds of a Fatal Accident in Cirrus Aircraft Crashes.

    Science.gov (United States)

    Alaziz, Mustafa; Stolfi, Adrienne; Olson, Dean M

    2017-06-01

    General aviation (GA) accidents have continued to demonstrate high fatality rates. Recently, ballistic parachute recovery systems (BPRS) have been introduced as a safety feature in some GA aircraft. This study evaluates the effectiveness and associated factors of the Cirrus Airframe Parachute System (CAPS) at reducing the odds of a fatal accident in Cirrus aircraft crashes. Publicly available Cirrus aircraft crash reports were obtained from the National Transportation Safety Board (NTSB) database for the period of January 1, 2001-December 31, 2016. Accident metrics were evaluated through univariate and multivariate analyses regarding odds of a fatal accident and use of the parachute system. Included in the study were 268 accidents. For CAPS nondeployed accidents, 82 of 211 (38.9%) were fatal as compared to 8 of 57 (14.0%) for CAPS deployed accidents. After controlling for all other factors, the adjusted odds ratio for a fatal accident when CAPS was not deployed was 13.1. The substantial increased odds of a fatal accident when CAPS was not deployed demonstrated the effectiveness of CAPS at providing protection of occupants during an accident. Injuries were shifted from fatal to serious or minor with the use of CAPS and postcrash fires were significantly reduced. These results suggest that BPRS could play a significant role in the next major advance in improving GA accident survival.Alaziz M, Stolfi A, Olson DM. Cirrus Airframe Parachute System and odds of a fatal accident in Cirrus aircraft crashes. Aerosp Med Hum Perform. 2017; 88(6):556-564.

  1. On preparation for accident management in LWR power stations

    International Nuclear Information System (INIS)

    1996-01-01

    Nuclear Safety Commission received the report from Reactor Safety General Examination Committee which investigated the policy of executing the preparation for accident management. The basic policy on the preparation for accident management was decided by Nuclear Safety Commission in May, 1992. This Examination Committee investigated the policy of executing the preparation for accident management, which had been reported from the administrative office, and as the result, it judged the policy as adequate, therefore, the report is made. The course to the foundation of subcommittee is reported. The basic policy of the examination on accident management by the subcommittee conforming to the decision by Nuclear Safety Commission, the measures of accident management which were extracted for BWR and PWR facilities, the examination of the technical adequacy of selecting accident sequences in BWR and PWR facilities and the countermeasures to them, the adequacy of the evaluation of the possibility of executing accident management measures and their effectiveness and the adequacy of the evaluation of effect to existing safety functions, the preparation of operation procedure manual, and education and training plan are reported. (K.I.)

  2. Severe accidents and terrorist threats at nuclear reactors

    International Nuclear Information System (INIS)

    Pollack, G.L.

    1987-01-01

    Some of the key areas of uncertainty are the nature of the physical and chemical interactions of released fission products and of the interactions between a molten core and concrete, the completeness and validity of the computer codes used to predict accidents, and the behavior of the containment. Because of these and other uncertainties, it is not yet possible to reliably predict the consequences of reactor accidents. It is known that for many accident scenarios, especially less severe ones or where the containment is not seriously compromised, the amount of radioactive material expected to escape the reactor is less, even much less, than was previously calculated. For such accidents, the predictions are easier and more reliable. With severe accidents, however, there is considerable uncertainty as to the predicted results. For accidents of the type that terrorists might cause - for example, where the sequence of failure would be unexpected or where redundant safety features are caused to fail together - the uncertainties are still larger. The conclusion, then, is that there are potential dangers to the public from terrorist actions at a nuclear reactor; however, because of the variety of potential terrorist threats and the incompleteness of the knowledge about the behavior of reactor components and fission products during accidents, the consequences cannot yet be assessed quantitatively

  3. 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

  4. Advanced calculus

    CERN Document Server

    Friedman, Avner

    2007-01-01

    This rigorous two-part treatment advances from functions of one variable to those of several variables. Intended for students who have already completed a one-year course in elementary calculus, it defers the introduction of functions of several variables for as long as possible, and adds clarity and simplicity by avoiding a mixture of heuristic and rigorous arguments.The first part explores functions of one variable, including numbers and sequences, continuous functions, differentiable functions, integration, and sequences and series of functions. The second part examines functions of several

  5. Casebook on electric safety accidents

    International Nuclear Information System (INIS)

    1987-09-01

    This book gives concentration on electric safety accidents in domestic and abroad, which introduces general electrical safety with property of electricity, safe equipment and maintenance and protection of electric shock. It lists the cases of accident caused of electricity in domestic like accident in power substation, utilization equipment, load system and another accident by electricity like death in electric shock another by electricity like death in electric shock in new building construction, the cases caused of electricity in abroad like damage in electric shock by high voltage electric transformer, electric shock in summer and earth fault accident by fault cooling tower.

  6. Radiological accidents balance in medicine

    International Nuclear Information System (INIS)

    Nenot, J.C.

    1995-01-01

    This work deals with the radiological accidents in medicine. In medicine, the radiation accidents on medical personnel and patients can be the result of over dosage and bad focusing of radiotherapy sealed sources. Sometimes, the accidents, if they are unknown during a time enough for the source to be spread and to expose a lot of persons (in the case of source dismantling for instance) can take considerable dimensions. Others accidents can come from bad handling of linear accelerators and from radionuclide kinetics in some therapies. Some examples of accidents are given. (O.L.). 11 refs

  7. Small break LOCA [loss of coolant accident] mitigation for Bellefonte

    International Nuclear Information System (INIS)

    Bayless, P.D.; Dobbe, C.A.

    1986-01-01

    Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

  8. The characters of emergency rescue and the measures to prevent accidents for nuclear-powered submarine

    International Nuclear Information System (INIS)

    Wang Yuexing

    1999-01-01

    The characteristics of emergency rescue and the measures for preventing and decreasing accidents in nuclear-powered submarine have been presented. The breakdown of equipment and human factors are the main reasons which lead to accidents. Four preventive measures are suggested: enhancing capabilities to take precautions against fire, seriously controlling the environmental factors which affect the health of the submariners, reinforcing the constitutions of the submariners, and working out emergency planning against serious accidents in advance

  9. Three Mile Island accident

    International Nuclear Information System (INIS)

    Barre, B.; Olivier, E.; Roux, J.P.; Pelle, P.

    2010-01-01

    Deluded by equivocal instrumentation signals, operators at TMI-2 (Three Mile Island - unit 2) misunderstood what was going on in the reactor and for 2 hours were taking inadequate decisions that turned a reactor incident into a major nuclear event that led to the melting of about one third of the core. The TMI accident had worldwide impacts in the domain of nuclear safety. The main consequences in France were: 1) the introduction of the major accident approach and the reinforcement of crisis management; 2) the improvement of the reactor design, particularly that of the pressurizer valves; 3) the implementation of safety probabilistic studies; 4) a better taking into account of the feedback experience in reactor operations; and 5) a better taking into account of the humane factor in reactor safety. (A.C.)

  10. The ultimate nuclear accident

    International Nuclear Information System (INIS)

    Abdus Salam, A.

    1988-01-01

    The estimated energy equivalent of Chernobyl explosion was the 1/150 th of the explosive energy equivalent of atomic bomb dropped on Hiroshima; while the devastation that could be caused by the world's stock pile of nuclear weapons, could be equivalent to 160 millions of Chernobyl-like incidents. As known, the number of nuclear weapons is over 50,000 and 2000 nuclear weapons are sufficient to destroy the world. The Three Mile Island and Chernobyl accidents have been blamed on human factors but also the human element, particularly in the form of psychological stresses on those operating the nuclear weapons, could accidentally bring the world to a nuclear catastrophe. This opinion is encouraged by the London's Sunday Times magazine which gave a graphic description of life inside a nuclear submarine. So, to speak of nuclear reactor accidents and not of nuclear weapons is false security. (author)

  11. 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)

  12. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  13. Severe accident management guidelines

    International Nuclear Information System (INIS)

    Uhle, Jennifer

    2014-01-01

    The events at Fukushima Daiichi have highlighted the importance of Severe Accident Management Guidelines (SAMGs). As the world has learned from the catastrophe and countries are considering changes to their nuclear regulatory programs, the content of SAMGs and their regulatory control are being evaluated. This presentation highlights several factors that are being addressed in the United States as rulemaking is underway pertaining to SAMGs. The question of how to be prepared for the unexpected is discussed with specific insights gleaned from Fukushima. (author)

  14. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Rassow, J.

    1986-01-01

    The documentation aims at giving a clearly arranged account of facts, interrelations and comparative evaluations of general interest. It deals with the course of events, atmospheric dispersion and fallout of the substances released and discusses the basic principles of the metering of radioactive radiation, the calculation of body doses and comparative evaluations with the radioactive exposure and risks involved by other sources. The author intends to contribute to an objective discussion about the Chernobyl reactor accident and nuclear energy as such. (DG) [de

  15. Radiation accident in Vietnam

    International Nuclear Information System (INIS)

    Wheatley, J.

    1994-01-01

    In November 1992 a Vietnamese research physicist was working with a microtron accelerator when he received a radiation overexposure that required the subsequent amputation of his right hand. A team from the International Atomic Energy Agency visited Hanoi in March 1993 to carry out an investigation. It was concluded that the accident occurred primarily because of a lack of safety systems, although the lack of both written procedures and training in basic radiation safety were also major contributors. (author)

  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. PREVENTION OF OCCUPATIONAL ACCIDENTS

    Directory of Open Access Journals (Sweden)

    Jovica Jovanovic

    2004-01-01

    Full Text Available Medical services, physicians and nurses play an essential role in the plant safety program through primary treatment of injured workers and by helping to identify workplace hazards. The physician and nurse should participate in the worksite investigations to identify specific hazard or stresses potentially causing the occupational accidents and injuries and in planning the subsequent hazard control program. Physicians and nurses must work closely and cooperatively with supervisors to ensure the prompt reporting and treatment of all work related health and safety problems. Occupational accidents, work related injuries and fatalities result from multiple causes, affect different segments of the working population, and occur in a myriad of occupations and industrial settings. Multiple factors and risks contribute to traumatic injuries, such as hazardous exposures, workplace and process design, work organization and environment, economics, and other social factors. With such a diversity of theories, it will not be difficult to understand that there does not exist one single theory that is considered right or correct and is universally accepted. These theories are nonetheless necessary, but not sufficient, for developing a frame of reference for understanding accident occurrences. Prevention strategies are also varied, and multiple strategies may be applicable to many settings, including engineering controls, protective equipment and technologies, management commitment to and investment in safety, regulatory controls, and education and training. Research needs are thus broad, and the development and application of interventions involve many disciplines and organizations.

  18. Radiological accident of Goiania

    International Nuclear Information System (INIS)

    Palacios, Elias; Gimenez, J.C.

    1988-01-01

    The accident of Goiania that took place in September, 1987, was the consequence of the dismantlement of the teletherapy equipment containing a Cs 137 source. The activity of the source was of about 5,2.10 13 Bq(1.400 Ci) and was made up by 1.10 -1 Kg of ClCs. This is one of the worst accidents, involving medical or industrial source, which happened up to the moment. The accident and the criteria adopted to face the emergency are described. The characteristics of the irradiation and superficial and internal contamination of the persons affected caused any dosimetric evaluation to be particularly difficult. The emergency control managed by capable persons was carried out with the expected difficulties in some areas within the Goiania city. About 90 % of the material could be recovered, in the removal works, in order to be treated as radioactive waste. Conclusions that should be taken into account in Argentina are reached. The authors of the article took part in the emergency by helping the brazilian Authorities within the Mutual Assistance Agreement between Argentina and Braxil. (M.E.L.) [es

  19. Thule accident 1968

    International Nuclear Information System (INIS)

    Melgaard, L.; Moeller Kristensen, H.

    1987-01-01

    On January 21, 1968 an American B-52 bomber crashed on the ice at Thule in Nothern Greenland. The bomber carried 4 nuclear weapons that were destroyed. The radioactive material of the bombs was spread over a large area of the ice. About 850 Danes stayed at the Thule base in 1968 for a shorter or longer period. Out of these 850 probably between 70 and 170 men took part in the clearing after the accident. Danish and American authorities establised that the radioactive contamination from the accident was too small to cause any health effects. For that reason the Danish authorities did not follow the Danish workers in order to show late effects, if any. In defiance of the authorities' very cocksure attitude towards possible late effets parts of the Danish press in the Summer 1986 started to be interested in the matter and to search for previous Thule-workers. Up till January 1987 aboput 600 workers have been contacted by the press, trade unions, and private persons. About 500 out of the 600 workers report on illness, cancers and deaths. This report tries to compile the accessible informations on the matter, to descibe the possible radioactive and chemical effects, to compare the Thule accident with a similar incident in Spain in 1966, and to propose a comprehensive health examination of all the workers. (LN)

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

    International Nuclear Information System (INIS)

    Butragueno, J.L.

    1980-01-01

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

  1. Accidents and human factors

    International Nuclear Information System (INIS)

    Nishiwaki, Y.; Kawai, H.; Morishima, H.; Terano, T.; Sugeno, M.

    1984-01-01

    When the TMI accident occurred it was 4 a.m., an hour when the error potential of the operators would have been very high. The frequency of car and train accidents in Japan is also highest between 4 a.m. and 6 a.m. The error potential may be classified into five phases corresponding to the electroencephalogramic pattern (EEG). At phase 0, when the delta wave appears, a person is unconscious and in deep sleep; at phase I, when the theta wave appears, he is very tired, sleepy and subnormal; at phase II, when the alpha wave appears, he is normal, relaxed and passive; at phase III, when the beta wave appears, he is normal, clear-minded and active; at phase IV, when the strong beta or epileptic wave appears, he is hypernormal, excited and incapable of normal judgement. Should an accident occur at phase II, the brain condition may jump to phase IV. At this phase the error or accident potential is maximum. The response of the human brain to different types of noises and signals may vary somewhat for different individuals and for different groups of people. Therefore, the possibility that such differences in brain functions may influence the mental structure would be worthy of consideration in human factors and in the design of man-machine systems. Human reliability and performance would be affected by many factors: medical, physiological and psychological, etc. The uncertainty involved in human factors may not necessarily be probabilistic, but fuzzy. Therefore, it would be important to develop a theory by which both non-probabilistic uncertainties, or fuzziness, of human factors and the probabilistic properties of machines can be treated consistently. From the mathematical point of view, probabilistic measure is considered a special case of fuzzy measure. Therefore, fuzzy set theory seems to be an effective tool for analysing man-machine systems. To minimize human error and the possibility of accidents, new safety systems should not only back up man and make up for his

  2. Radiation accident/disaster

    International Nuclear Information System (INIS)

    Kida, Yoshiko; Hirohashi, Nobuyuki; Tanigawa, Koichi

    2013-01-01

    Described are the course of medical measures following Fukushima Daiichi Nuclear Power Plant (FNPP) Accident after the quake and tsunami (Mar. 11, 2011) and the future task for radiation accident/disaster. By the first hydrogen explosion in FNPP (Mar. 12), evacuation of residents within 20 km zone was instructed, and the primary base for measures of nuclear disaster (Off-site Center) 5 km afar from FNPP had to work as a front base because of damage of communicating ways, of saving of injured persons and of elevation of dose. On Mar. 13, the medical arrangement council consisting from stuff of Fukushima Medical University (FMU), National Institute of Radiological Sciences, Nuclear Safety Research Association and Prefectural officers was setup in residents' hall of Fukushima City, and worked for correspondence to persons injured or exposed, where communication about radiation and between related organizations was still poor. The Off-site Center's head section moved to Prefectural Office on Mar. 15 as headquarters. Early in the period, all residents evacuated from the 20 km zone, and in-hospital patients and nursed elderly were transported with vehicles, >50 persons of whom reportedly died mainly by their base diseases. The nation system of medicare for emergent exposure had consisted from the network of the primary to third facilities; there were 5 facilities in the Prefecture, 3 of which were localized at 4-9 km distance from FNPP and closed early after the Accident; and the secondary facility of FMU became responsible to all exposed persons. There was no death of workers of FNPP. Medical stuff also measured the ambient dose at various places near FNPP, having had risk of exposure. At the Accident, the important system of command, control and communication was found fragile and measures hereafter should be planned on assumption of the worst scenario of complete damage of the infrastructure and communication. It is desirable for Disaster Medical Assistance Team which

  3. 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.

  4. Severe accident analysis methodology in support of accident management

    International Nuclear Information System (INIS)

    Boesmans, B.; Auglaire, M.; Snoeck, J.

    1997-01-01

    The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

  5. 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

  6. 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)

  7. 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)

  8. Generic implications of the Chernobyl accident

    International Nuclear Information System (INIS)

    Sege, G.

    1989-01-01

    The US Nuclear Regulatory Commission (NRC) staff's assessment of the generic implications of the Chernobyl accident led to the conclusion that no immediate changes in the NRC's regulations regarding design or operation of US commercial reactors are needed. However, further consideration of certain issues was recommended. This paper discusses those issues and the studies being addressed to them. Although 24 tasks relating to light water reactor issues are identified in the Chernobyl follow-up research program, only four are new initiatives originating from Chernobyl implications. The remainder are limited modifications of ongoing programs designed to ensure that those programs duly reflect any lessons that may be drawn from the Chernobyl experience. The four new study tasks discussed include a study of reactivity transients, to reconfirm or bring into question the adequacy of potential reactivity accident sequences hitherto selected as a basis for design approvals; analysis of risk at low power and shutdown; a study of procedure violations; and a review of current NRC testing requirements for balance of benefits and risks. Also discussed, briefly, are adjustments to ongoing studies in the areas of operational controls, design, containment, emergency planning, and severe accident phenomena

  9. Advanced information society(7)

    Science.gov (United States)

    Chiba, Toshihiro

    Various threats are hiding in advanced informationalized society. As we see car accident problems in motorization society light aspects necessarily accompy shady ones. Under the changing circumstances of advanced informationalization added values of information has become much higher. It causes computer crime, hacker, computer virus to come to the surface. In addition it can be said that infringement of intellectual property and privacy are threats brought by advanced information. Against these threats legal, institutional and insurance measures have been progressed, and newly security industry has been established. However, they are not adequate individually or totally. The future vision should be clarified, and countermeasures according to the visions have to be considered.

  10. 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.)

  11. Occupational accidents aboard merchant ships

    DEFF Research Database (Denmark)

    Hansen, H.L.; Nielsen, D.; Frydenberg, Morten

    2002-01-01

    Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may...... be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were...... aboard. Relative risks for notified accidents and accidents causing permanent disability of 5% or more were calculated in a multivariate analysis including ship type, occupation, age, time on board, change of ship since last employment period, and nationality. Foreigners had a considerably lower recorded...

  12. 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

  13. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  14. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-05-01

    Following the accident at Chernobyl nuclear reactor, WHO organized on 6 May 1986 in Copenhagen a one day consultation of experts with knowledge in the fields of meteorology, radiation protection, biological effects, reactor technology, emergency procedures, public health and psychology in order to analyse the development of events and their consequences and to provide guidance as to the needs for immediate public health action. The present report provides detailed information on the transportation and dispersion of the radioactive material in the atmosphere, especially volatile elements, during the release period 26 April - 5 May. Presented are the calculated directions and locations of the radioactive plume over Europe in the first 5 days after the accident, submitted by the Swedish Meteorological and Hydrological Institute. The calculations have been made for two heights, 1500m and 750m and the plume directions are grouped into five periods, covering five European areas. The consequences of the accident inside the USSR and the radiological consequences outside the USSR are presented including the exposure routes and the biological effects, paying particular attention to iodine-131 effects. Summarized are the first reported measured exposure rates above background, iodine-131 deposition and concentrations in milk and the remedial actions taken in various European countries. Concerning the cesium-137 problem, based on the UNSCEAR assessment of the consequences of the nuclear fallout, one concludes that the cesium contamination outside the USSR is not likely to cause any serious problems. Finally, the conclusions and the recommendations of the meeting, taking into account both the short-term and longer term considerations are presented

  15. Traffic accidents: an econometric investigation

    OpenAIRE

    Tito Moreira; Adolfo Sachsida; Loureiro Paulo

    2004-01-01

    Based on a sample of drivers in Brasilia's streets, this article investigates whether distraction explains traffic accidents. A probit model is estimated to determine the predictive power of several variables on traffic accidents. The main conclusion drawn from this study is that the proxies used to measure distraction, such as the use of cell phones and cigarette smoking in a moving vehicle, are significant factors in determining traffic accidents.

  16. Medical aspects of radiation accidents

    International Nuclear Information System (INIS)

    Messerschmidt, O.

    1990-01-01

    Reactor accidents and nuclear bomb explosions are compared including the release of radioactivity in an accident, results of risk studies, emergency measures of nuclear power plants, and evacuation of the population. The medical aspects refer to the prophylaxies of the thyroid gland, contamination and decontamination of body surfaces, recommendations of the ICRP, radiation injury after total body exposure and medical problems after a reactor accident. (DG)

  17. 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

  18. Nuclear accidents and protection

    International Nuclear Information System (INIS)

    Biocanin, R.; Amidzic, B.

    2005-01-01

    The numerous threats are our cruel reality. There is a great arsenal of nuclear weapons. Nuclear terrorism and nuclear accidents are always possible, especially during the transport and handling different nuclear material. Terrorist organisation also goes for coming into the possession of the nuclear means. Specific and important problem is human radioactive contamination in using nuclear energy for peaceful and military purpose. So, realisation of the universal and united system of NBCD gives us a possibility by using the modern communication equipment and very effective mobile units to react in a real time and successfully perform monitoring, alarming, protection and decontamination. (author) [sr

  19. The accident of Chernobyl

    International Nuclear Information System (INIS)

    1986-10-01

    RBMK reactors (reactor control, protection systems, containment) and the nuclear power plant of Chernobyl are first presented. The scenario of the accident is given with a detailed chronology. The actions and consequences on the site are reviewed. This report then give the results of the source term estimation (fision product release, core inventory, trajectories, meteorological data...), the radioactivity measurements obtained in France. Health consequences for the French population are evoked. The medical consequences for the population who have received a high level of doses are reviewed [fr

  20. Psychological response of accident

    International Nuclear Information System (INIS)

    Novikov, V.S.; Nikiforov, A.M.; Cheprasov, V.Yu.

    1996-01-01

    The psychological status of rescuers of consequences of Chernobyl[s accidents, having planned stationary examination and treatment of common somatic diseases, has been examined. THe age of men represented the study group was 35-54 years old. The results of medical-psychological examination showed the development in rescuers of common dysadaptation and stress state, characterized by depressive-hypochondriac state with high anxiety. The course of psychotherapeutic activities made possible to improve essentionally the psychological status of the patients. 12 refs., 3 figs., 1 tab

  1. Reactor accident in Chernobyl

    International Nuclear Information System (INIS)

    Sokalski, A.; Kowalski, A.

    1990-11-01

    The bibliography contains 1568 descriptions of papers devoted to Chernobylsk accident and recorded in ''INIS Atomindex'' to 30 June 1990. The descriptions were taken from ''INIS Atomindex'' and are presented in accordance with volumes of this journal (chronology of recording). Therefore all descriptions have numbers showing first the number of volume and then the number of record. The bibliography has at the end the detailed subject index consisting of 465 main headings and a lot of qualifiers. Some of them are descriptors taken from ''INIS Atomindex'' and some are key words taken from natural language. The index is in English as descriptions in the bibliography. (author)

  2. Accident prevention programme

    International Nuclear Information System (INIS)

    1978-01-01

    This study by the Steel Industry Safety and Health Commission was made within the context of the application by undertakings of the principles of accident and disease prevention previously adopted by the said Commission. It puts forward recommendations for the effective and gradual implementation of a programme of action on occupational health and safety in the various departments of an undertaking and in the undertaking as a whole. The methods proposed in this study are likely to be of interest to all undertakings in the metallurgical industry and other industrial sectors

  3. Serious accident in Peru

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A peruvian man, victim of an important accidental irradiation arrived on the Saturday twenty ninth of may 1999 to the centre of treatment of serious burns at the Percy military hospital (Clamart -France). The accident spent on the twentieth of February 1999, on the site of a hydroelectric power plant, in construction at 300 km at the East of Lima. The victim has picked up an industrial source of iridium devoted to gamma-graphy operations and put it in his back pocket; of trousers. The workman has serious radiation burns. (N.C.)

  4. Credible investigation of air accidents

    International Nuclear Information System (INIS)

    Smart, K.

    2004-01-01

    Within the United Kingdom the Air Accidents Investigation Branch (AAIB) has been used as a model for the other transport modes accident investigation bodies. Government Ministers considered that the AAIB's approach had established the trust of the public and the aviation industry in its ability to conduct independent and objective investigations. The paper will examine the factors that are involved in establishing this trust. They include: the investigation framework; the actual and perceived independence of the accident investigating body; the aviation industry's safety culture; the qualities of the investigators and the quality of their liaison with bereaved families those directly affected by the accidents they investigate

  5. 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)

  6. Effect of alternative aging and accident simulations on polymer properties

    International Nuclear Information System (INIS)

    Bustard, L.D.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

    1985-05-01

    The influence of accident irradiation, steam, and chemical spray exposures on the behavior of twenty-three age-preconditioned polymer sample sets (twenty-one different materials) has been investigated. The test program varied the following conditions: (1) Accident simulations of irradiation and thermodynamic (steam and chemical spray) conditions were performed both sequentially and simultaneously. (2) Accident thermodynamic (steam and chemical spray) exposures were performed both with and without air present during the exposures. (3) Sequential accident irradiations were performed both at 28 0 C and 70 0 C. (4) Age preconditioning was performed both sequentially and simultaneously. (5) Sequential aging irradiations were performed both at 27 0 C and 70 0 C. (6) Sequential aging exposures were performed using two sequences: (1) thermal followed by irradiation and (2) irradiation followed by thermal. We report both general trends applicable to a majority of the tested materials as well as specific results for each polymer. Our data base consists of ultimate tensile properties at the completion of the accident exposure for three XLPO and XLPE, five EPR and EPDM, two CSPE (HYPALON), one CPE, one VAMAC, one polydiallylphtalate, and one PPS material. We also report bend test results at completion of the accident exposures for two TEFZEL materials and permanent set after compression results for three EPR, one VAMAC, one BUNA N, one SILICONE, and one VITON material

  7. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    International Nuclear Information System (INIS)

    Park, S. Y.; Song, Y. M.

    2015-01-01

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building

  8. 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

  9. 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.

  10. 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

  11. Validation of severe accident management guidance for the wolsong plants

    International Nuclear Information System (INIS)

    Park, S. Y.; Jin, Y. H.; Kim, S. D.; Song, Y. M.

    2006-01-01

    Full text: Full text: The severe accident management(SAM) guidance has been developed for the Wolsong nuclear power plants in Korea. The Wolsong plants are 700MWe CANDU-type reactors with heavy water as the primary coolant, natural uranium-fueled pressurized, horizontal tubes, surrounded by heavy water moderator inside a horizontal calandria vessel. The guidance includes six individual accident management strategies: (1) injection into primary heat transport system (2) injection into calandria vessel (3) injection into calandria vault (4) reduction of fission product release (5) control of reactor building condition (6) reduction of reactor building hydrogen. The paper provides the approaches to validate the SAM guidance. The validation includes the evaluation of:(l) effectiveness of accident management strategies, (2) performance of mitigation systems or components, (3) calculation aids, (4) strategy control diagram, and (5) interface with emergency operation procedure and with radiation emergency plan. Several severe accident sequences with high probability is selected from the plant specific level 2 probabilistic safety analysis results for the validation of SAM guidance. Afterward, thermal hydraulic and severe accident phenomenological analyses is performed using ISAAC(Integrated Severe Accident Analysis Code for CANDU Plant) computer program. Furthermore, the experiences obtained from a table-top-drill is also discussed

  12. Development of the methodology and approaches to validate safety and accident management

    International Nuclear Information System (INIS)

    Asmolov, V.G.

    1997-01-01

    The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987-1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed. (orig.)

  13. Iodine behaviour in severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Dutton, L M.C.; Grindon, E; Handy, B J; Sutherland, L [NNC Ltd., Knutsford (United Kingdom); Bruns, W G; Sims, H E [AEA Technology, Harwell (United Kingdom); Dickinson, S [AEA Technology, Winfrith (United Kingdom); Hueber, C; Jacquemain, D [IPSN/CEA, Cadarache, Saint Paul-Lez-Durance (France)

    1996-12-01

    A description is given of analyses which identify which aspects of the modelling and data are most important in evaluating the release of radioactive iodine to the environment following a potential severe accident at a PWR and which identify the major uncertainties which affect that release. Three iodine codes are used namely INSPECT, IODE and IMPAIR, and their predictions are compared with those of the PSA code MAAP. INSPECT is a mechanistic code which models iodine behaviour in the aqueous aerosol, spray water and sump water, and the partitioning of volatile species between the aqueous phases and containment gas space. Organic iodine is not modelled. IODE and IMPAIR are semi-empirical codes which do not model iodine behaviour in the aqueous aerosol, but model organic iodine. The fault sequences addressed are based on analyses for the Sizewell `B` design. Two types of sequence have been analysed.: (a) those in which a major release of fission products from the primary circuit to the containment occur, e.g. a large LOCAS, (b) those where the release by-passes the containment, e.g. a leak into the auxiliary building. In the analysis of the LOCA sequences where the pH of the sump is controlled to be a value of 8 or greater, all three codes predict that the oxidation of iodine to produce gas phase species does not make a significant contribution to the source term due to leakage from the reactor building and that the latter is dominated by iodide in the aerosol. In the case where the pH of the sump is not controlled, it is found that the proportion of gas phase iodine increases significantly, although the cumulative leakage predicted by all three codes is not significantly different from that predicted by MAAP. The radiolytic production of nitric acid could be a major factor in determining the pH, and if the pH were reduced, the codes predict an increase in gas phase iodine species leaked from the containment. (author) 4 figs., 7 tabs., 13 refs.

  14. Improving performance with accident tolerant-fuels

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

    2017-01-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO 2 - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO 2 - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U 3 Si 2 , UN, and UC contain a higher uranium density and thermal conductivity than UO 2 , providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U 3 Si 2 , UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO 2 . These ATFs also have thermal conductivities approximately four times higher than that of UO 2 . A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO 2 - Zr as the fuel system of choice. (author)

  15. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    accidents and, secondly, the physical phenomena, studies and analyses described in Chapters 5 to 8. Chapter 5 is devoted to describing the physical phenomena liable to occur during a core melt accident, in the reactor vessel and the reactor containment. It also presents the sequence of events and the methods for mitigating their impact. For each of the subjects covered, a summary of the physical phenomena involved is followed by a description of the past, present and planned experiments designed to study these phenomena, along with their modelling, the validation of which is based on the test results. The chapter then describes the computer codes that couple all of the models and provide the best current state of knowledge of the phenomena. Lastly, this knowledge is reviewed while taking into account the gaps and uncertainties, and the outlook for the future is presented, notably regarding experimental programmes and the development of modelling and numerical simulation tools. Chapter 6 focuses on the behaviour of the containment enclosures during a core melt accident. After summarising the potential leakage paths of radioactive substances through the different containments in the case of the accidents chosen in the design phase, it presents the studies of the mechanical behaviour of the different containments under the loadings that can result from the hazards linked with the phenomena described in Chapter 5. Chapter 6 also discusses the risks of containment building bypass in a core melt accident situation. Chapter 7 presents the lessons learned regarding the phenomenology of core melt accidents and the improvement of nuclear reactor safety. Lastly, Chapter 8 presents a review of development and validation efforts regarding the main computer codes dealing with 'severe accidents', which draw on and build upon the knowledge mainly acquired through the research programmes: ASTEC (IRSN and GRS), MAAP-4 (FAI (US)) and used by EDF and by utilities in many other

  16. 1976 Hanford americium accident

    International Nuclear Information System (INIS)

    Heid, K.R.; Breitenstein, B.D.; Palmer, H.E.; McMurray, B.J.; Wald, N.

    1979-01-01

    This report presents the 2.5-year medical course of a 64-year-old Hanford nuclear chemical operator who was involved in an accident in an americium recovery facility in August 1976. He was heavily externally contaminated with americium, sustained a substantial internal deposition of this isotope, and was burned with concentrated nitric acid and injured by flying debris about the face and neck. The medical care given the patient, including the decontamination efforts and clinical laboratory studies, are discussed. In-vivo measurements were used to estimate the dose rates and the accumulated doses to body organs. Urinary and fecal excreta were collected and analyzed for americium content. Interpretation of these data was complicated by the fact that the intake resulted both from inhalation and from solubilization of the americium embedded in facial tissues. A total of 1100 μCi was excreted in urine and feces during the first 2 years following the accident. The long-term use of diethylenetriaminepentate (DTPA), used principally as the zinc salt, is discussed including the method, route of administration, and effectiveness. To date, the patient has apparently experienced no complications attributable to this extensive course of therapy, even though he has been given approximately 560 grams of DTPA. 4 figures, 1 table

  17. Serious reactor accidents reconsidered

    International Nuclear Information System (INIS)

    1987-12-01

    The chance is determined for damage of the reactor core and that sequel events will cause excursion of radioactive materials into the environment. The gravity of such an accident is expressed by the source term. It appears that the chance for such an accident varies with the source term. In general it is valid that how larger the source term how smaller the chance is for it and vice versa. The chance for excursion is related to two complexes of events: serious damage (meltdown) of the reactor core, and the escape of the liberated radionuclides into the environment. The results are an order of magnitude consideration of the relation between the extent of the source term and the chance for it. From the spectrum of possible source terms three representative ones have been chosen: a large, a medium and a relative small source term. This choice is in accordance with international considerations. The hearth of this study is the estimation of the chance for occurrence of the three chosen source terms for new light-water reactors. refs.; figs.; tabs

  18. 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

  19. Accident analysis. A review of the various accidents classifications

    International Nuclear Information System (INIS)

    Martin Martin, L.; Figueras, J.M.

    1982-01-01

    The objective of the accident analysis, in relation with the safety evaluation, environmental impact and emergency planning, should be to identify the total risk to the population and workers from potential accidents in the facility, analizing it over full spectrum of severity. (auth.)

  20. [The initial (I and II) and advanced (III and IV) stages of juvenile patellar chondromalacia. Its diagnosis by magnetic resonance using a 1.5-T magnet with FLASH sequences].

    Science.gov (United States)

    Macarini, L; Rizzo, A; Martino, F; Zaccheo, N; Angelelli, G; Rotondo, A

    1998-06-01

    Juvenile patellar chondromalacia is a common orthopedic disorder which can mimic other conditions; early diagnosis is mandatory to prevent its evolution into osteoarthrosis. In the early stages of patellar chondromalacia (I and II), the lesions originate in the deep cartilage layer and the joint surface is not affected. Arthroscopy can demonstrate joint surface changes only and give indirect information about deeper lesions. We investigated the yield of 2D FLASH MRI with 30 degrees flip angle and a dedicated coil in the diagnosis of patellar chondromalacia, especially in its early stages. Eighteen patients (mean age: 21 years) with clinically suspected patellar chondromalacia were examined with MRI; 13 of them were also submitted to arthroscopy. A 1.5 T unit with a transmit-and-receive extremity coil was used. We acquired T1 SE sequences (TR/TE: 500-700/15/20) and 2D T2* FLASH sequence (TR/TE/FA: 500-800/18/30 degrees). The field of view was 160-180 mm and the matrix 192 x 256, with 2-3 NEX. The images were obtained on the axial plane. The lesions were classified in 4 stages according to Shahriaree classification. Agreement between MR and arthroscopic findings was good in both early and advanced lesions in 12/13 cases. Early lesions appeared as hyperintense focal thickening of the hyaline cartilage (stage I) or as small cystic lesions within the cartilage and no articular surface involvement (stage II). The medial patellar facet was the most frequent site. Advanced lesions appeared as articular surface ulcerations, thinning and cartilage hypointensity (stage III); stage IV lesions presented as complete erosions of the hyaline cartilage and hypointense underlying bone. 2D FLASH MRI with 30 degrees flip angle can show the differences in water content in the cartilage and thus permit to detect early chondromalacia lesions in the deep cartilage.

  1. The Eight and a Half Year Journey of Undiagnosed AD: Gene Sequencing and Funding of Advanced Genetic Testing Has Led to Hope and New Beginnings

    Directory of Open Access Journals (Sweden)

    Illana Gozes

    2017-05-01

    Full Text Available BackgroundActivity-dependent neuroprotective protein (ADNP is one of the most prevalent de novo mutated genes in syndromic autism spectrum disorders, driving a general interest in the gene and the syndrome.AimThe aim of this study was to provide a detailed developmental case study of ADNP p.Tyr719* mutation toward improvements in (1 diagnostic procedures, (2 phenotypic scope, and (3 interventions.MethodsLongitudinal clinical and parental reports.ResultsAD (currently 11-year-old had several rare congenital anomalies including imperforate anus that was surgically repaired at 2 days of age. Her findings were craniofacial asymmetries, global developmental delay, autistic behaviors (loss of smile and inability to make eye contact at the age of 15 months, and slow thriving as she gradually matures. Comprehensive diagnostic procedures at 3 years resulted in no definitive diagnosis. With parental persistence, AD began walking at 3.5 years (skipping crawling. At the age of 8.5 years, AD was subjected to whole exome sequencing, compared to the parents and diagnosed as carrying an ADNP p.Tyr719* mutation, a causal recurring mutation in ADNP (currently ~17/80 worldwide. Brain magnetic resonance imaging demonstrated mild generalized cerebral volume loss with reduced posterior white matter. AD is non-verbal, communicating with signs and word approximations. She continues to make slow but forward developmental progress, and her case teaches newly diagnosed children within the ADNP Kids Research Foundation.ConclusionThis case study emphasizes the importance of diagnosis and describes, for the first time, early motor intervention therapies. Detailed developmental profile of selected cases leads to better treatments.

  2. Coffin-Siris syndrome and related disorders involving components of the BAF (mSWI/SNF) complex: historical review and recent advances using next generation sequencing.

    Science.gov (United States)

    Kosho, Tomoki; Miyake, Noriko; Carey, John C

    2014-09-01

    This issue of Seminars in Medical Genetics, American Journal of Medical Genetics Part C investigates the human diseases caused by mutations in the BAF complex (also known as the mammalian SWI/SNF complex) genes, particularly focusing on Coffin-Siris syndrome (CSS). CSS is a rare congenital malformation syndrome characterized by developmental delay or intellectual disability (ID), coarse facial appearance, feeding difficulties, frequent infections, and hypoplasia/aplasia of the fifth fingernails and fifth distal phalanges. In 2012, 42 years after the first description of CSS in 1970, five causative genes (SMARCB1, SMARCE1, SMARCA4, ARID1A, ARID1B), all encoding components of the BAF complex, were identified as being responsible for CSS through whole exome sequencing and pathway-based genetic screening. The identification of two additional causative genes (PHF6, SOX11) followed. Mutations in another BAF complex gene (SMARCA2) and (TBC1D24) were found to cause clinically similar conditions with ID, Nicolaides-Baraitser syndrome and DOORS syndrome, respectively. Also, ADNP was found to be mutated in an autism/ID syndrome. Furthermore, there is growing evidences for germline or somatic mutations in the BAF complex genes to be causal for cancer/cancer predisposition syndromes. These discoveries have highlighted the role of the BAF complex in the human development and cancer formation. The biology of BAF is very complicated and much remains unknown. Ongoing research is required to reveal the whole picture of the BAF complex in human development, and will lead to the development of new targeted therapies for related disorders in the future. © 2014 Wiley Periodicals, Inc.

  3. 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

  4. 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

  5. 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

  6. 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.

  7. A risk-based evaluation of the impact of key uncertainties on the prediction of severe accident source terms - STU

    International Nuclear Information System (INIS)

    Ang, M.L.; Grindon, E.; Dutton, L.M.C.; Garcia-Sedano, P.; Santamaria, C.S.; Centner, B.; Auglaire, M.; Routamo, T.; Outa, S.; Jokiniemi, J.; Gustavsson, V.; Wennerstrom, H.; Spanier, L.; Gren, M.; Boschiero, M-H; Droulas, J-L; Friederichs, H-G; Sonnenkalb, M.

    2001-01-01

    The purpose of this project is to address the key uncertainties associated with a number of fission product release and transport phenomena in a wider context and to assess their relevance to key severe accident sequences. This project is a wide-based analysis involving eight reactor designs that are representative of the reactors currently operating in the European Union (EU). In total, 20 accident sequences covering a wide range of conditions have been chosen to provide the basis for sensitivity studies. The appraisal is achieved through a systematic risk-based framework developed within this project. Specifically, this is a quantitative interpretation of the sensitivity calculations on the basis of 'significance indicators', applied above defined threshold values. These threshold values represent a good surrogate for 'large release', which is defined in a number of EU countries. In addition, the results are placed in the context of in-containment source term limits, for advanced light water reactor designs, as defined by international guidelines. Overall, despite the phenomenological uncertainties, the predicted source terms (both into the containment, and subsequently, into the environment) do not display a high degree of sensitivity to the individual fission product issues addressed in this project. This is due, mainly, to the substantial capacity for the attenuation of airborne fission products by the designed safety provisions and the natural fission product retention mechanisms within the containment

  8. Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

    International Nuclear Information System (INIS)

    Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

    2015-01-01

    The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

  9. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  10. 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)

  11. Fukushima accident - reasons and impacts

    International Nuclear Information System (INIS)

    Slugen, V.

    2011-01-01

    The Fukushima accident influenced dramatically the current view on safety of nuclear facilities. Consideration about possible impacts of natural catastrophe in design of nuclear facilities seems to be much more important than before. European commission is focused on the stress-tests at nuclear power plants. His paper will go more in details having in mind reasons and impacts of Fukushima accident (Author)

  12. 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.

  13. Occupational accidents among mototaxi drivers.

    Science.gov (United States)

    Amorim, Camila Rego; de Araújo, Edna Maria; de Araújo, Tânia Maria; de Oliveira, Nelson Fernandes

    2012-03-01

    The use of motorcycles as a means of work has contributed to the increase in traffic accidents, in particular, mototaxi accidents. The aim of this study was to estimate and characterize the incidence of occupational accidents among the mototaxis registered in Feira de Santana, BA. This is a cross-sectional study with descriptive and census data. Of the 300 professionals registered at the Municipal Transportation Service, 267 professionals were interviewed through a structured questionnaire. Then, a descriptive analysis was conducted and the incidence of accidents was estimated based on the variables studied. Relative risks were calculated and statistical significance was determined using the chi-square test and Fisher's exact test, considering p accidents were observed in 10.5% of mototaxis. There were mainly minor injuries (48.7%), 27% of them requiring leaves of absence from work. There was an association between the days of work per week, fatigue in lower limbs and musculoskeletal complaints, and accidents. Knowledge of the working conditions and accidents involved in this activity can be of great importance for the adoption of traffic education policies, and to help prevent accidents by improving the working conditions and lives of these professionals.

  14. B4C control rod behavior during severe accident sequences

    International Nuclear Information System (INIS)

    Steinbrueck, M.

    2003-01-01

    The oxidation kinetics of various types of boron carbides (pellets, powder) as well as the degradation of B 4 C control rod segments were investigated in the temperature range between 800 and 1600 deg C. Mass spectrometric gas analysis was used to determine oxidation rates in transient and isothermal tests. The oxidation kinetics of boron carbide are determined by the formation of a liquid boron oxide layer and its loss due to the reaction with surplus steam to form volatile boric acids and at temperatures above 1500 deg C by direct evaporation. Under these test conditions linear oxidation kinetics are established soon after oxidation has initiated. The oxidation kinetics are strongly influenced by the thermal-hydraulic boundary conditions, in particular by the steam flow rate. Only very low amounts of methane were ever produced in these tests. Enhanced degradation of B 4 C control rods starts with the rapid formation of eutectic melts in the systems B 4 C-stainless steel (SS) and SS-Zircaloy at temperatures above 1250 deg C. Initially, this melt is kept within a ZrO 2 scale externally formed at the Zircaloy guide tube. The absorber melt is rapidly oxidized after failure of the oxide shell and aggressively attacks adjacent fuel claddings. (author)

  15. 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

  16. 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.

  17. Barriers to learning from incidents and accidents

    NARCIS (Netherlands)

    Dechy, N.; Dien, Y.; Drupsteen, L.; Felicio, A.; Cunha, C.; Roed-Larsen, S.; Marsden, E.; Tulonen, T.; Stoop, J.; Strucic, M.; Vetere Arellano, A.L.; Vorm, J.K.J. van der; Benner, L.

    2015-01-01

    This document provides an overview of knowledge concerning barriers to learning from incidents and accidents. It focuses on learning from accident investigations, public inquiries and operational experience feedback, in industrial sectors that are exposed to major accident hazards. The document

  18. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

  19. 29 CFR 1960.29 - Accident investigation.

    Science.gov (United States)

    2010-07-01

    ... reflective of the seriousness of the accident. (b) In any case, each accident which results in a fatality or... evidence uncovered during accident investigations which would be of benefit in developing a new OSHA...

  20. Expression profiles of cancer stem cell markers: CD133, CD44, Musashi-1 and EpCAM in the cardiac mucosa-Barrett's esophagus-early esophageal adenocarcinoma-advanced esophageal adenocarcinoma sequence.

    Science.gov (United States)

    Mokrowiecka, Anna; Veits, Lothar; Falkeis, Christina; Musial, Jacek; Kordek, Radzislaw; Lochowski, Mariusz; Kozak, Jozef; Wierzchniewska-Lawska, Agnieszka; Vieth, Michael; Malecka-Panas, Ewa

    2017-03-01

    Barrett's esophagus (BE), which develops as a result of gastroesophageal reflux disease, is a preneoplastic condition for esophageal adenocarcinoma (EAC). A new hypothesis suggests that cancer is a disease of stem cells, however, their expression and pathways in BE - EAC sequence are not fully elucidated yet. We used a panel of putative cancer stem cells markers to identify stem cells in consecutive steps of BE-related cancer progression. Immunohistochemistry was performed on formalin-fixed, paraffin-embedded blocks from 58 patients with normal cardiac mucosa (n=5), BE (n=14), early EAC (pT1) from mucosal resection (n=17) and advanced EAC (pT1-T4) from postoperative specimens (n=22). Expression of the CD133, CD44, Musashi-1 and EpCAM was analyzed using respective monoclonal antibodies. All markers showed a heterogeneous expression pattern, mainly at the base of the crypts of Barrett's epithelium and EAC, with positive stromal cells in metaplastic and dysplastic lesions. Immuno-expression of EpCAM, CD44 and CD133 in cardiac mucosa was significantly lower (mean immunoreactivity score (IRS)=1.2; 0.0; 0.4; respectively) compared to their expression in Barrett's metaplasia (mean IRS=4.3; 0.14; 0.7; respectively), in early adenocarcinoma (mean IRS=4.4; 0.29; 1.3; respectively) and in advanced adenocarcinoma (mean IRS=6.6; 0.7; 2.7; respectively) (p<0.05). On the contrary, Musashi-1 expression was higher in BE and early ADC compared to GM and advanced ADC (NS). Our results suggest that the stem cells could be present in premalignant lesions. EpCAM, CD44 and CD133 expression could be candidate markers for BE progression, whereas Musashi-1 may be a marker of the small intestinal features of Barrett's mucosa. Copyright © 2016 Elsevier GmbH. All rights reserved.